ML18124A094

Redacted - Susquehanna Steam Electric Station, Units 1 & 2, Revision 68 to Final Safety Analysis Report, Chapter 12, Ensuring That Occupational Radiation Exposures Are as Low as Reasonably Achievable

ML18124A094
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Site:	Susquehanna
Issue date:	10/16/2017
From:	Susquehanna, Talen Energy
To:	Office of Nuclear Reactor Regulation
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SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA)

12.1.1 POLICY CONSIDERATIONS

12.1.1.1 Management Policy It is the policy of PP&L to maintain occupational radiation exposure As Low As Reasonably Achievable (ALARA) at the Susquehanna SES. This includes maintaining the annual dose to individuals working at the station ALARA, and keeping the annual integrated dose to station personnel ALARA. The management of this Company is firmly committed to performing all reasonable actions to ensure that radiation exposures are maintained ALARA.

Subsection 12.1.2 and Section 12.3 discuss the ALARA considerations that have been incorporated into the design of the Susquehanna SES.

Susquehanna SES will be operated and maintained in such a manner as to ensure occupational radiation exposures (ORE) are ALARA. The operational ALARA program is described in Section 12.5. Training programs will be established to assure personnel understand both why and how occupational radiation exposures will be maintained ALARA. A Station ALARA Committee has been established to ensure implementation of ALARA policy by various program

reviews.

12.1.1.2 Management Responsibilities

Figures 17.2-2 and 13.1-2 exhibit the management organizational structure for the

Susquehanna SES.

The Vice President-Nuclear Operations has the corporate responsibility for the ALARA program. The responsibility for the coordination and administration of the ALARA program is assigned to the General Manager-Nuclear Engineering and the General Manager-Susquehanna SES and their reports. They are responsible to ensure the policies and commitments contained in the PP&L ALARA Program are being properly implemented and maintained.

During the design and construction phase, the Susquehanna SES Project Manager is responsible to ensure that the design and construction of the facility is such that occupational exposures will be ALARA. This will include ensuring that, to the extent practicable:

a. Design concepts and station features reflect consideration of the activities of station personnel that might be anticipated and that might lead to personnel exposure to substantial sources of radiation and that station design features have been provided to reduce the anticipated exposures of station personnel to these sources of radiation.

b. Specifications for equipment reflect the objectives of ALARA, including among others, considerations of reliability, serviceability and limitations of internal accumulations of radioactive material.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-2 During the operational phase, the General Manager-Nuclear Engineering is responsible to ensure that the Station design remains in compliance with all applicable radiation safety standards found in 10CFR20, 10CFR50, 10CFR50.67, 40CFR190 and applicable regulatory guidance documents.

The Manager-Nuclear Modifications is responsible for ensuring the PP&L ALARA program is incorporated into the design of plant modifications and new facilities related to the SSES.

The Manager-Nuclear Technology is responsible to ensure that a radiation protection staff with health physics and radiological expertise is adequately maintained to support Nuclear Engineering, Susquehanna SES plant staff and other functional group activities as appropriate.

During the startup and operation phase, the General Manager-Susquehanna SES is responsible for ensuring radiation exposure is controlled in a manner consistent with ALARA requirements and is specifically responsible for the onsite radiation protection program. Additionally he is responsible for:

a. Ensuring support from all station personnel for the implementation of the Station ALARA program

b. Providing management oversight of the accumulation of personnel exposure at SSES and reviewing and concurring with annual personnel exposure goals

c. Ensuring resources needed to achieve ALARA goals and objectives are made available.

Additional General Manager-Susquehanna SES responsibilities are implemented through the Health Physics Supervisor.

Major ALARA responsibilities of the Health Physics Supervisor or designee, include the

following:

a. Participating in reviews of design changes for facilities and equipment that can affect potential radiation exposures;

b. Identifying locations, operations, and conditions, that have the potential for causing significant exposures to radiation;

c. Initiating and implementing an exposure control program which includes the establishment of manrem goals,

d. Developing plans, procedures, and methods for keeping radiation exposures of station personnel ALARA,

e. Reviewing, commenting on, and recommending changes in applicable procedures to maintain exposures ALARA;

f. Developing or participating in the development of appropriate Health Physics training programs related to work in radiation areas or involving radioactive material;

g. Supervising the radiation surveillance program to maintain data on exposures of and doses to station personnel by specific job functions and type of work; SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-3 h. Supervising the collection, analysis, and evaluation of data and information attained from radiological surveys and monitoring activities;

i. Supervising, training, and qualifying the radiation protection staff of the station; and

j. Ensuring that adequate radiation protection coverage is provided for station personnel during all working hours.

Chapter 13 provides additional information concerning responsibilities and reporting relationships at the Susquehanna SES.

12.1.1.3 Policy Implementation The management ALARA policy is implemented at the Susquehanna SES by the Health Physics Staff under the direction of the General Manager-Susquehanna SES and the Health Physics Supervisor. The policy implementation is formalized by the incorporation of ALARA philosophy and considerations into permanent plant procedures dealing specifically with ALARA concerns. The operational ALARA considerations identified in Subsections 12.1.3 and 12.5.3.2 are implemented by these procedures.

Subsection 12.5.3.7 describes the training program established to give appropriate station personnel the necessary knowledge to understand why and how they should maintain their ORE ALARA.

The Station ALARA Committee has been established to review the implementation of the Company ALARA Program. Specific responsibilities of the Station ALARA Committee include:

a. Assuring the effectiveness of the ALARA program as implemented at the Susquehanna SES.

b. Assuring that high exposure maintenance and modification tasks receive proper management attention ensuring they are planned in accordance with sound ALARA principles.

c. Reviewing, prioritizing and recommending potential action items for inclusion into the Nuclear Department long-term exposure reduction plan and monitoring status of action items included in the plan.

d. Administering the Employee ALARA Concerns program.

e. Assuring Station activities are conducted in an ALARA manner maintaining the balance between cost, schedule and personnel exposure.

f. Reviewing exposure goals, monitoring performance against these goals and taking action as appropriate when goals are jeopardized.

g. Monitoring individual personnel exposures to ensure they are minimized to the extent possible while maintaining overall collective exposures ALARA.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-4 12.1.2 DESIGN CONSIDERATIONS This subsection discusses the methods and features by which the policy considerations of Subsection 12.1.1 are applied. Provisions and designs for maintaining personnel exposures as low as reasonably achievable are presented in Subsections 12.3.1, 12.3.2 and 12.5.3.

Experiences and data from operating plants are evaluated to decide if and how equipment or facility designs could be improved to reduce overall plant personnel exposures. During plant design, operating reports and data such as that given in WASH 1311, NUREG-75/032, NUREG-109 and Compilation and Analysis of Data on occupational Radiation Exposure Experienced at Operating Nuclear Power Plants, AIF, September 1974, References 12.1-1, through 12.1-4 respectively, were reviewed to determine which operations, procedures or types of equipment were most significant in producing personnel exposures. Methods to mitigate such exposures were implemented wherever possible and practicable.

General design considerations and methods employed to keep in-plant radiation exposures ALARA have two objectives:

a) Minimizing the necessity for the amount of personnel time spent to radiation areas; and b) Minimizing radiation levels in routinely occupied plant areas and in the vicinity of plant equipment expected to require personnel attention.

Both equipment and facility designs are considered in keeping exposures ALARA during plant operations including normal operation, maintenance and repairs, refueling operations and fuel storage, in-service inspection and calibrations, radioactive waste handling and disposal, and other events of moderate frequency. The actual design features used are described in

Subsection 12.3.1.

12.1.2.2 Equipment General Design Considerations for ALARA The following equipment general design considerations to minimize the necessity for and amount of personnel time spent in a radiation area include, where practicable:

a) Reliability, durability, construction, and design features of equipment, components, and materials to reduce or eliminate the need for repair or preventive maintenance;

b) Servicing convenience including ease of disassembly and modular ization of components for replacement or removal to a lower radiation area for repair; c) Provisions, where practicable, to remotely or mechanically operate, repair, service, monitor, or inspect equipment; and d) Redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high and when no feasible method is available to reduce

radiation levels.

The following equipment general design considerations directed toward minimizing radiation levels proximate to equipment or components requiring personnel attention include, where practicable:

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-5 a) Provision for draining, flushing, or, if necessary, remote cleaning of equipment containing radioactive material;

b) Design of equipment, to minimize the buildup of radioactive material and to facilitate flushing of crud traps;

c) Utilization of high quality valves, valve packings, and gaskets to minimize leakage and spillage of radioactive materials;

d) Provisions for minimizing the spread of contamination into equipment service areas; and

e) Provisions for isolating equipment from radioactive process fluids.

12.1.2.3 Facility Layout General Design Considerations for ALARA Facility general design considerations to minimize the amount of personnel time spent in radiation areas include where practicable:

a) Locating equipment and instruments, which will require routine maintenance, calibration, or inspection for ease of access and a minimum of required occupancy time in radiation

fields; b) Arranging plant areas to allow remote or mechanical operation, service, monitoring, or inspection of highly radioactive equipment; and

c) Providing, for transportation of equipment or components requiring service to a lower radiation area.

Facility general design considerations directed toward minimizing radiation levels in plant access areas and in the vicinity of equipment requiring personnel attention include, where practicable:

a) Separating radiation sources and occupied areas (e.g., pipes containing potentially highly radioactive fluids do not pass through normally occupied areas);

b) Providing adequate shielding between radiation sources and access and service areas;

c) Locating appropriate equipment, instruments, and sampling sites in the lowest practicable radiation zone;

d) Providing means and adequate space for using movable shielding for sources within the service area when required; and

e) Providing means (e.g., curbing, drains and flush) to control contamination and to facilitate decontamination of potentially contaminated areas.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-6 12.1.2.4 ALARA Design Review During the design phase, Bechtel Power Corporation, as agents for PP&L, were given the basic responsibility for the performance of the ALARA design review. PP&L provided overall coordination of and input to this review. ALARA design reviews were completed on all required systems and areas. Recommended design modifications were made. In addition to intensive system/area ALARA design reviews, field routed small piping drawings were continually reviewed, often resulting in changes in routing, valve and operator types, and connection points.

During the operational phase, ALARA considerations are included in the scoping and design phase of modifications and changes to the design of the facility. The General Manager-Nuclear Engineering has the responsibility to:

a. Ensure that engineering personnel are adequately trained in ALARA design and engineering principles so that radiation exposure with respect to installation, operation and maintenance of plant modifications and new facilities is considered in each design.

b. Ensure an integrated level of involvement within the modifications organizations to effectively implement the PP&L ALARA program.

The Manager-Nuclear Modifications is responsible to:

a. Ensure ALARA/dose reduction opportunities are identified for inclusion into the design of modifications and new facilities.

b. Ensure the PP&L ALARA program is incorporated into the design of plant modifications and new facilities related to the Susquehanna SES.

The Supervisor-Operations Technology has the responsibility to determine the requirements of the ALARA design process and, in concert with the Health Physics Supervisor, ensure that ALARA design reviews are performed by qualified radiological personnel.

At the scoping phase, potential radiological impacts such as occupational and offsite exposure impacts, and radioactive waste generation are identified and considered during the design phase. Radiological personnel evaluate the proposed action and provide input into the scoping document. During the design phase, ALARA considerations consistent with the guidance contained in Regulatory Guidance 8.8 is ensured via the use of an ALARA design checklist. This checklist is completed by the design engineer and reviewed and approved by designated radiological personnel.

Procedures have been developed describing the training programs and requirements for personnel involved in the ALARA design process.

12.1.3 OPERATIONAL CONSIDERATIONS

To assure that the occupational radiation exposures are maintained as low as reasonably achievable (ALARA) during the operation of Susquehanna SES specific activities will be implemented.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-7 12.1.3.1 Procedure Development Station procedures will be prepared, reviewed, and approved in accordance with Section 13.5.

12.1.3.1.1 ALARA Procedures

To assure adequate emphasis on the necessity to minimize personnel exposures, ALARA procedures will be prepared. These procedures implement considerations of such topics as ALARA review of applicable Radiation Work Permits (RWP), worker feedback, special task training and evaluation of proposed changes in applicable facilities or equipment.

12.1.3.1.2 Station Procedures

Administrative requirements will be implemented to assure that applicable procedures developed by other plant disciplines have adequately incorporated the principle of minimizing personnel exposure. Station administrative documents will describe the criteria for selection of those procedures and revisions that will be reviewed by Health Physics. Recommendations made by Health Physics will normally be resolved with the appropriate plant discipline prior to submission for final review and approval.

12.1.3.2 Station Organization As described in Subsection 12.5.1, the Station organization provides the Health Physics Supervisor direct access to the Vice President-Nuclear Operations to assure uniform support of Health Physics and ALARA requirements. This organization will allow the Vice President-Nuclear Operations direct involvement in the review and approval of specific ALARA goals and objectives as well as review of data and dissemination of information related to the ALARA program.

The organization also provides a Health Physics Specialist-ALARA who is normally free from routine Health Physics activities to implement the Station ALARA program. This individual is primarily responsible for coordination of Station ALARA activities and will routinely interface with first line supervision in radiation work planning and post job review.

12.1.3.3 Operating Experience

The Radiation Work Permit process described in Subsection 12.5.3.2 will provide a mechanism for collection and evaluation of data relating to personnel exposure. Information collated by systems and/or components and job function will assist in evaluating design or procedure changes intended to minimize future radiation exposures.

12.1.3.4 Exposure Reduction

Specific exposure reduction techniques that will be employed at Susquehanna SES are described in Subsection 12.5.3.2. Procedures will assure that applicable station activities are completed with adequate preparation and planning; work is performed with appropriate Health SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-8 Physics recommendations and support; and results of post job data evaluation are applied to implement improvements.

In addition, the Health Physics staff, will at all times be vigilant for ways to reduce exposures by soliciting employee suggestions, evaluating origins of plant exposures, investigating unusual exposures, and assuring that adequate supplies and instrumentation are available.

PP&L management will perform periodic reviews of station programs to assure workers are receiving adequate instruction in ALARA and Health Physics requirements. Implementation of the Health Physics program, selected procedures, and past exposure records will also be reviewed. Management will perform formal re views of the Susquehanna SES Health Physics program at least once every three years and results will be forwarded to the Vice

President-Nuclear Operations and appropriate members of corporate management. The results of management reviews may also include recommendations on mechanisms which may reduce personnel exposure. The Vice President-Nuclear Operations will respond to noted recommendations or deficiencies and corrective action or improvements will be verified during subsequent reviews.

12.

1.4 REFERENCES

12.1-1 T. D. Murphy, WASH-1311, UC-78, "A Compilation of Occupational Radiation Exposure from Light Water Cooled Nuclear Power Plants 1969-1973," USNRC Radiological Assessment Branch, May 1974.

12.1-2 T. D. Murphy, et. al., NUREG-75/032, "Occupational Radiation Exposure at Light Water Cooled Power Reactors 1969-1974," USNRC Radiological Assessment Branch, June 1975.

12.1-3 T. D. Murphy, et. al., NUREG-0109, "Occupational Radiation Exposure at Light Water Cooled Power Reactors 1969-1975," USNRC Radiological Assessment

Branch, August 1976.

12.1-4 C. A. Pelletier, et. al., National Environmental Studies Project, "Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants," Atomic Industrial Forum, September 1974.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-1 12.2 RADIATION SOURCES In this section the sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity required for the design of personnel protective measures and for dose assessment are discussed and identified.

12.2.1 CONTAINED SOURCES The shielding design source terms are based on a noble gas fission product release rate of 0.1 Ci/sec (after 30 minutes decay) and the corresponding fission, activation, and corrosion product concentrations in the primary coolant. The sources in the primary coolant are discussed in Section 11.1 and listed in Tables 11.1-1 through 11.1-5. Throughout most of the primary coolant system, activation products, principally nitrogen-16, are the primary radiation sources for shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while at the same time providing for daughter product formation.

Basic reactor data and core region description used for this section are listed in Tables 12.2-1 through 12.2-5.

The data contained in these Tables is the original design basis for the plant and have not been revised for power uprate. Although doses from the reactor core will increase due to power uprate, these increases have no effect on normal operating doses to plant personnel or on equipment qualification. The increased doses from the core resulting from power uprate meet radiation shielding and zoning requirements inside containment.

In this subsection the design sources are presented by building location and system. General locations of the equipment discussed in this section are shown on the shielding and zoning drawings, Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, and A-530, Sh. 1. The layout of equipment in Unit 2 is not significantly different from the layout in Unit 1, therefore the shielding and zoning requirements shown on these drawings are applicable to both units. Detailed data on source descriptions for each shielded plant area are presented in Tables 12.2-38 through 12.2-40.

Shielding source terms presented in this section and associated tables are based on conservative assumptions regarding system and equipment operations and characteristics to provide reasonably conservative radioactivity concentrations for shielding design. Therefore, SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-2 the shielding source terms are not intended to approximate the actual system design radioactivity concentrations.

12.2.1.1 Drywell 12.2.1.1.1 Reactor Core The primary radiations within the drywell during full power operation are neutron and gamma radiation resulting from the fission process in the core. Tables 12.2-4 and 12.2-5 list the multigroup neutron and gamma ray fluxes at the outside surfaces of the reactor pressure vessel and the primary shield at the core midplane. The gamma fluxes include those resulting from capture or inelastic scattering of neutrons within the reactor pressure vessel and core shroud and the gamma radiation resulting from prompt fission and fission product decay.

The data contained in Tables 12.2-4 and 12.2-5 is the original design basis for the plant and have not been revised for power uprate. Although doses from the reactor core will increase due to power uprate, these increases have no effect on normal operating doses to plant personnel or on equipment qualification. The increased doses from the core resulting from power uprate meet radiation shielding and zoning requirements inside containment.

The largest radiation sources after reactor shutdown are the decaying fission products in the fuel. Tables 12.2-9A1 and 12.2-9A2 list the core gamma sources as a function of shutdown time. Secondary sources are the structural material activation of the RPV, its internals, and the piping and equipment located in the primary containment and also the activated corrosion products accumulated or deposited in the internals of the RPV, the primary coolant piping, and other process system piping in the primary containment.

12.2.1.1.2 Reactor Coolant System Sources of radiation in the reactor coolant system are fission products estimated to be released from fuel and activation and corrosion products that are circulated in the reactor coolant. These sources are listed in Tables 11.1-1 through 11.1-5 and their bases are discussed in Section 11.1. The nitrogen-16 concentration in the reactor coolant is assumed to be 50

µCi/gm of coolant at the reactor recirculation outlet nozzle.

12.2.1.1.3 Primary Steam System Radiation sources in the primary steam system piping include activation gases, principally nitrogen-16, and the corrosion and fission products carried over to the steam system.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-3 The nitrogen-16 concentration in the main steam is assumed to be 250

µCi/gm of steam leaving the reactor vessel at the main steam outlet nozzle. Fission product activity corresponds to an offgas release rate of 100,000

µCi/sec at 30 minutes delay from the reactor steam nozzle. Partition fractions for activity into the steam system are 100 percent for gases, 8percent by weight for halogens, and 0.1 percent by weight for particulates. These partition factors are applied to the reactor water concentrations as given in Table 11.1-2 through 11.1-5.

12.2.1.2 Reactor Building

12.2.1.2.1 Reactor Water Cleanup System

Radiation sources in the RWCU system consist of those radioisotopes carried in the reactor water. Nitrogen-16 is the predominant radiation source in the regenerative and nonregenerative heat exchangers and RWCU pumps and piping. The inventory of N-16 is based upon component transit times, as shown in Table 12.2-6. The main sources for the RWCU filter demineralizers, holding pumps, and the RWCU backwash receiving tank are the accumulated corrosion and fission products, based on the inlet reactor water concentrations given in Section 11.1. Table 12.2-7 provides the inventory of the accumulated isotopes in the filter

demineralizer, and Table 12.2-8 provides the inventory of isotopes in the RWCU backwash receiving tank.

12.2.1.2.2 Spent Fuel Handling and Transfer

The spent fuel assemblies are the predominant source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish near fission product buildup equilibrium for the reactor at rated power is used in determining the source strength. Shielding requirements for spent fuel transfer are based on the fission product activity present 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown. Source terms for spent fuel are discussed in Subsection 12.2.1.3.1 and are listed in Tables 12.2-9A1 and 12.2-9A2.

12.2.1.2.3 Residual Heat Removal System

The pumps, heat exchangers, and associated piping of the Residual Heat Removal (RHR)

System are potential carriers of radioactive materials. For plant shutdown, the RHR pumps and heat exchanger sources result from the radioactive isotopes carried in the reactor coolant, discussed in Subsection 12.2.1.1.2, after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown. The radioactive isotopic concentrations are listed in Table 12.2-10.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-4 12.2.1.2.4 Reactor Core Isolation Cooling System Components of the Reactor Core Isolation Cooling (RCIC) System that are potential radiation sources are the RCIC turbine and steam inlet and exhaust piping. Radioactivity in the turbine and piping is that present in the driving steam that has been extracted from the main steam system. The steam activity as discussed in Subsection 12.2.1.1.3, decayed for the appropriate transit time to the RCIC turbine, is used for the shielding calculations for this system, and is listed in Table 12.2-11.

12.2.1.2.5 High Pressure Coolant Injection System The radiation sources for the High Pressure Coolant Injection System are the HPCI turbine and the steam inlet and exhaust piping. The steam activity, as discussed in Subsections 12.2.1.1.3, decayed for the appropriate transit times is used for the shielding of this system as shown in Table 12.2-11.

12.2.1.2.6 Core Spray Systems Because the core spray, when testing, uses condensate from the condensate storage tank with very low radioactivity concentrations, no shielding is required.

12.2.1.3 Refueling Facilities 12.2.1.3.1 Spent Fuel Storage and Transfer The predominant radiation sources in the spent fuel storage and transfer areas are the spent fuel assemblies. Spent fuel assembly sources are discussed in Subsection 12.2.1.2.2. For shielding design, the spent fuel pool is assumed to contain the design maximum of 2840 fuel assemblies (Section 9.1) with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay. The activity of fission product radionuclides which serve as the basis for the spent fuel source term are shown in Tables 12.2-9A1 and 12.2-9A2.

Spent fuel assemblies are also a radiation source at the Independent Spent Fuel Storage Installation (ISFSI). For shielding design and dose rate determinations each Dry Shielded Canister (DSC) is assumed to contain 61 fuel assemblies with a minimum 5 year decay cooling in the spent fuel pool. Refer to Section 11.7.3 for ISFSI Source Terms.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-5 12.2.1.3.2 Fuel Pool Cooling and Cleanup System Sources in the Fuel Pool Cooling and Cleanup (FPCC) System are primarily a result of transfer of radioactive isotopes from the reactor coolant into the spent fuel pool during refueling operations. The total gross activity in the Fuel Pool water is assumed to be reactor coolant activities for fission, corrosion, and activation products (Tables 11.1-1 through 11.1-5), normalized to 1.0E-2

µCi/cc for shutdown conditions and 1.00E-04 µ Ci/cc for normal operations. This activity then undergoes subsequent decay and accumulation on the FPCC filter demineralizers (see Table 12.2-13). The FPCC filter demineralizer resins are back washed periodically into a backwash receiving tank. Shielding source terms for the backwash receiving tank are shown in Table 12.2-14.

12.2.1.4 Turbine Building 12.2.1.4.1 Primary Steam and Power Conversion Systems

Radiation sources for piping and equipment which contain primary steam are based on the radioactivity carried over into the steam from the reactor coolant and include fission product gases and halogens, corrosion and fission products, and gaseous activation products as discussed in Subsection 12.2.1.1.3. Steam density variations and the steam transit times through equipment and pipes are factored into the source term evaluation to account for volumetric dilution effects, radiological decay, and daughter product generation.

12.2.1.4.2 Condensate System

The sources in the condensate system are based on decayed main steam activities (Subsection 12.2.1.1.3). Eighty percent of the N-16 and 100 percent of the noble gases are assumed to be removed from the condensate system by the main condenser evacuation system. The gaseous activities are minor in the hotwell and negligible in the remainder of the condensate system. The hotwell has a minimum of three minutes holdup of condensate and therefore N-16 activity at the condenser outlet is negligible. Fission products, activated corrosion products, and the daughter products from the decay of fission product gases in transit through the turbine are the inlet sources to the condensate system.

These sources, as shown in Table 12.2-15, are present in the condensate pumps and piping and accumulate on the condensate filters and demineralizer resins. When spent, the condensate filters and demineralizer resins are discharged to the solid waste management system for packaging and offsite disposal. Table 12.2-16 provides the isotopic inventory utilized as the shielding source terms for the condensate demineralizers. Tables 12.2.46, SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-6 12.2.47, and 12.2.48 provide the shielding source terms for the condensate filter vessel, condensate backwash solution, and the condensate filter backwash receiving tank, respectively.

12.2.1.4.3 Offgas System Recombiner

Radioactive sources in the gas treatment system originate with the noble gases and non-condensible gases removed from the main condenser, and the activity entering with the

extraction driving steam to the main condenser evacuation system. The activity removed from the main condenser is based on the primary steam activity as described in Subsection 12.2.1.1.3, decayed for the total transit time to the steam jet air ejector. Eighty percent of the N-16 and 100 percent of the noble gases are assumed to be removed by the air ejector.

Activity in the extraction driving steam to the air ejector is the primary steam activity as described in Subsection 12.2.1.1.3, decayed by the transit time to the air ejector. The total quantity of activity in the offgas pipe and recombiner and source term assumptions are shown in Tables 12.2-17 and 12.2-18.

12.2.1.5 Radwaste Building

12.2.1.5.1 Liquid Waste Management System

The liquid waste management system (LWMS) sources are radioisotopes, including fission and activation products, present in the reactor coolant. The components of the LWMS contain varying degrees of radioactivity depending on the detailed system and equipment design.

The concentrations of radionuclides present in the process fluids at various locations in the system such as pipes, tanks, filters, and demineralizers are discussed in Section 11.2 and are listed in Tables 11.2-5 through 11.2-7. Shielding and associated radiation zoning for components of the LWMS are based on the design basis radioactivity concentrations given in Sections 11.1 and 11.2.

12.2.1.5.2 Solid Waste Management System

Wet and dry radioactive wastes are collected, treated, and stored in the solid radwaste facilities as discussed in Section 11.4. The radioactive wastes are processed by filtration, decanting, and ion-exchange treatment. The resultant volume reduced products (e.g., filter cakes, depleted resins) are dewatered for storage and offsite shipment. Processed liquids may be analyzed for reuse as condensate make-up, processed as radioactive waste, or diluted and discharged.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-7 Radioactive waste may be solidified in containers specified in the Solid Radioactive Waste Process Control Program, then washed to minimize external surface contaminants, and shipped or stored in concrete shielded compartments. The aforementioned operations may be accomplished utilizing remote container loading, transfer, capping facilities, and an overhead crane. Shielding and radiation zone designations of the solid radioactive waste areas are based on the maximum activity sources at zero decay presented in Tables 11.2-6 and 11.4-6, without any consideration of external container shielding credit.

12.2.1.5.3 Ambient Charcoal Offgas Treatment System

The charcoal offgas system as described in Section 11.3 is located in the radwaste building and primarily adsorbs the noble gases and daughter products remaining in the noncondensible gases removed from the main condenser after treatment in the recombiner offgas system.

The shielding of the components is based on the transit times for formation and accumulation of noble gas daughter products and the remaining xenon and krypton gases on the carbon beds.

The gases, after charcoal treatment, pass th rough a post HEPA filter where remaining particulates are trapped prior to exhausting. The concentration of the activity on the piping, equipment, and particulate and charcoal filters for shield design is shown in Tables 12.2-19

through 12.2-24.

12.2.1.6 Sources Resulting from Design Basis Accidents

Radiation sources and shielding requirements resulting from design basis accidents are presented and evaluated in Section 18.1.20 Plant Shielding (II.B.2).

12.2.1.7 Site Boundary N-16 Shine Dose The N-16 present in the reactor steam in the primary steam lines, turbines, and moisture separator can contribute to the site boundary dose as a result of high energy gamma emission.

The turbine shielding was designed to minimize shine dose. The N-16 shine dose rate at the site boundary was calculated based on the final turbine shielding design. The turbine operating floor component N-16 inventories are listed in Table 12.2-11.

12.2.1.8 Stored Radioactivity

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-8 Normally the only sources of activity not stored inside the plant structures are the refueling water storage tank (RWST) and the condensate storage tank (CST). A berm is provided around these tanks to retain leakage and provide for Zone 1 (<0.5 mrem/hr) access. Provisions have been made to recycle the water from both the condensate and refueling water storage tanks to the condensate demineralizer.

Under normal conditions the condensate storage tank contains concentrations of radionuclides that yield a surface exposure rate of less than 1. 5 mr/hr. The condensate storage tank isotopic inventory is shown in Table 12.2-29.

The refueling water storage tank is also expected to have a maximum contact exposure rate of less than1.5 mr/hr.

All spent fuel is stored in the spent fuel pool until it is placed in the spent fuel shipping cask for offsite transport or transferred to the Independent Spent Fuel Storage Installation (ISFSI) described in Section 11.7. Storage space is provided in the radwaste enclosure for interim storage of packaged solid waste materials. Radioactive wastes stored inside the plant structures are shielded in place such that there is normally Zone I access outside the structure.

A separate Steam Dryer Storage Facility (SDSF) is provided within the plant protected area for the storage shielding and radioactive decay of replaced Reactor steam dryers. The steam dryers are cut in half and packaged into steel containers for storage in the SDSF. They are not considered as radioactive waste but are treated as irradiated plant equipment. The SDSF is a

reinforced concrete vault with removable roof slab access only, meeting 10CFR20 dose limits. Low Level Radwaste is stored in the Low Level Radwaste Holding Facility (LLRWHF) and is described in Section 11.6.

12.2.1.9 Special Sources Special materials used in the radiochemistry laboratory and sealed sources used for calibration purposes are of the low activity level and are handled in accordance with station health physics procedures. Unsealed sources and radiochemistry samples are handled in hoods that exhaust to the ventilation system.

The radiation source for the Transverse Incore Probe System (TIP) is provided in Tables 12.2-41, 12.2-42, 12.2-43, and 12.2-44. The radiation source is based upon location within the core and residence time. As indicated in the tables, the TIP system consists of three components for shielding calculations, the fissionable material, non-fissionable material, and the cable. Sources are provided for each component as a function of irradiation and decay times.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-9 The reactor startup source is shipped to the site in a special cask designed for shielding. The source is transferred under water while in the cask and loaded into beryllium containers. This is then loaded into the reactor while remaining under water. The source is stored in the Spent Fuel Pool. Thus, no unique shielding requirements after reactor operation are required.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES

12.2.2.1 Sources of Airborne Radioactivity

The sources of airborne radioactivity are found in the various confined areas of the plant facility and are primarily from the process leakage of the systems carrying radioactive gases, steam, and liquids. Depending on the type of the system and its physical condition, such as system pressures and temperatures, the leakage will be as a gas, steam, liquid, or a mixture of these.

12.2.2.2 Production of Airborne Materials

Radioactive materials become airborne through a number of mechanisms. The primary production mechanisms are spraying, splashing, flashing, evaporation, and diffusion.

12.2.2.3 Locations of Sources of Airborne Radioactivity

The primary sources of airborne radioactivity are found in the reactor, turbine, and radwaste buildings. Within these structures, the radioactivity may be released in equipment cubicles, system compartments, valve and piping galleries, sampling stations, radwaste handling areas, cleaning and decontamination areas and repair shops.

12.2.2.4 Control of Airborne Radioactivity Ventilation is an effective means of controlling airborne radioactive materials. Ventilation flow paths are designed such that air from low potential airborne areas flows toward the higher potential airborne areas. This flow pattern will ensure that activity released in the above mentioned source locations, which usually have low personnel access requirements, will have little chance to escape to areas with a high personnel occupancy such as corridors, working aisles and operating floors.

12.2.2.5 Methodology for Estimating the Concentration of Airborne Radioactive Material Within the Plant

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-10 In order to estimate the airborne radioactive material concentrations at locations within the plant, the following methodology was used:

(1) Estimate the total airborne releases (in Curies per year) for each of the buildings of the plant; (2) Estimate a distribution for these releases among the various equipment areas of each building based on operating data and engineering judgment; (3) Determine the annual exhaust flow from each equipment area, (4) Calculate the resultant airborne radionuclide concentration (µCi/cc) in each equipment area based on the release distribution (Ci/yr) and exhaust flow rate (cc/yr).

The following subsections discuss each step in the above procedure in more detail.

12.2.2.6 Estimation of Total Airborne Releases Within the Plant

The estimated quantities of untreated airborne radioactive material produced in the buildings of the plant are given in Table 12.2-30. These releases were based upon NUREG-0016 Revision 1, "Calculation of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors." The quantities in Table 12.2-30 were generated as follows:

- All turbine building releases were originally reduced by a factor of five to take credit for the leakage collection system installed for valves in lines 2 1/2" and larger.

Valves in the turbine building were originally provided with valve stem packing leakoff connections. Research and testing has shown that improved packing provides an effective seal to prevent leakage into the Turbine Building. As a result, these leakoff connections are in the process of being removed and packing configurations changed, as appropriate, to conform with the new requirements. As part of this effort, leakoff isolation valves and piping will be removed (or abandoned in place) and the leakoff collection header piping will be removed or abandoned in place.

(NOTE: Releases assigned to the turbine building are assumed to include any control structure

releases.)

- The reactor building releases were taken to be the sum of releases listed in NUREG-0016 Revision 1 for the auxiliary building and containment building.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-11 - The radwaste building releases are "per reactor" and consequently were doubled for Susquehanna SES.

12.2.2.7 Distribution of Airborne Releases Within the Plant

The approach taken to determine the anticipated distribution of gaseous effluents assumed that all untreated airborne radioactive material originates only within the equipment areas of the plant. It was further assumed that a major percentage of the release is generated within a few specific areas of each building. These ar eas are identified as per NUREG-0016 Revision 1, Section 2.2.4, "Gaseous Releases From Building Ventilation Systems", as follows:

For the Reactor Building, 90% of the normal power operation releases are due to the reactor water cleanup (RWCU) pumps and filter/demineralizers (F/Ds), and the emergency core cooling systems (ECCS) and 10% of the releases are due to the fuel handling and drywell areas.

During shutdown, 10% of the releases are from the RWCU and ECCS Systems and 90% of the releases are from the fuel handling and drywell areas.

For the Turbine Building, 85% of the normal power operation releases are from the main condenser area and 15% of the releases are from miscellaneous areas such as the steam jet air ejector (SJAE) room, the turbine operating floor, the feedwater pump room, and the mechanical vacuum pump room. During shutdown, 50% of the releases are from the main condenser area and 50% of the releases are from miscellaneous areas.

For the Radwaste Building, during normal power operation and shutdown, 10% of the releases are from the solid waste handling areas and 90% of the releases are from the liquid waste handling areas.

Releases were assumed to be generated continuously throughout the year.

The selection and relative contributions of the major areas was determined from NUREG-0016 Revision 1, based on EPRI NP-495 and NRC studies. These studies provide data on the

important sources of Iodine-131 at operating BWR's and determined the relative release rate from each source. The relative release rates for all airborne radionuclides were then assumed to be directly proportional to the Iodine-131 release rates.

Table 12.2-31 lists the major airborne contributors in each building and the percentage of the total building release assigned to each. Tables 12.2-32 through 12.2-34 provide the specific equipment areas of the plant associated with the major contributors and the applicable exhaust SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-12 air flow rates. Note that only those equipment areas which have a significant potential for airborne radioactive material releases were included in the "other equipment areas" category.

12.2.2.8 Estimated Airborne Radioactive Material Concentrations Within the Plant

The airborne radionuclide concentrations for each equipment area were calculated using the following methodology. For a specific area, the appropriate building release (Table 12.2-30) was multiplied by the applicable release percentage for the area (Table 12.2-31) and divided by the area annual exhaust flow (Table 12.2-32, 12.2-33, or 12.2-34). The resultant concentrations are presented in Tables 12.2-35 through 12.2-37 and are compared to 10CFR20 Appendix B

requirements.

12.2.2.9 Changes to Source Data Since PSAR Source data has been updated to reflect changes in contained sources due to the implementation of power uprate, 24-month fuel cycle, design and operational changes, hydrogen water chemistry, condensate filtration, and NUREG-0016, Revision 1.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-9 THIS TABLE HAS BEEN INTENTIONALLY LEFT BLANK

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-11 STEAM N-16 SHIELDING SOURCE TERMS FOR TURBINE AND REACTOR BUILDING EQUIPMENT (1)(2) Equipment Decay Time (2) (Seconds)

N-16 (1) µCi/gm Steam Density gm/cc Reactor Pressure Vessel Nozzle 0 250 3.51E-02 (3) 24" Main Steam Line 0 250 3.51E-02 Main Steam Stop Valves 1.87 227.3 3.51E-02 High Pressure Turbine Inlet 2.30 218.0 3.51E-02 High Pressure Turbine Outlet 2.41 215.6 8.67E-03 Moisture Separator Inlet 2.41 215.6 8.67E-03 Moisture Separator Outlet 3.11 201.4 7.22E-03 Cross Around Piping 3.11 201.4 7.22E-03 Combined Intermediate Valve 3.11 201.4 7.22E-03 Low Pressure Turbine Inlet 3.11 201.4 7.22E-03 Low Pressure Turbine Outlet 3.36 196.6 9.28E-05 Steam to Reactor Feedwater Turbine 4" Turbine 10" 0.37 5.24 250 163.7 3.51E-02 7.30E-03 Reactor Feedwater Turbine Exhaust 5.24 163.7 1.07E-04 Feedwater Heaters

E-103 E-104 E-105 3.04 3.15 2.42 202.9 200.6 215.4 2.75E-03 4.71E-03 1.11E-02 Steam Seal Evaporator (Inlet Steam) 5.51 159.5 2.58E-03 RCIC Turbine (Inlet Steam) 1.0 247.2 3.51E-02 HPCI Turbine (Inlet Steam) 1.0 247.2 3.51E-02 Notes: 1. Values presented are based on plant operation with hydrogen water chemistry.

2. Decay time is based on a conservative estimate of the transit time at a Main Steam flow rate of 1.65E+07 lbs/hr.

3. 3.51E-02 = 3.51 X 10

-2

SSES-FSAR Table 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-15 CONDENSATE SHIELDING SOURCE TERMS(1)(2)(4) Isotope µCi/cc Isotope µCi/cc Isotope µCi/cc Br-83 4.75E-04(3) Ba-139 1.72E-04 Kr-83m 3.74E-06 Br-84 1.14E-03 Ba-140 9.34E-06 La-144 1.44E-06 Br-85 7.35E-04 Ba-141 3.32E-04 Cs-141 2.72E-04 I-131 3.38E-04 Ba-142 2.14E-04 Cs-140 2.17E-03 I-132 3.81E-03 Np-289 2.40E-04 Cs-139 8.01E-04 I-133 2.34E-03 Na-24 2.00E-06 Rb-90m 1.76E-04 I-134 9.34E-03 Cr-51 5.00E-07 Rb-88 3.30E-05 I-135 3.63E-03 Mn-56 4.96E-05 Sr-94 9.91E-06 Sr-89 3.10E-06 Co-58 5.00E-06 Y-94 1.31E-06 Sr-90 2.30E-07 Co-60 5.00E-07 Sr-95 1.05E-06 Sr-91 8.10E-05 Ni-65 2.97E-07 Y-95 1.20E-06 Sr-92 1.39E-04 W-187 3.00E-06 Sr-93 1.23E-04 Mo-99 2.20E-05 La-142 3.41E-06 Y-93 2.88E-07 Tc-99m 2.79E-04 La-141 1.77E-06 Y-93m 4.80E-05 Tc-101 1.27E-04 Y-92 9.04E-07 Rb-91 2.17E-03 Te-132 4.90E-05 Y-91m 1.20E-06 Rb-90 1.74E-03 Cs-134 1.60E-07 Xe-135 5.60E-06 Rb-89 2.51E-04 Cs-136 1.10E-07 Xe-135m 3.41E-05 N-13 6.09E-03 Cs-137 2.40E-07 Xe-133 3.20E-07 N-16 4.29E-04 Cs-138 2.68E-04 Kr-85m 1.38E-06 Notes:

(1) Basis for condensate isotopic sources:

a. Daughter products formed from deca y of noble gases in transit through the turbine and main condenser.

b. Coolant and noncoolant particulate activation and fission product carryover fraction of 1X10

-3 per unit weight into steam. c. Reaction water fission product halogen carryover fraction of 8X10

-2 per unit weight into steam.

(2) Isotopes with activity concentrations less than 10

-7 µCi/cc are not shown.

(3) 4.75E-04 = 4.75 X 10

-4 (4) Values shown are conservatively based on 2 minutes holdup. Minimum retention time in the Hotwell is estimated at 3 minutes.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-17 OFFGAS LINE FROM SJAE TO RECOMBINER SHIELDING SOURCE TERMS (1)(3) Isotope Curies Isotope Curies Isotope Curies Xe-133 3.49E-03(2) Ba-139 1.14E-02 Sr-91 1.00E-02 Xe-135m 1.13E-02 Cs-138 3.91E-03 Y-91 1.00E-02 Xe-135 9.10E-03 Cs-137 3.90E-03 Y-91m 5.93E-03 Xe-137 6.26E-02 Kr-83m 1.44E-03 Rb-90 1.12E-02 Xe-138 3.78E-02 Kr-85m 2.58E-03 Sr-90 7.01E-03 Xe-139 1.10E-01 Kr-87 8.33E-03 Y-90 7.01E-03 Xe-140 1.00E-01 Kr-88 8.34E-03 Rb-89 5.62E-03 Xe-141 1.94E-02 Kr-89 5.41E-02 Sr-89 5.60E-03 Xe-142 4.82E-03 Kr-90 1.07E-01 N-13 2.11E-02 Cs-142 1.00E-03 Kr-91 9.70E-02 N-16 7.79E+01 Cs-141 2.23E-03 Kr-92 4.79E-02 N-17 2.85E-02 Ba-141 2.02E-03 Kr-93 9.39E-03 O-19 5.27E-01 La-141 2.01E-03 Kr-94 1.39E-03 F-18 3.03E-03 Ce-141 2.01E-03 Rb-93 1.41E-03 Cs-140 1.08E-02 Rb-92 7.30E-03 Ba-140 1.04E-02 Sr-92 4.98E-03 La-140 1.04E-02 Y-92 4.98E-03 Cs-139 1.14E-02 Rb-91 1.04E-02

Notes:

(1) Bases: a. 100,000

µCi/sec noble gases (after 30 min. decay). b. Decay time from Reactor Nozzl e to Preheater Inlet, 5.23 seconds. c. Main Steam flow 1.65E+07 lbs/ hr. d. All particulates generated plateout on pipe surfaces. e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation. (2) 3.49E-03 = 3.49 X 10

-3 (3) Isotopes with activity less than 10

-3 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-18 OFF-GAS RECOMBINER SH IELDING SOURCE TERMS(1)(2) Isotope Curies Isotope Curies Isotope Curies Xe-133 3.44E-03(3) Cs-139 2.30E-01 Rb-92 8.28E-02 Xe-135m 1.12E-02 Ba-139 2.30E-01 Sr-92 8.59E-02 Xe-135 8.97E-03 Cs-138 7.98E-02 Y-92 8.59E-02 Xe-137 6.15E-02 Cs-137 7.96E-02 Rb-91 1.94E-01 Xe-138 3.72E-02 Kr-83m 1.42E-03 Sr-91 1.95E-01 Xe-139 1.07E-01 Kr-85m 2.54E-03 Y-91 1.95E-01 Xe-140 9.48E-02 Kr-87 8.21E-03 Y-91m 1.15E-01 Xe-141 1.32E-02 Kr-88 8.22E-03 Rb-90 2.22E-01 Xe-142 3.14E-03 Kr-89 5.32E-02 Sr-90 1.41E-01 Cs-142 6.85E-03 Kr-90 1.03E-01 Y-90 1.41E-01 Ba-142 7.46E-03 Kr-91 8.93E-02 Rb-89 1.14E-01 La-142 7.46E-03 Kr-92 3.81E-02 Sr-89 1.14E-01 Cs-141 3.07E-02 Kr-93 6.74E-03 Rb-88 1.76E-02 Ba-141 3. 09E-02 Rb-94 1.97E-03 N-13 2.08E-02 La-141 3.09E-02 Sr-94 2.11E-03 N-16 7.03E+01 Ce-141 3.09E-02 Y-94 2.11E-03 N-17 2.39E-02 Ba-140 2.05E-01 Rb-93 1.51E-02 O-19 5.07E-01 Cs-140 2.05E-01 Sr-93 1.56E-02 F-18 2.99E-03 La-140 2.05E-01 Y-93 1.56E-02 NOTES: (1) Bases a) 100,000

µCi/sec noble gases (after 30 minute decay). b) Decay time from Reactor Nozzle to entrance of recombiner, 6.22 seconds. c) Particulate daughter products accumu late for 40 years in Recombiner. d) Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell. e) Main steam flow 1.65E+07 lbs/hr. f) Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation. (2) Isotopes with activity less then 10

-3 Ci are not shown. (3) 3.44E-03 = 3.44 X 10

-3 SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-19 OFF-GAS LINE FROM RECOMBINER TO CHARCOAL SYSTEM SHIELDING SOURCE TERMS(1),(2) Isotope Curies Isotope Curies Isotope Curies Xe-133m 1.32E-02(3) Cs-138 4.48E-01 Y-91 4.45E-02 Xe-133 3.79E-01 Cs-137 3.24E-01 Y-91m 2.65E-02Xe-135m 1.16E+00 Kr-83m 1.55E-01 Rb-90 7.20E-01Xe-135 9.90E-01 Kr-85m 2.79E-01 Sr-90 1.90E-01Xe-137 5.34E+00 Kr-87 8.95E-01 Y-90 1.90E-01Xe-138 3.88E+00 Kr-88 9.01E-01 Rb-89 5.75E-01Xe-139 3.96E+00 Kr-89 4.39E+00 Sr-89 4.42E-01Xe-140 1.07E+00 Kr-90 2.9 E+00 Rb-88 1.15E-01Cs-140 4.27E-01 Kr-91 4.31E-01 N-13 2.08E+00Ba-140 1.09E-01 Kr-92 1.20E-02 N-16 1.80E+02La-140 1.09E-01 Rb-92 1.07E-02 N-17 1.76E-02 Cs-139 5.93E-01 Rb-91 1.67E-01 F-18 3.27E-01 Ba-139 4.02E-01 Sr-91 4.46E-02 NOTES:

(1) Bases a) 100,000

µCi/sec noble gases (after 30 minute decay). b) Main steam flow 1.65E+07 lbs/hr. c) Decay time from Reactor Nozzle to offgas line entrance, 45.5 seconds. d) Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation. e) Nitrogen partition factors: 80 percent for non-condensible gases; 20 percent to main condenser hotwell. (2) Isotopes with activity less then 10

-2 Ci are not shown. (3) 1.32E-02 = 1.32 X 10

-2

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-21 OFFGAS AMBIENT CHARCOAL FREON CHILLER SHIELDING SOURCE TERMS (1) (3)

Isotope Curies Isotope Curies Kr-87 1.62E-01(2) Rb-89 3.81E-01 Kr-88 1.66E-01 Sr-89 3.81E-01 Kr-89 4.07E-01 Rb-88 1.54E-01 Xe-135m 1.87E-01 N-13 3.11E-01 Xe-135 1.84E-01 Xe-137 5.64E-01 Xe-138 6.37E-01 Cs-138 5.94E-01 Cs-137 3.18E-01

NOTES:

(1) Bases: a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to inlet of Freon Chiller, 472 seconds. c. Particulate daughter products accumulate for 40 years in equipment. d. Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell. e. Main Steam flow 1.65E+07 lbs /hr. f. Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation.

(2) 1.62E-01 = 1.62 X 10

-1 (3) Isotopes with activities less than10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-22 OFFGAS CHARCOAL GUARD BED SHIELDING SOURCE TERMS (1) (3) Isotope Curies Isotope Curies Kr-87 1.13E-01(2) Rb-89 3.64E-01 Kr-88 1.15E-01 Sr-89 3.64E-01 Kr-89 2.77E-01 Rb-88 1.51E-01 Xe-135m 1.30E-01 N-13 2.15E-01 Xe-135 1.28E-01 Xe-137 3.85E-01 Xe-138 4.42E-01 Cs-138 5.78E-01 Cs-137 3.05E-01

NOTES:

(1) Bases: a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to Charcoal Guard Bed, 486 seconds. c. Particulate daughter products accumulate for 40 years in equipment. d. Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell.

e. Main Steam flow 1.65E+07 lbs/hr. f. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 1.13E-01 = 1.13 X 10

-1.

(3) Isotopes with activity less than 10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-23 OFFGAS CHARCOAL BEDS SHIELDING SOURCE TERMS (1)(3)(4)(5)

Isotope Curies Isotope Curies F-18dd 3.87E+00 Kr-83m 3.10E+01 Xe-131m 4.79E+00 Kr-85m 1.29E+02 Xe-133m 4.75E+01(2) Kr-85 8.05E+00 Xe-133 2.08E+03 Kr-87 1.21E+02 Xe-135m 2.89E+01 Kr-88 2.70E+02 Xe-135 9.36E+02 Kr-89 1.24E+01 Xe-137 2.13E+01 Rb-89 1.26E+01 Xe-138 1.11E+02 Sr-89 1.25E+01 Xe-139 1.82E-01 Rb-88 2.70E+02 Cs-139 1.97E-01 N-13 1.25E+01 Ba-139 1.97E-01 Cs-137 1.29E+01 Cs-138 1.11E+02 NOTES:

(1) Bases:

a. 100,000

µCi/sec noble gas release rate after 30 min. decay.

b. Decay time from Reactor Nozzle to inlet of Main Charcoal Beds (1 st bed), 503 seconds.

c. Particulate daughter products accumulate for 40 years in equipment.

d. Main Steam flow 1.65E+07 lbs/hr.

e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 4.75E+01 = 4.75 X 10

+1.

(3) Activity per charcoal bed (1 only).

(4) Representative of each of the five charcoal beds.

(5) Isotopes with activity less than 10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-24 OFFGAS POST HEPA FILTER SHIELDING SOURCE TERMS (1)(3) Isotope Curies Xe-131m 1.25E-05(2) Xe-133 2.12E-03 Kr-83m 2.08E-05 Kr-85m 5.21E-04 Kr-85 5.62E-05 Kr-87 2.82E-05 Kr-88 4.72E-04 Rb-88 3.06E-03 N-13 3.01E-03 F-18 1.29E-02

NOTES:

(1) Bases:

a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to outlet of ambient Charcoal Absorbers: Kryptons - 1.16E+05 seconds Xenons - 2.05E+06 seconds

c. Particulate daughter products accumulate for 40 years in equipment.

d. Main Steam flow 1.65E+07 lbs /hr. e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 1.25E-05 = 1.25 X 10

-5.

(3) Isotopes with activity less than 10

-5 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 2 of 2 TABLE 12.2-30 ESTIMATED AIRBORNE RADIOACTIVE RELEASES PRIOR TO TREATMENT (CURIES /YEAR)

(1) Notes:

1. Based on NUREG-0016, Revision 1, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water

Reactors", (BWR - GALE Code), Section 2.2.4.

2. Turbine Building releases reduced by a factor of 5 due to installation of the Process Valve Stem Leakoff Collection System.

3. Reactor Building is the sum of th e Containment & Auxiliary Building.

4. 4.00E+00 = 4.00 X 10

0.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-32 TURBINE BUILDING AIRBORNE SOURCE DESCRIPTIONS AREA DESIGNATION ROOM NUMBERS AND AREAS INCLUDED TOTAL EXHAUST FLOW (CFM) TOTAL ANNUAL (1) EXHAUST (cc/yr) Condenser 30, 31, 32, 34, 37, 38, 39, 113, 211 23,100 3.44E+14(2) Miscellaneous Areas C-10, C-12, C-130, C212 C220, C301, C-400, C-900, C-912, 33, 35, 42, 43, 45, 53, 59, 111, 112, 114,115, 116, 117, 118, 119, 120, 121, 122, 123, 124, 125, 126, 128, 130, 210, 212, 214, 215, 216, 217, 300, 415, 416, 417, 419, 421 183,900 2.74E+15 Notes:

1. Based on continuous release of 365 days per year.

2. 3.44E+14 = 3.44 x 10 14

SSES-FSARNIMSRev.60FSARRev.65Page1of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable1646'Offgaslinesfromrecombinertoradwaste8"GBC-1068"diam.cyl.1.17x10

-30.32212.2-19 30 31 656'656'Condensatepumpsandlines1P-102A,B,C,D48",30"and18"diam.cyl.1.00.37512.2-15 32 34 656'656'RecombinerRecombiner1S-1255'diam.x6'-2"ht.cyl.1.21x10-30.7512.2-1833656'SteamPackingexhauster1E-11036"diam.cyl.30"diam.cyl.1.00.37511.1-1,-2,-3,-455656'Vacuumpump1P-10512"diam.cyl.1.4x10

-30.40611.1-1,-2,-3,-456656'RecombinerpipevalvemanifoldfromSJAE8"diam.and10"diam.cyl.1.21x10-30.32212.2-17C-11656'ChemdraintankandpumpsOT-114OP-132A,B3'-6"diam.6'-4"ht.cyl.1.00.18711.2-642656'CondensatedemineralizerpipewayVariousresinlines30",24",14",3"diam.cyl.1.00.21612.2-1612.2-15152656'Regenerationwastesurgetanks1T-106A,B7'diam.13'highcyl.1.00.512.2-16151656'Neutralizingpumps1P130A,B2"diam.cyl.1.00.15412.2-1643656'Chemicalwaste1T130A,B12'diam.1.00.19712.2-16Neutralizingtanks20'length37656'RFPturbine27"x75"RFT3'diam.cyl.2.04x10

-40.2512.2-1138656'exhaustductduct11.1-1,-2,-3,-439656' SSES-FSARNIMSRev.60FSARRev.65Page2of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable36656'MaincondenserandsteampipingIE-108A,B,C,RFWductcondensatelines3'diam.36"diam.cyl.2.04x10-41.00.2511.1-1,-2,-3,-412.2-11 12.2-1559656'CondensateFiltrationBackwashReceivingTank1T-1879'diam.x24'1cyl1.00.2512.2-48117676'Deepbed1F-106G11'diam.cyl.1.01.512.2-6118676'Condensatedemin.1F-106F3'deep1.01.512.2-16119676'1F-106E1.01.512.2-16120676'1F-106D1.01.512.2-16121676'1F-106C1.01.512.2-16123676'1F-106B1.01.512.2-16124676'1F-106A1.01.512.2-16125676'1F-106H1.01.512.2-16114676'RFPturbine105C4"diam.cyl.0.0330.33712.2-11115676'RFPturbine105B10"diam.cyl.0.00610.36511.1-1,-2,-3,-4116676'RFPturbine105A111676'SJAE1E-1098"diam.cyl.1.21x10

-30.32212.2-17112676'FutureSJAE3'-6"diam.x12'hcyl.1.21x10-31.08113676'Maincondensersteampiping1E-108A,B,C42",28",26",16"diam.cycl..00623,.036.00216,0.6250.37511.1-1,-2,-3,-4,12.2-11.00786125676'Anionandcation1T-1574'6"øx14'hcyl.1.00.2512.2-16regeneration1T-158,1596'6"x14'hcyl.

SSES-FSARNIMSRev.60FSARRev.65Page3of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable110676'Turbinebldg.1C-132A,B3/8"SStubingVaries-11.1-1,-2,-3,-4samplestation130676'CondensateFilterBundles1F135A,B,C,D,E,F6'diam.x4'-2"hcyl.0.9751.58412.2-46133(1)676'CondensateFilterBundles2F125G6'diam.x4'-2"hcyl.0.9751.58412.2-46 130,133(1),C-100676'BackwashPiping12"HCD-107(130)12"HCD-207(133,C-100)12'diam.cyl.1.00.2512.2-47211699'Maincondenserandsteampiping1E108A,B,C,24"DBB-101102,203,10424"diam.cyl.28"diam.cyl.

42"diam.cyl..036.0328 0.006740.940.375 0.62511.1-1,-2,-3,-412.2-11212699'Feedwaterheaters1E1036'øx45'cyl.0.7149/1611.1-1,-2,-3,-4214699'Feedwaterheaters1E1046'øx40'cyl.0.8159/1612.2-11215699'Feedwaterheaters1E1056'øx36'cyl.0.8539/16300714'Mainsteamtunnel24"-DBB-101,24"diam.cyl..0360.9711.1-1,-2,-3,-4102,103,10412.2-11411729'Mainsteamtunnel24"-DBB-101,24"diam.cyl..0360.9711.1-1,-2,-3,-4102,103,10412.2-11416729'Steamseal1E-12851/2'øx34'cyl.0.4475/811.1-1,-2,-3,-4evaporator12-EBB-11312"diam.cyl.2.25-30.68712.2-11419729'Moistureseparator1T-104A,B10'-8"x67.3'cyl.0.7680.7511.1-1,-2,-3,-4421729'and42"cross-oversteamlines42"-GESupplied42"diam.cyl.0.00630.62512.2-11 SSES-FSARNIMSRev.60FSARRev.65Page4of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable420729'HP&LPturbines1G-10128"diam.cyl..03281.411.1-1,-2,-3,-428",42"GESupplied42"diam.cyl..006230.62512.2-11531762'H&VSystem1F-156,1572'x2'x1'(63units)0.490.25N/A532762'(HEPA&charcoal)1F-158315ft 3(carbon)Note:1.Room133onlyappliestoUnit2.

SSES-FSAR Table Rev. 54 FSAR Rev. 65 Page 1 of 1 Table 12.2-45 This Table Has Been Deleted

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 1 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr H 3 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.57E+04 7.55E+04 7.47E+04 7.37E+04 7.17E+04 6.40E+04 Na 24 2.73E+02 2.73E+02 2.66E+02 2.60E+02 1.86E+02 8.79E+01 2.91E+00 4.46E-13 0.00E+00 Cr 51 5.65E+06 5.65E+06 5.65E+06 5.65E+06 5.61E+06 5.52E+06 5.11E+06 2.67E+06 5.95E+05 6.26E+04 6.08E+02 7.04E-06 Mn 54 4.27E+05 4.27E+05 4.27E+05 4.27E+05 4.27E+05 4.26E+05 4.23E+05 4.00E+05 3.50E+05 2.87E+05 1.90E+05 3.75E+04 Fe 55 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.86E+06 1.79E+06 1.68E+06 1.48E+06 8.86E+05 Mn 56 1.21E+07 1.21E+07 1.05E+07 9.17E+06 1.40E+06 1.90E+04 7.47E-05 0.00E+00 Co 58 5.91E+05 5.91E+05 5.91E+05 5.91E+05 5.90E+05 5.86E+05 5.68E+05 4.41E+05 2.45E+05 1.02E+05 1.67E+04 1.32E+01 Fe 59 1.28E+05 1.28E+05 1.28E+05 1.28E+05 1.28E+05 1.27E+05 1.21E+05 3.02E+04 3.16E+04 7.79E+03 4.35E+02 4.97E+03 Ni 59 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 Co 60 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.16E+05 3.09E+05 2.99E+05 2.80E+05 2.35E+05 Ni 63 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.54E+04 3.54E+04 3.53E+04 3.48E+04 Cu 64 2.65E+02 2.65E+02 2.58E+02 2.51E+02 1.72E+02 7.17E+01 1.41E+00 2.28E-15 0.00E+00 Ni 65 5.57E+04 5.57E+04 4.86E+04 4.23E+04 6.17E+03 7.57E+01 1.89E-07 0.00E+00 Ga 73 4.97E+03 4.97E+03 4.64E+03 4.32E+03 1.59E+03 1.63E+02 5.65E-03 0.00E+00 Ge 73m 4.91E+03 4.91E+03 4.58E+03 4.26E+03 1.57E+03 1.60E+02 5.58E-03 0.00E+00 As 76 2.56E+03 2.56E+03 2.53E+03 2.50E+03 2.08E+03 1.36E+03 2.05E+02 1.49E-05 5.05E-22 0.00E+00 Ge 77 7.31E+04 7.31E+04 7.10E+04 6.88E+04 4.48E+04 1.68E+04 2.02E+02 4.83E-15 0.00E+00 As 77 2.26E+05 2.26E+05 2.25E+05 2.23E+05 2.04E+05 1.60E+05 4.61E+04 6.72E-01 4.61E-12 8.71E-29 0.00E+00 Ge 78 7.22E+05 7.22E+05 5.70E+05 4.50E+05 1.65E+04 8.56E+00 1.43E-14 0.00E+00 As 78 7.32E+05 7.32E+05 7.13E+05 6.70E+05 8.25E+04 1.25E+02 1.39E-12 0.00E+00 Br 82 3.81E+05 3.81E+05 3.78E+05 3.75E+05 3.27E+05 2.38E+05 5.81E+04 2.77E-01 1.46E-13 0.00E+00 Br 83 1.29E+07 1.29E+07 1.18E+07 1.04E+07 1.40E+06 1.38E+04 1.28E-05 0.00E+00 Kr 83m 1.31E+07 1.31E+07 1.29E+07 1.26E+07 3.64E+06 5.27E+04 5.39E-05 0.00E+00 Kr 85 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.47E+06 1.46E+06 1.44E+06 1.39E+06 1.22E+06 Kr 85m 2.68E+07 2.68E+07 2.51E+07 2.32E+07 7.87E+06 6.62E+05 9.63E+00 0.00E+00 Rb 86 2.17E+05 2.17E+05 2.16E+05 2.16E+05 2.14E+05 2.09E+05 1.86E+05 7.10E+04 7.62E+03 2.68E+02 2.73E-01 4.31E-13 Kr 87 5.37E+07 5.37E+07 4.13E+07 3.15E+07 6.94E+05 1.13E+02 1.03E-15 0.00E+00 Kr 88 7.45E+07 7.45E+07 6.60E+07 5.84E+07 1.05E+07 2.12E+05 4.93E-03 0.00E+00 Rb 88 7.64E+07 7.64E+07 7.16E+07 6.46E+07 1.18E+07 2.38E+05 5.51E-03 0.00E+00 Zr 89 5.52E+04 5.52E+04 5.50E+04 5.48E+04 5.15E+04 4.47E+04 2.37E+04 9.55E+01 2.84E-04 1.46E-12 1.74E-29 0.00E+00 Sr 89 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.02E+08 9.78E+07 6.85E+07 3.01E+07 8.79E+06 6.91E+05 3.09E+01

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 2 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Y 89m 1.68E+05 1.63E+05 6.43E+04 6.40E+04 6.07E+04 5.39E+04 3.25E+04 6.46E+03 2.80E+03 8.18E+02 6.43E+01 2.87E-03 Sr 90 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Y 90 1.36E+07 1.36E+07 1.36E+07 1.36E+07 1.35E+07 1.34E+07 1.32E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Sr 91 1.31E+08 1.31E+08 1.26E+08 1.21E+08 7.30E+07 2.28E+07 1.21E+05 2.23E-15 0.00E+00 Y 91 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.29E+08 9.48E+07 4.65E+07 1.60E+07 1.78E+06 3.11E+02 Y 91m 7.57E+07 7.57E+07 7.53E+07 7.40E+07 4.64E+07 1.44E+07 7.64E+04 1.41E-15 0.00E+00 Sr 92 1.39E+08 1.39E+08 1.22E+08 1.08E+08 1.80E+07 2.99E+05 3.01E-03 0.00E+00 Y 92 1.40E+08 1.40E+08 1.39E+08 1.37E+08 6.53E+07 4.42E+06 4.04E+00 0.00E+00 Zr 93 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 Y 93 1.07E+08 1.07E+08 1.04E+08 1.01E+08 6.23E+07 2.08E+07 1.48E+05 3.74E-14 0.00E+00 Zr 95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.91E+08 1.89E+08 1.84E+08 1.39E+08 7.23E+07 2.73E+07 3.67E+06 1.35E+03 Nb 95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.79E+08 1.21E+08 5.36E+07 7.92E+06 2.98E+03 Nb 95m 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.12E+06 2.10E+06 1.63E+06 8.50E+05 3.21E+05 4.32E+04 1.59E+01 Nb 96 3.12E+05 3.12E+05 3.07E+05 3.02E+05 2.45E+05 1.53E+05 1.80E+04 1.63E-04 4.44E-23 0.00E+00 Zr 97 1.90E+08 1.90E+08 1.86E+08 1.82E+08 1.36E+08 7.09E+07 3.70E+06 2.84E-05 0.00E+00 Nb 97 1.91E+08 1.91E+08 1.90E+08 1.88E+08 1.46E+08 7.13E+07 3.72E+06 3.06E-05 0.00E+00 Nb 97m 1.80E+08 1.80E+08 1.76E+08 1.73E+08 1.30E+08 6.73E+07 3.51E+06 2.70E-05 0.00E+00 Mo 99 2.02E+08 2.02E+08 2.01E+08 2.00E+08 1.86E+08 1.57E+08 7.37E+07 1.05E+05 2.79E-02 3.84E-12 0.00E+00 Tc 99 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.22E+03 2.22E+03 2.22E+03 2.22E+03 2.22E+03 Tc 99m 1.79E+08 1.79E+08 1.79E+08 1.79E+08 1.73E+08 1.51E+08 7.14E+07 1.01E+05 2.70E-02 3.72E-12 0.00E+00 Ru103 1.72E+08 1.72E+08 1.72E+08 1.71E+08 1.70E+08 1.69E+08 1.60E+08 1.01E+08 3.50E+07 7.15E+06 2.71E+05 6.80E-01 Rh103m 1.71E+08 1.71E+08 1.71E+08 1.71E+08 1.70E+08 1.68E+08 1.60E+08 1.01E+08 3.50E+07 7.14E+06 2.71E+05 6.79E-01 Ru105 1.19E+08 1.19E+08 1.13E+08 1.05E+08 3.51E+07 2.89E+06 3.78E+01 0.00E+00 Rh105 1.11E+08 1.11E+08 1.12E+08 1.12E+08 1.05E+08 8.02E+07 1.96E+07 9.55E+01 5.26E-11 1.74E-29 0.00E+00 Rh105m 3.38E+07 3.38E+07 3.22E+07 2.99E+07 1.00E+07 8.25E+05 1.08E+01 0.00E+00 Ru106 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rh106 7.38E+07 7.37E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rh106m 2.39E+06 2.39E+06 2.04E+06 1.74E+06 1.85E+05 1.11E+j03 1.10E-07 0.00E+00 Ag109m 4.14E+07 4.14E+07 4.04E+07 3.94E+07 2.77E+07 1.23E+07 3.22E+05 7.79E-02 7.12E-02 6.22E-02 4.71E-02 1.57E-02 Ag110 1.87E+07 1.82E+07 6.23E+03 6.23E+03 6.23E+03 6.22E+03 6.17E+03 5.74E+03 4.86E+03 3.78E+03 2.26E+03 2.98E+02 Ag110m 4.58E+05 4.58E+05 4.58E+05 4.58E+05 4.58E+05 4.57E+05 4.53E+05 4.22E+05 3.57E+05 2.78E+05 1.67E+05 2.19E+04

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 3 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Pd111 6.66E+06 6.66E+06 2.88E+06 1.29E+06 8.10E+04 1.08E+04 1.24E+00 0.00E+00 Pd111m 2.83E+05 2.83E+05 2.65E+05 2.49E+05 1.03E+05 1.38E+04 1.57E+00 0.00E+00 Ag111 6.72E+06 6.72E+06 6.71E+06 6.70E+06 6.52E+06 6.13E+06 4.65E+06 4.13E+05 1.56E+03 3.59E-01 1.18E-08 0.00E+00 Ag111m 6.71E+06 6.71E+06 3.04E+06 1.39E+06 1.01E+05 1.34E+04 1.54E+00 0.00E+00 Pd112 3.02E+06 3.02E+06 2.97E+06 2.92E+06 2.32E+06 1.37E+06 1.28E+05 1.51E-04 3.82E-25 0.00E+00 Ag112 3.03E+06 3.03E+06 3.03E+06 3.02E+06 2.64E+06 1.60E+06 1.51E+05 1.78E-04 4.49E-25 0.00E+00 In113m 3.47E+05 3.47E+05 3.47E+05 3.47E+05 3.46E+05 3.45E+05 3.39E+05 2.90E+05 2.02E+05 1.18E+05 3.85E+04 4.73E+02 Sn113 3.47E+05 3.47E+05 3.47E+05 3.47E+05 3.46E+05 3.45E+05 3.38E+05 2.90E+05 2.02E+05 1.18E+05 3.84E+04 4.72E+02 Ag113 1.69E+06 1.69E+06 1.60E+06 1.50E+06 6.05E+05 7.64E+04 7.05E+00 0.00E+00 Cd113m 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.32E+03 4.29E+03 4.23E+03 4.13E+03 3.74E+03 In114 5.63E+04 5.62E+04 3.54E+04 3.54E+04 3.53E+04 3.50E+04 3.35E+04 2.33E+04 1.01E+04 2.86E+03 2.13E+02 7.72E-03 In114m 3.71E+04 3.71E+04 3.71E+04 3.71E+04 3.69E+04 3.65E+04 3.51E+04 2.44E+04 1.05E+04 2.98E+03 2.23E+02 8.10E-03 Cd115 9.02E+05 9.02E+05 8.94E+05 8.94E+05 8.17E+05 6.62E+05 2.61E+05 8.02E+01 6.26E-07 4.32E-19 0.00E+00 Cd115m 4.20E+04 4.20E+04 4.19E+04 4.19E+04 4.18E+04 4.13E+04 3.95E+04 2.64E+04 1.04E+04 2.56E+03 1.44E+02 1.68E-03 In115m 9.02E+05 9.02E+05 9.02E+05 9.02E+05 8.63E+05 7.21E+05 2.84E+05 9.02E+01 1.15E+00 2.83E-01 1.59E-02 1.86E-07 Cd117 8.86E+05 8.86E+05 7.79E+05 6.76E+05 9.63E+04 1.12E+03 2.22E-06 0.00E+00 Cd117m 2.03E+05 2.03E+05 1.83E+05 1.66E+05 3.90E+04 1.44E+03 5.13E-04 0.00E+00 In117 6.62E+05 6.62E+05 6.56E+05 6.38E+05 2.02E+05 4.12E+03 6.63E-04 0.00E+00 In117m 8.10E+05 8.10E+05 8.02E+05 7.79E+05 2.34E+05 4.13E+03 2.74E-05 0.00E+00 Sn117m 2.49E+06 2.49E+06 2.49E+06 2.48E+06 2.44E+06 2.36E+06 2.03E+06 5.39E+05 2.53E+04 2.58E+02 2.04E-02 1.38E-18 Sn119m 2.46E+06 2.46E+06 2.46E+06 2.46E+06 2.46E+06 2.45E+06 2.43E+06 2.29E+06 1.99E+06 1.60E+06 1.04E+06 1.84E+05 Sn121 1.54E+06 1.54E+06 1.52E+06 1.51E+06 1.26E+06 8.35E+05 1.32E+05 4.17E+02 4.16E+02 4.16E+02 4.12E+02 4.03E+02 Sn121m 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.91E+02 1.91E+02 1.89E+02 1.85E+02 Sb122 2.13E+05 2.13E+05 2.12E+05 2.12E+05 1.96E+05 1.65E+05 7.65E+04 9.66E+01 1.98E-05 1.83E-15 0.00E+00 Te123m 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.63E+02 9.47E+02 8.17E+02 5.76E+02 3.42E+02 1.17E+02 1.70E+00 Sn123 1.36E+05 1.36E+05 1.36E+05 1.36E+05 1.36E+05 1.35E+05 1.33E+05 1.16E+05 8.40E+04 5.18E+04 1.92E+04 3.80E+02 Sn123m 9.25E+05 9.25E+05 5.53E+05 3.29E+05 2.31E+02 1.42E-05 0.00E+00 Sb124 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.64E+04 9.31E+04 6.89E+04 3.45E+04 1.22E+04 1.45E+03 3.22E-01 Sn125 5.57E+05 5.57E+05 5.56E+05 5.55E+05 5.44E+05 5.18E+05 4.18E+05 6.45E+04 8.61E+02 1.33E+00 2.19E-06 0.00E+00 Sn125m 2.07E+06 2.07E+06 2.35E+05 2.65E+04 1.39E-09 0.00E+00 0.00E+00 Sb125 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.35E+06 1.29E+06 1.21E+06 1.07E+06 6.43E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 4 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Te125m 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.04E+05 3.06E+05 3.04E+05 2.93E+05 2.61E+05 1.57E+05 Sb126 4.64E+04 4.64E+04 4.64E+04 4.64E+04 4.56E+04 4.39E+04 3.71E+04 8.68E+03 3.15E+02 1.36E+01 1.16E+01 1.16E+01 Sb126m 5.45E+04 5.45E+04 1.83E+04 6.18E+03 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 Sn127 3.78E+06 3.78E+06 3.21E+06 2.72E+06 2.70E+05 1.38E+03 6.56E-08 0.00E+00 Sb127 9.40E+06 9.40E+06 9.40E+06 9.32E+06 8.94E+06 7.95E+06 4.61E+06 4.28E+04 8.71E-01 7.95E-08 2.60E-22 0.00E+00 Te127 9.32E+06 9.32E+06 9.32E+06 9.32E+06 9.24E+06 8.63E+06 5.78E+06 1.38E+06 9.09E+05 5.13E+05 1.58E+05 1.52E+03 Te127m 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.57E+06 1.36E+06 9.32E+05 5.25E+05 1.61E+05 1.55E+03 Sb128 1.61E+06 1.61E+06 1.57E+06 1.53E+06 9.17E+05 2.67E+05 1.05E+03 1.50E-18 0.00E+00 Sb129 3.47E+07 3.47E+07 3.25E+07 3.00E+07 9.93E+06 8.02E+05 9.47E+00 0.00E+00 0.00E+00 Te129 3.29E+07 3.29E+07 3.26E+07 3.19E+07 1.53E+07 5.10E+06 3.95E+06 2.31E+06 6.71E+05 1.05E+05 2.29E+03 6.54E-04 Te129m 6.66E+06 6.66E+06 6.66E+06 6.66E+06 6.64E+06 6.56E+06 6.17E+06 3.61E+06 1.05E+06 1.63E+05 3.58E+03 1.02E-03 Xe129m 5.16E+03 5.16E+03 5.15E+03 5.14E+03 5.03E+03 4.78E+03 3.77E+03 4.97E+02 4.62E+00 4.15E-03 2.21E-09 0.00E+00 I130 2.64E+06 2.64E+06 2.58E+06 2.51E+06 1.70E+06 6.91E+05 1.22E+04 7.79E-12 0.00E+00 Te131 9.09E+07 9.09E+07 6.98E+07 4.42E+07 4.03E+06 2.79E+06 5.28E+05 2.90E-01 1.02E-15 0.00E+00 Te131m 2.15E+07 2.15E+07 2.12E+07 2.11E+07 1.80E+07 1.24E+07 2.35E+06 1.28E+00 4.56E-15 0.00E+00 I131 1.07E+08 1.07E+08 1.07E+08 1.07E+08 1.05E+08 1.00E+08 7.87E+07 8.40E+06 4.76E+04 2.03E+01 2.35E-06 0.00E+00 Xe131m 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.45E+06 1.38E+06 5.05E+05 2.00E+04 1.11E+02 2.29E-03 7.54E-22 Te132 1.54E+08 1.54E+08 1.54E+08 1.53E+08 1.44E+08 1.25E+08 6.59E+07 2.61E+05 7.46E-01 3.61E-09 2.76E-26 0.00E+00 I132 1.57E+08 1.57E+08 1.57E+08 1.56E+08 1.48E+08 1.28E+08 6.78E+07 2.69E+05 7.72E-01 3.71E-09 2.84E-26 0.00E+00 Cs132 4.35E+03 4.35E+03 4.35E+03 4.33E+03 4.20E+03 3.91E+03 2.83E+03 1.76E+02 2.87E-01 1.89E-05 4.67E-14 0.00E+00 Te133m 9.86E+07 9.86E+07 6.82E+07 4.69E+07 2.44E+05 1.49E+00 0.00E+00 I133 2.22E+08 2.22E+08 2.20E+08 2.18E+08 1.74E+08 1.02E+08 9.24E+06 8.63E-03 1.25E-23 0.00E+00 Xe133 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.06E+08 1.51E+08 4.99E+06 1.80E+03 1.22E-02 2.82E-13 0.00E+00 Xe133m 6.99E+06 6.99E+06 6.99E+06 6.98E+06 6.86E+06 6.29E+06 3.00E+06 8.48E+02 4.78E-06 2.03E-18 0.00E+00 I134 2.45E+08 2.45E+08 2.15E+08 1.75E+08 1.53E+06 5.71E+00 0.00E+00 Cs134 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.29E+07 2.24E+07 2.12E+07 1.95E+07 1.64E+07 8.40E+06 Cs134m 4.81E+06 4.81E+06 4.27E+06 3.79E+06 7.16E+05 1.59E+04 5.69E-04 0.00E+00 I135 2.11E+08 2.11E+08 2.00E+08 1.90E+08 9.09E+07 1.68E+07 8.40E+03 0.00E+00 Xe135 7.03E+07 7.03E+07 7.56E+07 8.02E+07 1.01E+08 5.62E+07 4.00E+05 1.18E-15 0.00E+00 Xe135m 4.63E+07 4.63E+07 3.58E+07 3.18E+07 1.48E+07 2.74E+06 1.38E+03 0.00E+00 Ba135m 4.19E+04 4.19E+04 4.13E+04 4.09E+04 3.45E+04 2.35E+04 4.12E+03 1.17E-03 9.17E-19 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 5 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Cs136 7.34E+06 7.34E+06 7.33E+06 7.33E+06 7.21E+06 6.96E+06 5.94E+06 1.51E+06 6.41E+04 5.60E+02 3.24E-02 6.31E-19 Ba136m 8.40E+05 8.25E+05 8.17E+05 8.17E+05 8.10E+05 7.79E+05 6.66E+05 1.70E+05 7.18E+03 6.27E+01 3.63E-03 7.07E-20 Cs137 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 172E+07 1.70E+07 1.62E+07 Ba137m 1.65E+07 1.65E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.63E+07 1.63E+07 1.62E+07 1.60E+07 1.53E+07 Cs138 2.05E+08 2.05E+08 1.51E+08 8.94E+07 1.15E+04 1.21E-05 0.00E+00 Ba139 1.95E+08 1.95E+08 1.70E+08 1.34E+08 4.32E+06 1.66E+03 7.15E-13 0.00E+00 Ba140 1.96E+08 1.96E+08 1.96E+08 1.96E+08 1.93E+08 1.86E+08 1.57E+08 3.84E+07 1.47E+06 1.11E+04 4.69E-01 2.69E-18 La140 2.09E+08 2.09E+08 2.09E+08 2.09E+08 2.08E+08 2.03E+08 1.78E+08 4.42E+07 1.70E+06 1.28E+04 5.40E-01 3.10E-18 La141 1.78E+08 1.78E+08 1.72E+08 1.60E+08 4.70E+07 2.77E+06 8.17E+00 0.00E+00 Ce141 1.80E+08 1.80E+08 1.80E+08 1.80E+08 1.79E+08 1.76E+08 1.66E+08 9.55E+07 2.64E+07 3.88E+06 7.46E+04 1.28E-02 La142 1.74+08 1.74E+08 1.54E+08 1.24E+08 5.10E+06 3.42E+03 1.82E-11 0.00E+00 Pr142 7.33E+06 7.33E+06 7.20E+06 7.07E+06 5.49E+06 3.07E+06 2.26E+05 3.38E-05 0.00E+00 Ce143 1.67E+08 1.67E+08 1.66E+08 1.64E+08 1.42E+08 1.02E+08 2.24E+07 4.55E+01 3.32E-12 0.00E+00 Pr143 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.60E+08 1.44E+08 3.90E+07 1.82E+06 1.83E+04 1.41E+00 8.71E-17 Ce144 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.50E+08 1.41E+08 1.21E+08 9.78E+07 6.22E+07 1.05E+07 Pr144 1.52E+08 1.52E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.50E+08 1.41E+08 1.21E+08 9.78E+07 6.22E+07 1.05E+07 Pr144m 2.12E+06 2.12E+06 2.12E+06 2.12E+06 2.12E+06 2.11E+06 2.09E+06 1.97E+06 1.70E+06 1.37E+06 8.71E+05 1.47E+05 Pr145 1.13E+08 1.13E+08 1.08E+08 1.02E+08 4.53E+07 7.11E+06 1.70E+03 0.00E+00 Nd147 7.24E+07 7.24E+07 7.23E+07 7.22E+07 7.09E+07 6.80E+07 5.62E+07 1.09E+07 2.47E+05 8.40E+02 7.02E-03 6.59E-23 Pm147 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.52E+07 2.52E+07 2.43E+07 2.28E+07 1.99E+07 1.18E+07 Pm148 2.02E+07 2.02E+07 2.02E+07 2.01E+07 1.94E+07 1.78E+07 1.21E+07 5.38E+05 4.55E+04 1.00E+04 4.46E+02 2.11E-03 Pm148m 3.88E+06 3.88E+06 3.88E+06 3.87E+06 3.86E+06 3.81E+06 3.63E+06 2.35E+06 8.56E+05 1.89E+05 8.40E+03 4.00E-02 Nd149 4.16E+07 4.16E+07 3.47E+07 2.84E+07 1.70E+06 2.75E+03 7.49E-10 0.00E+00 Pm149 6.46E+07 6.46E+07 6.44E+07 6.42E+07 5.94E+07 4.82E+07 1.89E+07 5.45E+03 3.72E-05 2.09E-17 0.00E+00 Pm150 5.53E+05 5.53E+05 4.87E+05 4.27E+05 6.99E+04 1.12E+03 9.09E-06 0.00E+00 Pm151 2.17E+07 2.17E+07 2.15E+07 2.13E+07 1.80E+07 1.21E+07 2.10E+06 5.07E-01 2.73E-16 0.00E+00 Sm151 6.80E+04 6.80E+04 6.80E+04 6.80E+04 6.81E+04 6.83E+04 6.87E+04 6.87E+04 6.86E+04 6.85E+04 6.82E+04 6.72E+04 Eu152m 2.48E+04 2.48E+04 2.38E+04 2.30E+04 1.37E+04 4.15E+03 1.96E+01 1.37E-19 0.00E+00 Sm153 5.31E+07 5.31E+07 5.27E+07 5.23E+07 4.71E+07 3.71E+07 1.26E+07 1.10E+03 4.73E-07 4.21E-21 0.00E+00 Gd153 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.05E+05 7.49E+05 6.30E+05 4.87E+05 2.86E+05 3.52E+04 Eu154 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.04E+06 1.02E+06 9.77E+05 8.29E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 6 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Eu155 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.29E+05 4.18E+05 4.04E+05 3.75E+05 2.78E+05 Sm156 2.57E+06 2.57E+06 2.48E+06 2.38E+06 1.42E+06 4.38E+05 2.16E+03 2.25E-17 0.00E+00 Eu156 2.74E+07 2.74E+07 2.74E+07 2.73E+07 2.70E+07 2.62E+07 2.29E+07 6.98E+06 4.51E+05 7.41E+03 1.58E+00 5.20E-15 Gd159 2.74E+07 2.74E+07 2.69E+07 2.65E+07 2.04E+07 1.12E+07 7.61E+05 5.77E-05 0.00E+00 Tb160 7.20E+06 7.20E+06 7.19E+06 7.19E+06 7.17E+06 7.13E+06 6.92E+06 5.40E+06 3.03E+06 1.28E+06 2.17E+05 1.97E+02 Tb161 4.33E+06 4.33E+06 4.32E+06 4.31E+06 4.19E+06 3.92E+06 2.90E+06 2.13E+05 5.14E+02 6.09E-02 5.04E-10 0.00E+00 Dy165 5.38E+05 5.38E+05 4.66E+05 4.02E+05 5.03E+04 4.34E+02 2.25E-07 0.00E+00 Dy166 2.09E+03 2.09E+03 2.09E+03 2.08E+03 1.96E+03 1.71E+03 9.24E+02 4.63E+00 2.26E-05 2.44E-13 1.74E-29 0.00E+00 Ho166 8.58E+04 8.58E+04 8.41E+04 8.33E+04 6.99E+04 4.69E+04 8.43E+03 8.10E+00 3.95E-05 4.26E-13 1.74E-29 0.00E+00 Hf175 3.62E+03 3.62E+03 3.62E+03 3.62E+03 3.61E+03 3.58E+03 3.48E+03 2.69E+03 1.49E+03 6.10E+02 9.78E+01 7.05E-02 Lu176m 1.64E+03 1.64E+03 1.50E+03 1.36E+03 3.58E+02 1.70E+01 1.86E-05 0.00E+00 Lu177 6.84E+02 6.84E+02 6.82E+02 6.81E+02 6.60E+02 6.17E+02 4.52E+02 3.12E+01 3.46E-01 1.93E-01 8.63E-02 3.72E-03 Hf180m 1.12E+04 1.12E+04 1.05E+04 9.86E+03 4.10E+03 5.45E+02 6.24E-02 0.00E+00 Hf181 2.38E+05 2.38E+05 2.38E+05 2.38E+05 2.37E+05 2.35E+05 2.23E+05 1.46E+05 5.47E+04 1.25E+04 6.07E+02 3.93E-03 W181 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.41E+03 1.39E+03 1.20E+03 8.48E+02 5.09E+02 1.76E+02 2.70E+00 Ta182 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.57E+04 2.53E+04 2.16E+04 1.51E+04 8.79E+03 2.87E+03 3.51E+01 Ta183 5.46E+04 5.46E+04 5.45E+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 0.00E+00 W183m 6.40E+04 6.29E+04 5.45E+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 0.00E+00 W185 3.52E+04 3.52E+04 3.52E+04 3.52E+04 3.51E+04 3.49E+04 3.39E+04 2.67E+04 1.54E+04 6.69E+03 1.21E+03 1.43E+00 Re186 2.58E+04 2.58E+04 2.57E+04 2.56E+04 2.43E+04 2.15E+04 1.24E+04 1.05E+02 1.73E-03 1.16E-10 1.99E-25 0.00E+00 W187 4.42E+05 4.42E+05 4.36E+05 4.30E+05 3.51E+05 2.21E+05 2.74E+04 3.77E-04 2.75E-22 0.00E+00 W188 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.94E+03 1.89E+03 1.45E+03 8.02E+02 3.25E+02 5.12E+01 3.49E-02 Re188 1.73E+05 1.73E+05 1.72E+05 1.69E+05 1.28E+05 6.75E+04 5.37E+03 1.47E+03 8.10E+02 3.29E+02 5.17E+01 3.52E-02 Os191 7.72E+01 7.72E+01 7.64E+01 7.64E+01 7.63E+01 7.49E+01 6.59E+01 2.05E+01 1.38E+00 2.40E-02 5.75E-06 3.06E-20 Os191m 5.82E+01 5.82E+01 5.67E+01 5.52E+01 3.81E+01 1.63E+01 3.62E-01 1.66E-15 0.00E+00 Ir192 3.01E+01 3.01E+01 3.01E+01 3.01E+01 3.00E+01 2.98E+01 2.90E+01 2.27E+01 1.29E+01 5.55E+00 9.78E-01 1.06E-03 Np236m 4.84E+02 4.84E+02 4.77E+02 4.70E+02 3.78E+02 2.31E+02 2.51E+01 1.13E-07 6.12E-27 0.00E+00 U237 1.00E+08 1.00E+08 1.00E+08 9.93E+07 9.70E+07 9.02E+07 6.64E+07 4.60E+06 1.02E+04 4.48E+02 4.37E+02 3.97E+02 Pu237 6.71E+02 6.71E+02 6.70E+02 6.70E+02 6.67E+02 6.60E+02 6.30E+02 4.23E+02 1.69E+02 4.23E+01 2.47E+00 3.35E-05 Np238 4.70E+07 4.70E+07 4.67E+07 4.64E+07 4.22E+07 3.39E+07 1.27E+07 2.55E+03 7.44E+00 7.43E+00 7.41E+00 7.34E+00 Pu238 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.57E+05 4.58E+05 4.63E+05 4.69E+05 4.75E+05 4.82E+05 4.81E+05 SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 7 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Np239 2.12E+09 2.12E+09 2.12E+09 2.11E+09 1.93E+09 1.59E+09 6.59E+08 3.16E+05 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Pu239 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.84E+04 4.87E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 Np240 3.93E+06 3.93E+06 2.81E+06 2.01E+06 1.82E+04 3.90E-01 6.20E-16 7.40E-16 1.01E-15 1.41E-15 2.25E-15 5.53E-15 Pu240 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 Pu241 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.91E+07 1.89E+07 1.87E+07 1.83E+07 1.66E+07 Am241 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.55E+04 2.57E+04 2.79E+04 3.29E+04 4.02E+04 5.51E+04 1.11E+05 Am242m 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.65E+03 1.65E+03 1.65E+03 1.63E+03 Am242 1.14E+07 1.14E+07 1.12E+07 1.09E+07 8.10E+06 4.04E+06 1.81E+05 1.65E+03 1.64E+03 1.64E+03 1.64E+03 1.62E+03 Cm242 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.66E+06 6.59E+06 5.91E+06 4.58E+06 3.12E+06 1.42E+06 6.47E+04 Pu243 4.19E+07 4.19E+07 3.90E+07 3.64E+07 1.37E+07 1.46E+06 6.17E+01 3.93E-05 3.93E-05 3.93E-05 3.93E-05 3.93E-05 Am243 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Cm243 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.86E+03 2.84E+03 2.80E+03 2.67E+03 Am244 1.38E+07 1.38E+07 1.33E+07 1.29E+07 7.95E+06 2.66E+06 1.90E+04 4.79E-15 0.00E+00 Cm244 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.89E+05 3.87E+05 3.83E+05 3.76E+05 3.48E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 1 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr H 3 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.41E+02 1.39E+02 1.35E+02 1.21E+02 Na 24 4.16E-01 4.16E-01 4.07E-01 3.97E-01 2.85E-01 1.34E-01 4.45E-02 6.81E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cr 51 9.20E+03 9.20E+03 9.19E+03 9.19E+03 9.12E+03 8.97E+03 8.32E+03 4.34E+03 9.68E+02 1.02E+02 9.90E-01 1.15E-08 Mn 54 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.64E+02 6.59E+02 6.22E+02 5.45E+02 4.46E+02 2.96E+02 5.84E+01 Fe 55 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.64E+03 3.58E+03 3.43E+03 3.22E+02 2.84E+03 1.71E+03 Mn 56 1.95E+04 1.95E+04 1.70E+04 1.49E+04 2.27E+03 3.07E+01 1.21E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Co 58 8.15E+02 8.15E+02 8.14E+02 8.14E+02 8.12E+02 8.07E+02 7.83E+02 6.07E+02 3.38E+02 1.40E+02 2.29E+01 1.82E-02 Fe 59 2.18E+02 2.18E+02 2.18E+02 2.18E+02 2.17E+02 2.15E+02 2.05E+02 1.37E+02 5.37E+01 1.32E+01 7.38E-01 8.44E-06 Ni 59 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 Co 60 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.26E+02 6.21E+02 6.07E+02 5.88E+02 5.50E+02 4.23E+02 Ni 63 7.67E+01 7.67E+01 7.67E+01 7.67E+01 7.66E+01 7.66E+01 7.66E+01 7.66E+01 7.65E+01 7.64E+01 7.61E+01 7.51E+01 Cu 64 1.01E+00 1.01E+00 9.85E-01 9.58E-01 6.54E-01 2.73E-01 5.37E-03 8.69E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ni 65 9.17E+01 9.17E+01 7.99E+01 6.97E+01 1.02E+01 1.25E-01 3.12E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ga 73 6.64E+00 6.64E+00 6.20E+00 5.77E+00 2.13E+00 2.17E-01 7.55E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ge 73m 6.58E+00 6.56E+00 6.12E+00 5.69E+00 2.10E+00 2.14E-01 7.46E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 76 5.54E+00 5.54E+00 5.47E+00 5.40E+00 4.49E+00 2.95E+00 4.42E-01 3.23E-08 1.09E-24 0.00E+00 0.00E+00 0.00E+00 Ge 77 8.69E+01 8.69E+01 8.44E+01 8.18E+01 5.32E+-01 2.00E+01 2.41E-01 5.74E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 77 2.64E+02 2.64E+02 2.63E+02 2.61E+02 2.38E+02 1.87E+02 5.40E+01 7.87E-04 5.41E-15 9.13E-32 0.00E+00 0.00E+00 Ge 78 9.12E+02 9.12E+02 7.20E+02 5.69E+02 2.08E+01 1.08E-02 1.80E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 78 9.26E+02 9.26E+02 9.02E+02 8.48E+02 1.04E+02 1.57E-01 1.75E-15 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br 82 7.95E+02 7.95E+02 7.89E+02 7.81E+02 6.81E+02 4.97E+02 1.21E+02 5.78E-04 3.05E-16 0.00E+00 0.00E+00 0.00E+00 Br 83 1.45E+04 1.45E+04 1.32E+04 1.17E+04 1.57E+03 1.55E+01 1.44E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 83m 1.47E+04 1.47E+04 1.46E+04 1.42E+04 4.10E+03 5.93E+01 6.08E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 85 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.48E+03 2.45E+03 2.41E+03 2.33E+03 2.05E+03 Kr 85m 2.88E+04 2.88E+04 2.69E+04 2.49E+04 8.44E+03 7.10E+02 1.03E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 86 4.64E+02 4.64E+02 4.63E+02 4.63E+02 4.58E+02 4.47E+02 4.00E+02 1.52E+02 1.63E+01 5.73E-01 5.83E-04 9.22E-16 Kr 87 5.70E+04 5.70E+04 4.38E+04 3.34E+04 7.35E+02 1.20E-01 1.10E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 88 7.80E+04 7.80E+04 6.91E+04 6.11E+04 1.11E+04 2.23E+02 5.16E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 2 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Rb 88 8.02E+04 8.02E+04 7.50E+04 6.76E+04 1.24E+04 2.48E+02 5.76E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 89 8.71E+01 8.71E+01 8.67E+01 8.64E+01 8.12E+01 7.05E+01 3.73E+01 1.50E-01 4.48E-07 2.30E-15 2.28E-32 0.00E+00 Sr 89 1.07E+05 1.07E+05 1.07E+05 1.07E+05 1.06E+05 1.05E+05 1.01E+05 7.07E+04 3.11E+04 9.04E+03 7.13E+02 3.19E-02 Y 89m 3.10E+02 3.01E+02 9.62E+01 9.58E+01 9.06E+01 7.98E+01 4.64E+01 6.73E+00 2.89E+00 8.41E-01 6.63E-02 2.96E-06 Sr 90 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.23E+04 2.23E+04 2.21E+04 2.18E+04 2.08E+04 Y 90 2.36E+04 2.36E+04 2.36E+04 2.36E+04 2.35E+04 2.33E+04 2.28E+04 2.24E+04 2.23E+04 2.21E+04 2.19E+04 2.08E+04 Sr 91 1.41E+05 1.41E+05 1.36E+05 1.31E+05 7.86E+04 2.45E+04 1.30E-02 2.40E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 91 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.43E+05 1.39E+05 1.02E+05 5.00E+04 1.72E+04 1.92E+03 3.35E-01 Y 91m 8.15E+04 8.15E+04 8.10E+04 7.97E+04 4.99E+04 1.56E+04 8.23E+01 1.52E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr 92 1.55E+05 1.55E+05 1.36E+05 1.20E+05 2.00E+04 3.34E+02 3.36E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 92 1.56E+05 1.56E+05 1.55E+05 1.52E+05 7.28E+04 4.93E+03 4.51E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 93 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 Y 93 1.23E+05 1.23E+05 1.20E+05 1.16E+05 7.19E+04 2.40E+04 1.71E+02 4.32E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 95 2.33E+05 2.33E+05 2.33E+05 2.33E+05 2.32E+05 2.30E+05 2.23E+05 1.68E+05 8.79E+04 3.32E+04 4.46E+03 1.64E+00 Nb 95 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.33E+05 2.17E+05 1.47E+05 6.51E+04 9.62E+03 3.61E+00 Nb 95m 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.55E+03 1.98E+03 1.03E+03 3.90E+02 5.25E+01 1.93E-02 Nb 96 4.78E+02 4.78E+02 4.71E+02 4.64E+02 3.78E+02 2.35E+02 2.77E+01 2.50E+07 6.80E-26 0.00E+00 0.00E+00 0.00E+00 Zr 97 2.43E+05 2.43E+05 2.38E+05 2.34E+05 1.75E+05 9.09E+04 4.74E+03 3.63E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb 97 2.45E+05 2.45E+05 2.44E+05 2.41E+05 1.88E+05 9.13E+04 4.76E+03 3.92E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb 97m 2.30E+05 2.30E+05 2.26E+05 2.22E+05 1.66E+05 8.62E+04 4.50E+03 3.46E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Mo 99 2.61E+05 2.61E+05 2.59E+05 2.58E+05 2.40E+-5 2.02E+05 9.50E+04 1.35E+02 3.60E-05 4.96E-15 0.00E+00 0.00E+00 Tc 99 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 Tc 99m 2.32E+05 2.32E+05 2.32E+05 2.31E+05 2.24E+05 1.95E+05 9.20E+04 1.30E+02 3.48E-05 4.80E-15 0.00E+00 0.00E+00 Ru103 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.47E+05 2.44E+05 2.31E+05 1.46E+05 5.07E+04 1.03E+04 3.93E+02 9.83E-04 Rh103m 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.46E+05 2.44E+05 2.31E+05 1.46E+05 5.06E+04 1.03E+04 3.92E+02 9.82E-04 Ru105 1.89E+05 1.89E+05 1.80E+05 1.66E+-5 5.58E+04 4.59E+03 6.01E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rh105 1.74E+05 1.74E+05 1.74E+05 1.74E+05 1.64E+05 1.25E+05 3.07E+04 1.50E-01 8.25E-14 2.28E-32 0.00E+00 0.00E+00 Rh105m 5.36E+04 5.36E+04 5.12E+04 4.74E+04 1.59E+04 1.31E+03 1.71E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 3 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Ru106 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.25E+05 1.19E+05 1.06E+05 8.99E+04 6.36E+04 1.63E+04 Rh106 1.36E+05 1.36E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.25E+05 1.19E+05 1.06E+05 8.99E+04 6.36E+04 1.63E+04 Rh106m 4.69E+03 4.69E+03 4.00E+03 3.41E+03 3.63E+02 2.17E+00 2.16E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag109m 7.74E+04 7.74E+04 7.57E+04 7.38E+04 5.18E+04 2.30E+04 6.03E+02 3.55E-04 3.24E-04 2.83E-04 2.15E-04 7.18E-05 Ag110 4.66E+04 4.53E+04 1.74E+01 1.74E+01 1.74E+01 1.74E+01 1.73+01 1.60E+01 1.36E+01 1.06E+01 6.33E+00 8.34E-01 Ag110m 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.27+03 1.18E+03 9.99E+02 7.78E+02 4.65E+02 6.13E+01 Pd111 1.14E+04 1.14E+-04 4.94E+03 2.21E+03 1.35E+02 1.80E+01 2.06E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pd111m 4.71E+02 4.71E+02 4.43E+02 4.16E+02 1.72E+02 2.29E+01 2.62E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag111 1.16E+04 1.16E+04 1.16E+04 1.16E+04 1.13+04 1.06E+04 8.01E+03 7.13E+02 2.68E+00 6.20E-04 2.03E-11 0.00E+00 Ag111m 1.16E+04 1.16E+04 5.22E+03 2.38E+03 1.68E+02 2.24E+01 2.57E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pd112 5.01E+03 5.01E+03 4.93E+03 4.85E+03 3.85E+03 2.27E+03 2.12E+02 2.51E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag112 5.03E+03 5.03E+03 5.02E+03 5.01E+03 4.38E+03 2.67E+03 2.49E+02 2.95E-07 7.45E-28 0.00E+00 0.00E+00 0.00E+00 In113m 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.24E+02 5.14E+02 4.40E+02 3.06E+02 1.78E+02 5.84E+01 7.17E-01 Sn113 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.23E+02 5.14E+02 4.40E+02 3.06E+02 1.78E+02 5,84E+01 7.17E-01 Ag113 2.74E+03 2.74E+03 2.58E+03 2.42E+03 9.81E+02 1.24E+02 1.14E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cd113m 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.11E+01 1.10E+01 1.07E+01 9.69E+00 In114 1.36E+02 1.35E+02 8.61E+01 8.61E+01 8.57E+01 8.49E+01 8.14E+01 5.66E+01 2.44E+01 6.93E+00 5.18E-01 1.88E-05 In114m 9.00E+01 9.00E+01 8.99E+01 8.99E+01 8.95E+01 8.87E+01 8.51E+01 5.91E+01 2.55E+01 7.24E+00 5.41E-01 1.96E-05 Cd115 1.43E+03 1.43E+03 1.42E+03 1.41E+03 1.29E+03 1.05E+03 4.13E+02 1.27E-01 9.90E-10 6.85E-22 0.00E+00 0.00E+00 Cd115m 7.03E+01 7.03E+01 7.03E+01 7.03E+01 7.00E+01 6.93E+01 6.61E+01 4.41E+01 1.74E+01 4.29E+00 2.41E-01 2.82E-06 In-115m 1.43E+03 1.43E+03 1.43E+03 1.43E+03 1.37E+03 1.14E+03 4.50E+02 1.43EE-01 1.92E-03 4.74E-04 2.66E-05 3.12E-10 Cd117 1.34E+03 1.34E+03 1.17E+03 1.02E+03 1.46E+02 1.69E+00 3.34E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cd117m 3.09E+02 3.09E+02 2.79E+02 2.52E+02 5.95E+01 2.19E+00 7.81E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 In117 1.00E+03 1.00E+03 9.94E+02 9.67E+02 3.06E+02 6.25E+00 1.01E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 In117m 1.23E+03 1.23E+03 1.21+03 1.18E+03 3.53E+02 6.24E+00 4.16E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sn117m 3.80E+03 3.80E+03 3.80E+03 3.79E+03 3.74E+03 3.61E+03 3.11E+03 8.24E+02 3.87E+01 3.94E-01 3.13E-05 2.12E-21 Sn119m 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.84E+03 3.81E+03 3.58E+03 3.11E+03 2.52E+03 1.62E+03 2.88E+02 Sn121 2.31E+03 2.31E+03 2.28E+03 2.25E+03 1.89E+03 1.25E+03 1.99E+02 8.96E-01 8.93E-01 8.91E-01 8.86E-01 8.63E-01

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 4 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Sn121m 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.96E-01 3.95E-01 3.92E-01 3.83E-01 Sb122 5.06E+02 5.06E+02 5.02E+02 5.00E+02 4.64E+02 3.91E+02 1.81E+02 2.28E-01 4.68E-08 4.33E-18 0.00E+00 0.00E+00 Te123m 8.01E+00 8.01E+00 8.00E+00 8.00E+00 8.00E+00 7.96E+00 7.82E+00 6.74E+00 4.76E+00 2.83E+00 9.65E-01 1.40E-02 Sn123 1.97E+02 1.97E+02 1.97E+02 1.97E+02 1.96E+02 1.95E+02 1.92E+02 1.67E+02 1.21E+02 7.48E+01 2.76E+01 5.49E-01 Sn123m 1.34E+03 1.34E+03 8.03E+02 4.78E+-2 3.35E-01 2.07E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb124 2.39E+02 2.39E+02 2.39E+02 2.39E+02 2.38E+02 2.36E+02 2.29E+02 1.69E+02 8.48E+01 3.01E+01 3.56E+00 7.92E-04 Sn125 8.09E+02 8.09E+02 8.08E+02 8.07E+02 7.89E+02 7.52E+02 6.07E+02 9.36E+01 1.25E+00 1.94E-03 3.18E-09 0.00E+00 Sn125m 3.08E+03 3.08E+03 3.49E+02 3.93E+01 2.06E-12 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb125 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.43E+03 2.33E+03 2.19E+03 1.92E+03 1.16E+03 Te125m 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.55E+02 5.30E+02 4.69E+02 2.83E+02 Sb126 8.31E+01 8.31E+01 8.31E+01 8.29E+01 8.16E+01 7.86E+01 6.65E+01 1.56E+01 5.67E-01 2.88E-02 2.53E-02 2.53E-02 Sb126m 9.43E+01 9.42E+01 3.17E+01 1.07E+01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 Sn127 5.62E+03 5.62E+03 4.77E+03 4.04E+03 4.01E+02 2.04E+00 9.74E-11 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb127 1.39E+04 1.39E+04 1.38E+04 1.38E+04 1.32E+04 1.17E+04 6.81E+03 6.31E+01 1.28E-03 1.18E-10 3.84E-25 0.00E+00 Te127 1.38E+04 1.38E+04 1.38E+04 1.38E+04 1.36E+04 1.28E+04 8.54E+03 2.04E+03 1.35E+03 7.63E+02 2.35E+02 2.26E+00 Te127m 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.34E+03 2.02E+03 1.38E+03 7.79E+02 2.40E+02 2.31E+00 Sb128 2.32E+03 2.32E+03 2.26E+03 2.20E+03 1.31E+03 3.84E+02 1.51E+00 2.15E-21 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb129 4.77E+04 4.77E+04 4.46E+04 4.12E+04 1.37E+04 1.10E+03 1.31E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te129 4.54E+04 4.54E+04 4.50E+04 4.39E+04 2.10E+04 7.03E+03 5.45E+03 3.19E+03 9.25E+02 1.45E+02 3.16E+00 9.02E-07 Te129m 9.19E+03 9.19E+03 9.19E+03 9.19E+03 9.17E+03 9.05E+03 8.51E+03 4.98E+03 1.44E+03 2.25E+02 4.94E+00 1.41E-06 Xe129m 2.23E+01 2.23E+01 2.22E+01 2.22E+01 2.17E+01 2.06E+01 1.63E+01 2.15E+00 2.00E-02 1.79E-05 9.55E-12 0.00E+00 I130 6.42E+03 6.42E+03 6.27E+03 6.10E+03 4.12E+03 1.68E+03 2.97E+01 1.89E-14 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te131 1.18E+05 1.18E+05 9.07E+04 5.76E+04 5.70E+03 3.94E+03 7.46E+02 4.09E-04 1.45E-18 0.00E+00 0.00E+00 0.00E+00 Te131m 3.04E+04 3.04E+04 3.01E+04 2.98E+04 2.53E+04 1.75E+04 3.32E+03 1.82E-03 6.45E-18 0.00E+00 0.00E+00 0.00E+00 I131 1.41E+05 1.41E+05 1.41E+05 1.41E+05 1.38E+05 1.32E+05 1.04E+05 1.11E+04 6.28E+01 2.68E-02 3.11E-09 0.00E+00 Xe131m 2.15E+03 2.15E+03 2.15E+03 2.15E+03 2.13E+03 2.11E+03 1.97E+03 7.02E+02 2.75E+01 1.52E-01 3.14E-06 1.03E-24 Te132 2.01E+05 2.01E+05 2.01E+05 2.00E+05 1.88E+05 1.63E+05 8.61E+04 3.41E+02 9.75E-04 4.71E-12 3.61E-29 0.00E+00 I132 2.06E+05 2.06E+05 2.05E+05 2.05E+05 1.93E+05 1.68E+05 8.87E+04 3.51E+02 1.00E-03 4.86E-12 3.72E-29 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 5 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Cs132 9.34E+00 9.34E+00 9.32E+00 9.30E+00 9.01E+00 8.39E+00 6.09E+00 3.77E-01 6.15E-04 4.05E-08 1.00E-16 0.00E+00 Te133m 1.24E+05 1.24E+05 8.59E+04 5.90E+04 3.08E+02 1.87E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I133 2.85E+05 2.85E+05 2.83E+05 2.80E+05 2.24E+05 1.32E+05 1.19E-05 1.11E-05 1.60E-26 0.00E+00 0.00E+00 0.00E+00 Xe133 2.74E+05 2.74E+05 2.74E+05 2.74E+05 2.73E+05 2.65E+05 1.95E+05 6.43E+03 2.31E+00 1.57E-05 3.63E-16 0.00E+00 Xe133m 9.22E+03 9.22E+03 9.21E+03 9.20E+03 9.03E+03 8.25E+03 3.93E+03 1.11E+00 6.25E-09 2.65E-21 0.00E+00 0.00E+00 I134 3.12E+05 3.12E+05 2.72E+05 2.20E+05 1.90E+03 7.08E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs134 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.56E+04 5.43E+04 5.14E+04 4.73E+04 3.99E+04 2.04E+04 CS134m 9.91E+03 9.91E+03 8.80E+03 7.81E+03 1.47E+03 3.27E+01 1.17E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I135 2.74E+05 2.74E+05 2.60E+05 2.46E+05 1.18E+05 2.18E+04 1.09E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe135 7.50E+04 7.50E+04 8.26E+04 8.90E+04 1.22E+05 7.02E+04 5.08E+02 1.50E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe135m 6.25E+04 6.25E+04 4.70E+04 4.14E+04 1.92E+04 3.55E+03 1.78E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba135m 1.82E+02 1.82E+02 1.80E+02 1.78E+02 1.50E+02 1.02E+02 1.79E+01 5.10E-06 3.98E-21 0.00E+00 0.00E+00 0.00E+00 Cs136 1.54E+04 1.54E+04 1.54E+04 1.53E+04 1.51E+04 1.46E+04 1.25E+04 3.17E+03 1.34E+02 1.17E+00 6.79E-05 1.32E-21 Ba136m 1.77E+03 1.73E+03 1.72E+03 1.72E+03 1.69E+03 1.63E+03 1.40E+03 3.55E+02 1.50E+01 1.31E-01 7.61E-06 1.48E-22 Cs137 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.28E+04 3.27E+04 3.23E+04 3.08E+04 Ba137m 3.14E+04 3.14E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.11E+04 3.10E+04 3.08E+04 3.05E+04 2.91E+04 Cs138 2.58E+05 2.58E+05 1.90E+05 1.12E+05 1.44E+01 1.53E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba139 2.46E+05 2.46E+05 2.13E+05 1.69E+05 5.43E+05 2.09E+00 9.00E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba140 2.47E+05 2.47E+05 2.46E+05 2.46E+05 2.42E+05 2.34E+05 1.99E+05 4.83E+04 1.85E+03 1.39E+01 5.91E-04 3.39E-21 La140 2.70E+05 2.70E+05 2.70E+05 2.70E+05 2.67E+05 2.60E+05 2.26E+05 5.57E+04 2.13E+03 1.60E+01 6.81E-04 3.91E-21 La141 2.24E+05 2.24E+05 2.16E+05 2.01E+05 5.90E+04 3.48E+03 1.03E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ce141 2.25E+05 2.25E+05 2.25E+05 2.25E+05 2.24E+05 2.21E+05 2.08E+05 1.19E+05 3.31E+04 4.86E+03 9.35E+01 1.60E-05 La142 2.17E+05 2.17E+05 1.90E+05 1.54E+05 6.33E+03 4.25E+00 2.26E-14 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pr142 1.72E+04 1.72E+04 1.69E+04 1.66E+04 1.29E+04 7.20E+03 5.29E+02 7.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ce143 2.04E+05 2.04E+05 2.03E+05 2.01E+05 1.73E+05 1.24E+05 2.73E+04 5.55E-02 4.06E-15 0.00E+00 0.00E+00 0.00E+00 Pr143 1.97E+05 1.97E+05 1.97E+05 1.97E+05 1.96E+05 1.95E+05 1.76E+05 4.74E+04 2.21E+03 2.22E+01 1.72E-03 1.06E-19 Ce144 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E+05 1.76E+05 1.52E+05 1.22E+05 7.77E+04 1.31E+04 Pr144 1.90E+05 1.90E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E+05 1.76E+05 1.52E+05 1.22E+05 7.77E+04 1.31E+04

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 6 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Pr144m 2.65E+03 2.65E+03 2.64E+03 2.64E+03 2.64E+03 2.64E+03 2.62E+03 2.46E+03 2.12E+03 1.71E+03 1.09E+03 1.84E+02 Pr145 1.40E+05 1.40E+05 1.33E+05 1.26E+05 5.59E+04 8.75E+03 2.09E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nd147 9.29E+04 9.29E+04 9.28E+04 9.27E+04 9.10E+04 8.73E+04 7.22E+04 1.40E+04 3.17E+02 1.08E+00 9.01E-06 8.47E-26 Pm147 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.15E+04 3.16E+04 3.16E+04 3.05E+04 2.85E+04 2.50E+04 1.47E+04 Pm148 2.88E+04 2.88E+04 2.87E+04 2.87E+04 2.76E+04 2.54E+04 1.73E+04 7.46E+02 5.61E+01 1.23E+01 5.50E-01 2.60E-06 Pm148m 4.78E+03 4.78E+03 4.78E+03 4.78E+03 4.75E+03 4.70E+03 4.47E+03 2.89E+03 1.05E+03 2.33E+02 1.04E+01 4.92E-05 Nd149 5.70E+04 5.70E+04 4.76E+04 3.90E+04 2.34E+03 3.77E+00 1.03E-12 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm149 9.22E+04 9.22E+04 9.19E+04 9.16E+04 8.47E+04 6.88E+04 2.69E+04 7.77E+00 5.30E-08 2.99E-20 0.00E+00 0.00E+00 Pm150 1.00E+03 1.00E+03 8.81E+02 7.74E+02 1.27E+02 2.02E+00 1.65E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm151 3.15E+04 3.15E+04 3.13E+04 3.10E+04 2.61E+04 1.77E+04 3.05E+03 7.37E-04 3.96E-19 0.00E+00 0.00E+00 0.00E+00 Sm151 9.85E+01 9.85E+01 9.85E+01 9.86E+01 9.87E+01 9.90E+01 9.95E+01 9.96E+01 9.95E+01 9.93E+01 9.89E+01 9.74E+01 Eu152m 3.63E+01 3.63E+01 3.50E+01 3.37E+01 2.00E+01 6.10E+00 2.88E-02 2.01E-22 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sm153 1.04E+05 1.04E+05 1.03E+05 1.02E+05 9.22E+04 7.26E+04 2.47E+04 2.15E+00 9.25E-10 8.24E-24 0.00E+00 0.00E+00 Gd153 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.04E+03 9.67E+02 8.13E+02 6.28E+02 3.69E+02 4.54E+01 Eu154 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.39E+03 2.36E+03 2.32E+03 2.22E+03 1.89E+03 Eu155 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.00E+03 9.78E+02 9.42E+02 8.74E+02 6.50E+02 Sm156 4.23E+03 4.23E+03 4.07E+03 3.93E+03 2.34E+03 7.20E+02 3.56E+00 3.70E-20 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu156 7.32E+04 7.32E+04 7.31E+04 7.31E+04 7.22E+04 7.00E+04 6.11E+04 1.87E+04 1.21E+03 1.98E+01 4.21E-03 1.39E-17 Gd159 4.36E+04 4.36E+04 4.28E+04 4.21E+04 3.24E+04 1.78E+04 1.21E+03 9.18E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tb160 1.64E+04 1.64E+04 1.64E+04 1.64E+04 1.63E+04 1.62E+04 1.58E+04 1.23E+04 6.91E+03 2.92E+03 4.95E+02 4.50E-01 Tb161 8.08E+03 8.08E+03 8.06E+03 8.05E+03 7.81E+03 7.30E+03 5.41E+03 3.97E+02 9.57E-01 1.13E-04 9.41E-13 0.00E+00 Dy165 3.06E+03 3.06E+03 2.65E+03 2.29E+03 2.86E+02 2.47E+00 1.28E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Dy166 1.45E+01 1.45E+01 1.45E+01 1.44E+01 1.36E+01 1.19E+01 6.43E+00 3.21E-02 1.57E-07 1.69E-15 6.85E-32 0.00E+00 Ho166 7.57E+02 7.57E+02 7.48E+02 7.38E+02 6.18E+02 4.13E+02 7.14E+01 5.01E-02 2.44E-07 2.63E-15 9.13E-32 0.00E+00 Hf175 2.86E+00 2.86E+00 2.86E+00 2.86E+00 2.85E+00 2.83E+00 2.75E+00 2.13E+00 1.17E+00 4.82E-01 7.70E-02 5.57E-05 Lu176m 2.60E+00 2.60E+00 2.36E+00 2.14E+00 5.65E-01 2.67E-02 2.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Lu177 1.12E+00 1.12E+00 1.12E+00 1.11E+00 1.08E+00 1.01E+00 7.40E-01 5.10E-02 6.23E-04 3.53E-04 1.59E-04 6.83E-06 Hf180m 1.19E+01 1.19E+01 1.12E+01 1.05E+01 4.35E+00 5.80E-01 6.64E-05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 7 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay Time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Hf181 4.48E+02 4.48E+02 4.48E+02 4.47E+02 4.45E+02 4.40E+02 4.19E+02 2.74E+02 1.03E+02 2.36E+01 1.14E+00 7.39E-06 W181 1.87E+00 1.87E+00 1.87E+00 1.87E+00 1.87E+00 1.86E+00 1.83E+00 1.58E+00 1.12E+00 6.68E-01 2.32E-01 3.55E-03 Ta182 5.61E+01 5.61E+01 5.61E+01 5.61E+01 5.60E+01 5.58E+01 5.48E+01 4.68E+01 3.26E+01 1.90E+01 6.21E+00 7.60E-02 Ta183 1.54E+02 1.54E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 0.00E+00 W183m 1.69E+02 1.67E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 0.00E+00 W185 5.96E+01 5.96E+01 5.96E+01 5.96E+01 5.94E+01 5.91E+01 5.74E+01 4.52E+01 2.60E+01 1.13E-01 2.05E+00 2.42E-03 Re186 5.05E+01 5.05E+01 5.03E+01 5.01E+01 4.75E+01 4.21E+01 2.42E+01 2.05E-01 3.39E-06 2.27E-13 3.89E-28 0.00E+00 W187 5.17E+02 5.17E+02 5.10E+02 5.02E+02 4.10E+02 2.58E+02 3.19E+01 4.41E-07 3.22E-25 0.00E+00 0.00E+00 0.00E+00 W188 2.72E+00 2.72E+00 2.71E+00 2.71E+00 2.71E+00 2.69E+00 2.61E+00 2.01E+00 1.11E+00 4.50E-01 7.09E-02 4.83E-05 Re188 3.39E+02 3.39E+02 3.36E+02 3.31E+02 2.50E+02 1.31E+02 9.44E+00 2.03E+00 1.12E+00 4.55E-01 7.16E-02 4.88E-05 Os191 6.33E-01 6.33E-01 6.33E-01 6.33E-01 6.29E-01 6.17E-01 5.43E-01 1.69E-01 1.13E-02 1.98E-04 4.74E-08 2.52E-22 Os191m 4.70E-01 4.70E-01 4.58E-01 4.46E-01 3.08E-01 1.32E-01 2.93E-03 1.34E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ir192 3.14E-01 3.14E-01 3.14E-01 3.14E-01 3.13E-01 3.12E-01 3.03E-01 2.37E-01 1.35E-01 5.80E-02 1.02E-02 1.12E-05 Np236m 1.15E+00 1.15E+00 1.14E+00 1.12E+00 9.01E-01 5.51E-01 5.99E-02 2.69E-10 1.46E-29 0.00E+00 0.00E+00 0.00E+00 U237 1.67E+05 1.67E+05 1.66E+05 1.66E+05 1.61E+05 1.50E+05 1.11E+05 7.65E+03 1.69E+01 7.36E-01 7.16E-01 6.50E-01 Pu237 2.52E+00 2.52E+00 2.52E+00 2.52E+00 2.51E+00 2.48E+00 2.37E+00 1.59E+00 6.33E-01 1.59E-01 9.28E-03 1.26E-07 Np238 1.11E+05 1.11E+05 1.11E+05 1.10E+05 9.99E+04 8.03E+04 3.01E+04 6.05E+00 1.35E-02 1.35E-02 1.35E-02 1.33E-02 Pu238 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.39E+03 1.40E+03 1.41E+03 1.43E+03 1.45E+03 1.44E+03 Np239 3.26E+06 3.26E+06 3.26E+06 3.24E+06 2.98E+06 2.45E+06 1.01E+06 4.94E+02 1.23E+01 1.23E+01 1.23E+01 1.23E+01 Pu239 6.13E+01 6.13E+01 6.13E+01 6.13E+01 6.14E+01 6.16E+01 6.19E+01 6.22E+01 6.22E+01 6.22E+01 6.22E+01 6.22E+01 Np240 7.18E+03 7.18E+03 5.13E+03 3.67E+03 3.33E+01 7.13E-04 4.03E-17 4.55E-17 5.72E-17 7.48E-17 1.11E-16 2.54E-16 Pu240 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.33E+02 Pu241 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.13E+04 3.10E+04 3.07E+04 2.99E+04 2.72E+04 Am241 4.63E+01 4.63E+01 4.63E+01 4.63E+01 4.64E+01 4.65E+01 4.69E+01 5.05E+01 5.86E+01 7.08E+01 9.54E+01 1.86E+02 Am242m 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.00E+00 3.00E+00 3.00E+00 2.99E+00 2.96E+00 Am242 2.46E+04 2.46E+04 2.40E+04 2.35E+04 1.74E+04 8.70E+03 3.88E+02 2.99E+00 2.99E+00 2.98E+00 2.98E+00 2.95E+00 Cm242 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.69E+04 1.52E+04 1.17E+04 8.01E+03 3.64E+03 1.65E+02 Pu243 1.22E+05 1.22E+05 1.14E+05 1.06E+05 4.00E+04 4.26E+03 1.80E-01 9.63E-07 9.63E-07 9.63E-07 9.63E-07 9.63E-07

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 8 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay Time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Am243 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 Cm243 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.04E+01 1.03E+01 9.80E+00 Am244 6.13E+04 6.13E+04 5.92E+04 5.72E+04 3.54E+04 1.18E+04 8.44E+01 2.13E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cm244 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.60E+03 2.58E+03 2.53E+03 2.34E+03

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES

Specific design features for maintaining personnel exposures ALARA are discussed in this subsection.

12.3.1.1 Common Equipment and Component Designs for ALARA

This subsection describes the design features used for several general classes of equipment or components. These classes of equipment are common to many of the plant systems; thus, the features employed for each system to maintain minimum exposures are similar and are discussed by equipment class in the following paragraphs.

Filters: Whenever practicable, filters that accumulate radioactive material are supplied with the means either to backflush the filter remotely or to perform cartridge replacement with semi-remote tools (i.e., long handled tools). For cartridge filters, adequate space is provided to allow removing, cask loading, and transporting the cartridge to the solid radwaste area.

Demineralizers: Demineralizers for radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin tanks prior to dewatering or solidification and that fresh resin can be loaded into the demineralizer remotely. Underdrains and downstream strainers are designed for full system pressure drop. The demineralizers and piping are designed with provisions for being flushed.

Evaporators: Evaporators are provided with chemical addition connections to allow the use of chemicals for descaling operations. Space is provided to allow uncomplicated removal of heating tube bundles. To the extent practicable, the more radioactive components are separated from those that are less radioactive by a shield wall.

Pumps: Wherever practicable, pumps, in radioactive areas are purchased with mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space for access to the pumps for servicing. Small pumps are installed in a manner that allows easy removal if necessary. All pumps in radioactive waste systems are provided with flanged connections for ease in removal. Generally, pump casings are provided with drain connections for draining the pump for maintenance.

Tanks: Whenever practicable, tanks are provided with sloped bottoms and bottom outlet connections. Overflow lines are directed to the waste collection system in order to control contamination within plant structures.

Heat Exchangers: Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to-tube sheet joints welded to minimize leakage. Impact baffles are provided and tube side and shell side velocities are limited to minimize erosive effects.

Instruments: Instrument devices are located in low radiation zones and away from radiation sources when practicable. Where practicable primary instrument devices, which for functional SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-2 reasons are located in high radiation zones, are designed for easy removal to a lower radiation zone for calibration. Where practicable transmitters and readout devices are located in low radiation zones, such as corridors and the control room, for servicing. Some instruments in high radiation zones are provided in duplicate to reduce access and service time required.

Seals are provided on instrument sensing lines on process piping that may contain highly radioactive materials to reduce the servicing time required to keep the lines free of solids.

Instrument and sensing line connections are located in such a way as to avoid corrosion product and radioactive gas buildup.

Valves: To minimize personnel exposures from valve operations, motor-operated, air operated, or other remotely actuated valves are used to the maximum extent practicable.

When practicable, valves are located in valve galleries so that they are shielded separately from the major components that accumulate radioactivity. Long runs of exposed piping are minimized in valve galleries. In areas where manual valves are used on frequently operated process lines, either remote valve operators or shielding is normally provided to minimize personnel exposure.

For equipment located in Zone V areas, remote actuators are provided for frequently operated valves associated with system operation. All other valve operations are either infrequent or performed with equipment in the shutdown mode. To the maximum practicable extent, simple straight reach rods will be used to allow operators to retain the feel of whether the valves are tightly closed or not. Where practicable valves with reach rods are installed with their stems horizontal so that the reach rods are also horizontal but above the heads of personnel to allow ready access.

For valves in radiation areas, provisions are made to drain adjacent radioactive components where practicable.

Wherever practicable, valves for clean, non-radioactive systems are separated from radioactive sources and are located in readily accessible areas.

Manually operated valves in the filter/demineralizer valve compartments required for normal operation and shutdown are equipped with reach rods extending through or over the valve gallery wall. Personnel are not required to enter the valve gallery during flushing operations. The valve gallery shield walls are designed for maximum expected filter backflush activities during flushing operations.

Full ported valves are used in systems expected to contain radioactive solids.

Special valve designs with minimum internal crevices are normally used where crud trapping could become a problem, especially for piping carrying spent resin or evaporator bottoms.

Piping: The piping in pipe chases is designed for the lifetime of the unit. The number of valves or instrumentation in the pipe chases has been reduced to maximum extent practicable. Where radioactive piping is routed through areas that require routine maintenance, pipe chases are normally provided to reduce the radiation contribution from these pipes. Wherever practicable, piping containing radioactive material is routed to minimize radiation exposure to the unit personnel.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-3 Floor Drains: Floor drains and properly sloped floors are provided for each room or cubicle, within which are serviceable components containing radioactive liquids. If a radioactive drain line must pass through a zone lower than that at which it will terminate, proper shielding is provided. Local gas traps or porous seals are not used on radwaste floor drains. Gas traps are provided at the common sump or tank.

Lighting: Multiple electric lights are provided for each cell or room containing highly radioactive components so that the burnout of a single lamp will not require entry and immediate replacement of the defective lamp. Normally, incandescent lights that require less time for servicing are provided to minimize personnel exposure. The fluorescent lights used in some areas do not require frequent service because of the increased life of the tubes.

HVAC: The HVAC system design provides for rapid replacement of the filter elements and housings.

Sample Stations: Sample stations for routine sampling of process fluids are located in accessible areas. Shielding is provided at the sample stations as required to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling. Ventilation, drains or other means of contamination control are provided where necessary. The counting room and laboratory facilities are described in Section 12.5.

Clean Services: Whenever practicable, active components of clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways.

12.3.1.2 Common Facility and Layout Designs for ALARA

This subsection describes the design features used for standard type plant processes and layout situations. These features are used in conjunction with the general equipment designs described in Subsection 12.3.1.1 and include the features discussed in the following paragraphs.

Valve Galleries: Valve galleries are provided with shielded entrances for personnel protection. In many cases the valve galleries are divided by shielding or distance into subcompartments that service only two or three components and are further subdivided by fin walls so that personnel are only exposed to the valves and piping associated with one component at any given location. Threshold berms and floor drains are provided to control radioactive leakage. To facilitate decontamination in valve galleries, concrete surfaces are covered with a smooth surfaced coating which will allow easy decontamination.

Piping: Pipes carrying radioactive materials are routed through controlled access areas zoned for that level of activity. Each piping run is analyzed to determine the potential radioactivity level and surface dose rate. Where radioactive piping must be routed through corridors or other low radiation zone areas, shielded pipeways are normally provided. Whenever practicable, valves and instruments are not placed in radioactive pipeways. Whenever practicable, equipment compartments contain only piping associated with equipment in that compartment.

Where practicable piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be eliminated. Where possible, thermal expansion loops are raised rather than dropped. In radioactive systems, the use of non-removable SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-4 backing rings in the piping joints is minimized to eliminate a potential crud trap for radioactive materials. Piping carrying resin slurries or evaporator bottoms is run vertically as much as possible.

Whenever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe.

Field Run Piping: All routing of radioactive process piping, large and small, is reviewed by the design engineering office.

Penetrations: To minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between the source and the accessible areas. If offsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If these two methods are not used, then baffle shield walls or grouting the area around the penetration are provided.

Contamination Control: Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from certain radioactive systems are piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain. All-welded piping systems are used on contaminated systems to the maximum extent practicable to reduce system leakage and crud buildup at joints. The valves in some radioactive systems are provided with leak-off connections piped directly to the collection system.

Decontamination of potentially contaminated areas within the plant is facilitated by the application of suitable smooth surfaced coatings to the concrete floors and walls.

Floor drains with properly sloping floors are provided in potentially contaminated areas of the plant.

In addition, radioactive and potentially radioactive drains are separated from non-radioactive drains. Systems that become highly radioactive, such as the radwaste slurry transport system, are provided with flush and drain connections. Certain systems have provisions for chemical and mechanical cleaning prior to maintenance.

Equipment Layout: In systems where process equipment is a major radiation source (such as fuel pool cleanup, radwaste, condensate demineralizer, etc.), pumps, valves, and instruments are normally separated from the process component. This allows servicing and maintenance of items in reduced radiation zones. Control panels are located in lowest practicable radiation zones.

Major components (such as tanks, demineralizers, and filters) in radioactive systems are isolated in individual shielded compartments insofar as practicable.

Provision is made on some major plant components for removal of these components to lower radiation zones for maintenance.

Labyrinth entrance way shields or shielding doors are provided for compartments from which radiation could stream to access areas and exceed the radiation zone dose limits for those areas.

For potentially high radiation components (such as filters and demineralizers), completely enclosed shielded compartments with hatch openings are used.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-5 Equipment in non-radioactive systems that requires lubrication is located outside radiation areas.

Wherever practicable, lubrication of equipment in radiation areas is achieved with the use of tube type extensions to reduce exposure during maintenance.

Figures 12.3-1, 12.3-2, 12.3-3, 12.3-4, 12.3-5, and 12.3-6 provide layout arrangements for demineralizers, filters, spent resin storage tanks, hydrogen recombiners, and their associated valve compartments or galleries.

Exposure from routine in-plant inspection is controlled by locating, whenever possible, inspection points in shielded low background radiation areas. Radioactive and non-radioactive systems are separated as far as practicable to limit radiation exposure from routine inspection of non-radioactive systems. For radioactive systems, emphasis is placed on space and ease of motion in a shielded inspection area. Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is normally provided. In high radiation areas where routine surveillance is required, remote viewing devices are provided when practicable.

Facilities for Handling Sealed and Unsealed Radioactive Material:

As discussed in Subsection 12.2.1.9, special material used in the radiochemistry laboratory require the design of special handling equipment. For unsealed materials, the following is provided:

a) Exhaust hoods that exhaust to the ventilation system are located in areas such as sample stations and the radiochemistry laboratory.

b) Decontamination facilities, radiochemistry laboratory, controlled zone shop, instrument repair shops and washdown area are situated at various locations in the plant and are described in Subsection 12.5.2.

c) An area for the repair and maintenance of removed control rod drives is provided in the reactor building in close proximity to the control rod drive removal hatch.

12.3.1.3 Radiation Zoning and Access Control Access to areas inside the plant structures and plant yards is regulated and controlled. Each radiation zone defines the radiation level range to which the aggregate of all contributing sources must be attenuated by shielding.

All plant areas are categorized into radiation zones according to expected radiation levels and anticipated personnel occupancy, with consideration given toward maintaining personnel external exposures ALARA and within the standards of 10CFR20. Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal operation and shutdown; for maintenance occupancy requirements, and for general access requirements to determine appropriate zoning. Radiation zone categories used and their descriptions are given in Table 12.3-1 and the specific zoning for each plant area is shown on Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-6 B1N-100, Sh. 1. Note that the radiation zoning for Unit 1 is not significantly different from those for Unit 2, therefore Dwgs. A 511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1 are also representative of Unit

2. Attachment of the new common tool room facility to Unit 2 turbine building (Ref. Dwg. M-231, Sh. 1) represents a unique feature not represented by the radiation zoning for Unit 1 turbine building at elevation 676' (Dwg. A-514, Sh. 1). The radiation zoning for the tool room is Zone II. Where possible, frequently accessed areas, i.e., corridors, are shielded for Zone I and Zone II access.

The control of ingress or egress of plant operating personnel to controlled access areas and procedures employed to ensure that radiation levels and allowable working time are within the limits prescribed by 10CFR20 as described in Section 12.5.

12.3.1.4 Control of Activated Corrosion Products

In order to minimize the radiation exposure associated with the deposition of activated corrosion products in reactor coolant and auxiliary systems, the following steps have been taken:

(1) The reactor coolant system consists mainly of austenitic stainless steel, carbon steel and low alloy steel components. Nickel content of these materials is low, and it is controlled in accordance with applicable ASME material specifications.

A small amount of nickel base material (Inconel 600) is employed in the reactor vessel internal components. Inconel 600 is required where components are attached to the reactor vessel shell and the coefficient of expansion must match the thermal expansion characteristics of the low alloy vessel steel. Inconel 600 was selected because it provides the proper thermal expansion characteristics, adequate corrosion resistance and can be readily fabricated and welded.

(2) Materials employed in the reactor coolant system are purchased to ASME material specification requirements. No special controls on levels of cobalt impurities are specified.

(3) Hardfacing and wear materials having a high percentage of cobalt are restricted to applications where no satisfactory alternate materials are available. The EPRI cobalt reduction guidelines (Ref. 12.3-24) are utilized to the ex tent practical.

(4) A high temperature filtration system was not employed in the Reactor Water Clean-up System. The reasons for this included:

a) Lack of quantitative data on the removal efficiency for insoluble cobalt by the high temperature filter; b) Uncertainty in the deposition model including the relative effectiveness of cobalt removal on deposition rate;

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-7 c) Doubtful cost-effectiveness in an area where other methods under study (such as decontamination) may prove better at reducing dose rates while also being more cost-effective.

(5) Items 1, 2, and 3 above also apply to valve materials in contact with reactor coolant. Valve packing materials are selected primarily for their properties in the particular environment.

(6) Subsections 12.1.2.2, 12.3.1.1, and 12.3.1.2 describe the various flushing, draining, testing, and chemical addition connections which have been incorporated into the design of piping and equipment which handle radioactive materials. If decontamination is to be performed, these connections would be used for that purpose.

(7) The plant is designed with a powdered resin, pressure precoat clean-up system for the primary coolant in the reactor and a full flow deep bed condensate demineralizer and filter vessel system for the feedwater. See Dwgs. M-116, Sh. 1, M-116, Sh. 2, M-116, Sh. 3, M-144, Sh. 1, M-144, Sh. 2, M1-G33-16, Sh. 1, M1-G33-18, Sh. 1 and M-145, Sh. 1.

(8) A chemistry control program has been developed and implemented at SSES to reduce crud buildup.

12.3.2 SHIELDING

In this subsection the bases for the nuclear radiation shielding and the shielding configurations are

discussed.

12.3.2.1 Design Objectives

The basic objective of the plant radiation shielding is to reduce personnel exposures, in conjunction with a program of controlled personnel access to and occupancy of radiation areas, to levels that are within the dose regulations of 10CFR50 and are as low as reasonably achievable (ALARA) within the dose regulations of 10CFR20. Shielding and equipment layout and design are considered in ensuring that exposures are kept ALARA during anticipated personnel activities in areas of the plant containing radioactive materials.

Basic plant conditions considered in the nuclear radiation shielding design are normal operation at full-power, and plant shutdown.

The shielding design objectives for the plant during normal operation, including anticipated operational occurrences, and for Shutdown operations are:

a. To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10CFR20 b. To ensure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-related operations required for each plant equipment and instrumentation area

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-8 c. To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.

d. To sufficiently shield the control room so that the total dose from all post-accident sources (Rem-TEDE).(calculated in Chapter 15) in the event of design basis accidents will not exceed the limits of 10CFR50.67.

12.3.2.2 General Shielding Design Shielding is provided to attenuate direct radiation through walls and penetrations and scattered radiation to less than the upper limit of the radiation zone for each area shown in Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1. Since the layout for Unit 2 is not significantly different from that of Unit 1, the minimum shielding requirements (see Subsection 12.3.2.3) indicated on those drawings are applicable to both Units. General locations of the plant areas and equipment discussed in this subsection are also shown on those drawings.

The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 145 lb/ft

3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design ensures protection on an equivalent shielding basis as determined by the characteristics of the concrete block selected. Compliance of concrete radiation shield design with Regulatory Guide 1.69 is discussed in Section 3.13. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.

Special features employed to maintain radiation exposures ALARA in routinely occupied areas such as valve operating stations and sample stations are described in Subsections 12.3.1.1 and 12.3.1.2.

12.3.2.2.1 Reactor Building Shielding Design

During reactor operation, the steel-lined, reinforced concrete drywell wall and the reactor building walls protect personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and associated equipment within the reactor building. The reactor vessel shield wall, drywell wall, and various equipment compartment walls together with the reactor building walls minimize the radiation levels outside the reactor building.

Where personnel and equipment hatches or penetrations pass through the drywell wall, additional shielding is designed to attenuate the radiation to below the required level defined by the radiation zone outside the drywell wall during normal operation and shutdown and to acceptable emergency levels as defined by 10CFR50 during design basis accidents.

12.3.2.2.2 Reactor Building Interior Shielding Design Inside Drywell Structure: Areas within the drywell are designed as Zone V areas and are normally inaccessible during plant operation. The reactor vessel shield provides shielding for access in the SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-9 drywell during shutdown, and reduces the activation of and radiation damage to drywell equipment and materials.

Outside Drywell Structure: The drywell wall is designed to reduce radiation levels in normally occupied areas of the reactor building from sources within the drywell to less than the maximum level for Zone II.

Penetrations and hatch openings in the drywell wall are shielded, as necessary, to meet adjacent area radiation zoning levels. Shielding requirements for the personnel, equipment, and CRD removal hatch openings are shown on Dwg. A-522, Sh. 1 in the areas numbered 412, 413, and 402, respectively. Drywell piping and electrical penetrations are shielded by providing either local shields within the penetration assembly or a shielded penetration room. Shielded piping penetration room locations and bulk shielding requirements are shown on Dwgs. A-521, Sh. 1, A-522, Sh. 1, and A-523, Sh. 1. These rooms, numbered 202, 204, 205; 403, 411, 501, 504, 506, 515; are designated radiation Zone V during reactor power operation and are provided with personnel access controls. Electrical penetrations which are not located within these rooms are provided with supplementary local shielding as needed to meet outside zoning levels.

The components of the reactor water cleanup (RWCU) system described in Section 5.4.8 are located in shielded compartments which are designed as Zone V, restricted access areas. Shielding is provided for each piece of equipment in the RWCU system consistent with its postulated maximum activity Subsection 12.2.1 and with the access and zoning requirements of the adjacent areas. This equipment includes:

a) Regenerative heat exchanger b) Non-regenerative heat exchanger

c) RWCU pumps and piping

d) RWCU filter demineralizers and holdup pumps

e) RWCU backwash receiving tank and piping.

The traversing in-core probe (TIP) system is located inside a shielded compartment to protect personnel from the neutron activated portion of the TIP cable.

Main steamlines are located within shielded structures from the drywell wall to the reactor building

wall.

Spent fuel is a primary source of radiation during refueling. Because of the extremely high activity of the fission products contained in the spent fuel assemblies and the proximity of Zone II areas, shielding is provided for areas surrounding the fuel transfer canal and pool to ensure that radiation levels remain below zone levels specified for adjacent areas.

After reactor shutdown, the Residual Heat Removal (RHR) System pumps and heat exchangers are in operation to remove heat from the reactor water. It is anticipated that the radiation levels in the vicinity of this equipment will temporarily reach Zone V levels due to corrosion and fission products in the reactor water. Shielding is designed to attenuate radiation from RHR equipment SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-10 during shutdown cooling operations to levels consistent with the radiation zoning requirements of adjacent areas.

During functional testing operations of the Reactor Core Isolation Cooling (RCIC) System and the High Pressure Coolant Injection (HPCI), the steam driven turbine and the inlet and exhaust piping are shielded consistent with the maximum steam activities in the lines and the access zone requirements of surrounding areas.

The concrete shield walls surrounding the spent fuel cask loading, storage, and transfer areas, as well as the shield walls surrounding the fuel transfer and storage areas, are designed to provide Zone II maximum dose rates in accessible areas outside of the shield walls.

Water in the spent fuel pool provides shielding above the spent fuel transfer and storage areas.

Direct radiation levels at the fuel handling equipment are calculated to be less than 2.5 mrem/hr from spent fuel during normal operations.

Water is also used as shielding material above the steam dryer and separator storage area.

Concrete walls and water in the pool are designed to provide Zone II dose rates in adjacent accessible areas during storage of the dryer and separator.

The Fuel Pool Cooling and Cleanup (FPCC) System (see Section 9.1.3) shielding is based on the maximum activity discussed in Subsection 12.2.1 and the access and zoning requirements of adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, pumps and piping, filter demineralizers, and backwash receiving tank.

12.3.2.2.3 Radwaste Building Shielding Design

Shielding is provided as necessary around the following equipment in the radwaste building to ensure that the radiation zone and access requirements are met for surrounding areas.

a) Laundry drain tank and pumps b) Chemical waste tank and pumps c) Radwaste evaporators d) Radwaste evaporator tanks and pumps e) Liquid radwaste collection tanks and pumps f) Liquid radwaste surge tanks g) Liquid radwaste sample tanks and pumps h) Reactor water cleanup phase separator and pumps i) Waste sludge phase separator and pumps j) Spent resin tank k) Waste filling and capping station l) Waste liner transfer and storage areas m) Liquid radwaste demineralizer and piping SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-11 n) Waste mixing tanks o) Liquid radwaste filters p) Gaseous radwaste equipment.

12.3.2.2.4 Turbine Building Shielding Design

Radiation shielding is provided around the following equipment in the turbine building to ensure that zone access requirements (Dwgs. A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, and A-518, Sh. 1) are met for the following surrounding areas:

a) Condensate filter demineralizers and piping b) Regeneration waste surge tanks and pumps c) Chemical waste neutralizing tanks and pumps d) Reactor feed pump turbines and piping e) Condensate pumps and piping f) Main condensers and hotwell g) Mechanical vacuum pump h) Recombiners and piping i) Steam packing exhauster j) Condensate demineralizer resin regeneration tanks k) Air ejectors and gland steam condensers l) Feedwater heaters, heater drains, and piping m) Main steam piping n) Steam seal evaporator and drain tank o) Moisture separator and drain tanks p) High pressure and low pressure turbines q) Offgas piping r) Ultrasonic resin cleanser s) Condensate filter vessels t) Condensate filter vessel backwash receiving tank

Areas within most of these shield walls have high radiation levels and limited access.

12.3.2.2.5 Control Room Shielding Design

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-12 "Radiation shielding is provided, as necessary, for the control building and the control structure envelope in order to ensure that the radiation zoning and access requirements as presented on Dwgs. A-511, Sh. 1 and A-512, Sh. 1 are satisfied during normal operation. In addition, shielding is provided to permit access and occupancy of the areas of the control structure envelope in which critical safety functions are performed under post accident conditions with radiation doses limited to 5 rem TEDE from all contributing modes of exposure for the duration of any accident described in Chapter 15 (in accordance with 10CFR50.67).

An isometric drawing of the control and reactor building shielding is provided on Figure 12.3-29 to show the relationship of potential post accident sources to control structure habitability zones. The parameters used in the assessment of control structure envelope habitability during normal and abnormal station operating conditions, including post accident requirements as discussed in Sections 6.4.9.4, and in Appendix 15B."

12.3.2.2.6 Diesel Generator Building Shielding Design

There are no radiation sources in the diesel generator building; therefore, no shielding is required for the building.

12.3.2.2.7 Miscellaneous Plant Areas and Plant Yard Areas

Sufficient shielding and/or radiation protection controls are provided for all plant buildings and designated yard areas containing radiation sources or radioactive material such that radiation levels are maintained within regulatory Units. Some operations, such as loading solidified waste into shield casks, and storage of radiation materials, require access controls. These areas are surrounded by a security fence and closed off from areas accessible to the general public.

12.3.2.2.8 Counting Room Shielding Because the counting room contains sensitive instruments to radioactivity measurements, it is imperative that the background radiation levels are minimized. To accomplish this, no fly ash was used in the concrete mix for the walls and slabs surrounding the counting room. Fly ash normally contains a relatively large amount of slow decaying radioactive isotopes. In addition, the shield walls and slabs were sized to maintain a background radiation level of less than 130 mrem/year for anticipated operational occurrences and 45 mrem/year for normal operation.

12.3.2.3 Shielding Calculational Methods

The shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Subsection 12.2.1. The thickness of each shield wall surrounding radioactive equipment is determined by approximating as closely as possible the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to energy group sources using data from standard Refs. 12.3-1, 12.3-2, 12.3-3, 12.3-4, and 12.3-5.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-13 The geometric model assumed for shielding evaluation of tanks, heat exchangers, filters, demineralizers, and evaporators is a finite cylindrical volume source. For shielding evaluation of piping, the geometric model is a finite shielded cylinder. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source. Typical computer codes that are used for shielding analysis are listed in Table 12.3-2. Shielding attenuation data used in those codes include gamma class attenuation coefficients (Ref. 12.3-6),

gamma buildup factors (Ref. 12.3-7), neutron-gamma multigroup cross sections (Ref. 12.3-20), and albedos (Ref. 12.3-12).

The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area.

Shielding requirements are evaluated at the point of maximum radiation dose through any wall.

Therefore, the actual anticipated radiation levels in the greater region of each plant area is less than this maximum dose and therefore less than the radiation zone upper limit.

Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are designed using a general purpose gamma-ray scattering code G 3 (Ref. 12.3-11). The mazes are constructed so that the scattered dose rate plus the transmitted dose rate through the shield wall from all contributing sources is below the upper limit of the radiation zone specified for each plant area.

2.3.3 VENTILATION

The plant heating, ventilating, and air conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Detailed HVAC system descriptions are provided in Section 9.4. Control Structure habitability is discussed in Section 6.4.

12.3.3.1 Design Objectives The systems are designed to operate such that the in-plant airborne activity levels for normal operation (including anticipated operational occurrences) in the general personnel access areas are within the limits of 10CFR20. The systems operate to reduce the spread of airborne radioactivity during normal and anticipated abnormal operating conditions.

During post-accident conditions, the ventilation system for the plant control room provides a suitable environment for personnel and equipment and ensures continuous occupancy in this area.

The plant ventilation systems are designed to comply with the airborne radioactivity release limits for offsite areas during normal operation.

12.3.3.2 Design Criteria

Design criteria for the plant HVAC systems include the following:

a) During normal operation and anticipated operational occurrences, airborne radioactivity levels to which plant personnel are exposed is ALARA and within the limits specified in 10CFR20.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-14 b) During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10CFR20 and 10CFR50.

c) The plant siting dose guidelines of 10CFR50.67 will be satisfied following those hypothetical accidents, described in Chapter 15, which involve a release of radioactivity from the plant.

d) The dose to control room personnel shall not exceed the limits specified in 10CFR50.67 following those hypothetical accidents, described in Chapter 15, which involve a release of radioactivity from the plant.

12.3.3.3 Design Guidelines In order to accomplish the design objectives, the following guidelines are followed wherever practicable.

12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity

a) Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.

b) Equipment vents and drains are piped directly to a collection device connected to the collection system instead of allowing any contaminated fluid to flow across the floor to the floor drain.

c) All-welded piping systems are used on contaminated systems to the maximum extent practicable to reduce system leakage. If welded piping systems are not used, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.

d) The valves in some systems are provided with leak-off connections piped directly to the collection system.

e) Suitable coatings are applied to the concrete floors of potentially contaminated areas to facilitate decontamination.

f) Where practicable, metal diaphragm or bellows seat valves are used on those systems where essentially no leakage can be tolerated.

g) Contaminated equipment has design features that minimize the potential for airborne contamination during maintenance operations. These features may include flush connections on pump casings for draining and flushing the pump prior to maintenance or flush connections on piping systems that could become highly radioactive.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-15 h) Exhaust hoods are used in plant areas to facilitate processing of radioactive samples by drawing contaminants away from the personnel breathing areas and into the ventilation and filtering systems.

i) Equipment decontamination facilities are ventilated to ensure control of released contamination and minimize personnel exposure and the spread of contamination.

12.3.3.3.2 Guidelines to Control Airborne Radioactivity a) The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination under normal conditions.

b) In building compartments with a potential for contamination, a greater volumetric flow is exhausted from the area than is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.

c) Floor and equipment drain collector tank vents are piped to a collection header and processed by the tank vent filter system.

d) Exhaust air is routed through a prefilter and HEPA filters or a combination of prefilter, HEPA and charcoal filters where necessary before release to the atmosphere to reduce onsite and offsite airborne concentrations.

e) Air is supplied to each principal building via separate supply intakes and duct systems.

f) Redundant, Seismic Category I systems and components are provided for portions of the ventilation system required for safe shutdown of the reactor and to mitigate the consequences of design basis accidents. Included herein are the plant control room ventilation system, the reactor building recirculation system, the standby gas treatment system, and coolers and selected engineered safety feature equipment rooms.

g) Air being discharged from potentially contaminated areas of the Turbine Building and the Reactor Building is passed through prefilters, HEPA filters and charcoal adsorbing filters.

Air being discharged from the Radwaste Building is passed through prefilters and HEPA filters. Means are provided to isolate these areas upon indication of contamination to prevent the discharge of contaminants to the environment.

h) Suitable containment isolation valves are installed in accordance with General Design Criteria 54 and 56, including valve controls, to ensure that the containment integrity is maintained. See additional discussion in Subsections 3.1.2.5.5, 3.1.2.5.7 and 6.2.4.

12.3.3.3.3 Guidelines to Minimize Personnel Exposure from HVAC Equipment

a) Ventilation fans and filters are provided with adequate access space to permit servicing with minimum personnel radiation exposure. The HVAC system is designed to allow rapid replacement of components.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-16 b) Ventilation ducts are designed to minimize the buildup of radioactive contamination within the ducts to the maximum extent practicable. Welded seams are used to join ductwork segments and internal obstructions are avoided wherever practicable.

c) Access and service of ventilation systems in potentially radioactive areas are provided by component location to minimize operator exposure during maintenance, inspection, and testing as follows:

1) The outside air supply units and building exhaust system components are enclosed in ventilation equipment rooms. These equipment rooms are located in radiation Zone II and are accessible to the operators. Work space will be provided around each unit for anticipated maintenance, testing, and inspection. Filter-adsorber units generally comply with the access and service requirements of Regulatory Guide 1.52. (Refer to response to Regulatory Guide 1.52 in Section 3.13.)

Local cooling equipment, servicing the building requirements, will normally be located in areas of low contamination potential radiation Zones I or II.

d) While the majority of the activity in the filter train is removed by simply removing the contaminated filters, further decontamination of the internal structure is facilitated by the proximity of electrical outlets for operation of decontamination equipment, and water supply for washdown of the interior, if necessary. Drains are provided on the filter housing for removal of contaminated water.

e) Active elements of the atmospheric cleanup systems are designed to permit easy removal.

f) Access to active elements is direct from working platforms to simplify element handling. Space is provided on the platforms for accommodating safe personnel movement during replacement of components, including the use of necessary material-handling facilities and during any in-place testing operations.

g) The clear space for doors is a minimum of 3 ft by 7 ft.

h) The filters are designed with replaceable units that are clamped in place against compression seals. The filter housing is designed, tested, and proven to be airtight with bulkhead type doors that are closed against compression seals.

i) Filter systems in which radioactive materials could accumulate to produce significant radiation fields external to the ductwork are appropriately located and shielded to reduce exposure to personnel and equipment.

j) Filters in all systems are changed based upon the airflow and the pressure drop across the filter bank. Charcoal adsorbers are changed based on the residual adsorption capacity of the bed as measured by the testing of carbon samples taken from the removable canisters located in the carbon bed. The testing of the carbon adsorbers and all other components is described in Table 9.4-1.

12.3.3.4 Design Description

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-17 The ventilation systems serving the following structures are assumed to be potentially radioactive and are discussed in detail in Section 9.4.

a) Reactor Building b) Radwaste Building c) Turbine Building Although the control room is considered to be a non-radioactive area, radiation protection is provided to ensure habitability (see Section 6.4).

Ventilation system design parameters for the four systems are given in Tables 9.4-2, 9.4-3, 9.4-6, and 9.4-7.

A typical layout of a potentially radioactive filter unit is given on Figure 12.3-3.

12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION

12.3.4.1 Area Radiation Monitoring

The area radiation monitoring system supplements the personnel and area radiation survey provisions of the plant health physics program described in Section 12.5 to ensure compliance with the personnel radiation protection guidelines of 10CFR20, 10CFR50, 10CFR70, and Regulatory Guides 1.21, 8.2, 8.8 and 8.12 as discussed below.

The area radiation monitors function to:

a) Alert plant personnel of abnormally high radiation levels which, if unnoticed, could result in inadvertent exposures.

b) Inform the control room operator of the occurrence and approximate location of abnormal radiation level increase.

c) Comply with the requirements of 10CFR50 Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.

d) Assist in the detection of unauthorized or inadvertent movement of radioactive material in the plant.

e) Supplement other systems including process radiation monitoring, leak detection, etc., in detecting abnormal migrations of radioactive material in or from the process streams.

The area radiation monitoring system has no function related to the safe shutdown of the plant, or to the quantitative monitoring of the release of radioactive material to the environment.

The combination of the airborne radioactivity monitoring system in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns throughout the plant, plant equipment layout, lack of significant radioactive airborne SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-18 sources in normally occupied areas (radiation Zones I and II), and administrative control of access into applicable radiation and high radiation areas is sufficient to ensure that airborne radioactivity levels are safe in terms of the required duration of personnel access. A general review of these concepts follows:

a) Significant releases of airborne radioactive materials within the plant are detected by effluent monitors as described in Table 11.5-1 in Section 11.5. Air flow patterns are normally from occupied areas to non-occupied areas and from low airborne radioactive material areas to high airborne radioactive material areas. Readouts and alarms are located locally and in the control room.

b) Additional Continuous Air Monitors (CAMS) with local readout and alarm are located in selected areas of potential airborne concentrations throughout the reactor turbine and radwaste building (s).

c) Administrative controls for limiting exposure to airborne radioactivity concentrations greater than 10 Derived Air Concentration (DAC)-HRS. as specified in 10CFR20 Appendix B to Sections 20.1001 - 20.2401, Table 1 Col. 3 are as follows:

1) Routine airborne radioactivity surveys of various accessible radiation zones within the plant. The routine monitoring schedule and frequency is delineated in Station Health Physics Procedures. These locations may be modified with consideration of plant operating status.

2) Access to airborne radioactivity areas with concentrations greater than 30% of the applicable Derived Air Concentration and/or Derived Air Concentration mixture are controlled via a Radiation Work Permit (RWP). Entry into and/or work in the area are preceded by a survey sufficient to determine the radiological conditions present and protection required for these conditions. This information is specified on the RWP. 3) Access to areas where the potential for high radioactive airborne concentration exist due to work conditions is controlled via the RWP process.

12.3.4.1.1 Criteria for Area Monitor Selection The following design criteria are applicable to the area radiation monitoring system.

Energy Dependence: The detector-indicator and trip unit should be responsive to gamma radiation over an energy range of 80 KeV to 6 MeV. The energy dependence should not exceed

+/-20 percent of point for an exposure rate of approximately 50 mr/hr from 100 KeV to 3 MeV and there should be response from 80 KeV to 6 MeV.

Accuracy: The overall accuracy within the design range of temperature, humidity, line voltage, and line frequency variation should be such that the actual reading relative to the true reading, including susceptibility and energy dependence (100 KeV to 3 MeV), should be within 9.5 percent of equivalent linear full scale recorder output for any decade.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-19 Reproducibility: At design center the reading shall be reproducible within

+/-10 percent of point with constant geometry.

ENVIRONMENTAL CONDITIONS PARAMETER SENSOR LOCATION CONTROL ROOM DESIGN CENTER RANGE DESIGN CENTER RANGE Temperature (degrees C) 25 0 to 60 25 5 to 50 Relative Humidity (Percent) 50 20 to 100 50 20 to 90 12.3.4.1.2 Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel unknowingly to receive high radiation doses (e.g., in excess of 10CFR20 limits) in a short period of time because of system failure or improper personnel action. Any plant area that meets one or more of the following criteria is monitored:

a) Zone I areas which, during normal plant operation, including refueling, could exceed the radiation limit of 0.5 mrem/hr upon system failure or personnel error or which will be continuously occupied following an accident requiring plant shutdown b) Zone II areas where personnel could otherwise unknowingly receive high levels of radiation exposure due to system failure or personnel error c) Area monitors are in accordance with General Design Criterion 63 of 10CFR50 Appendix A. 12.3.4.1.3 System Description (Area Radiation Monitoring)

General The area radiation monitoring system is shown in diagram form in Dwgs. M-137, Sh. 1 and M-137, Sh. 2. Each channel consists of a combined sensor/converter unit, a local auxiliary unit (readout with visual and audible alarm), a combined indicator/trip unit, a shared power supply, and a shared multipoint recorder. The exception to this is that the accident range monitors, channels 48 through 57, do not have audible alarms. The location of each area radiation detector is indicated on the shielding and zoning drawings, Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1, and with the exception of the LLRW Holding Facility, is listed in Table 12.3-7. With the exception of channels 11 and 12, the detector locations are the same for both Units 1 and 2. In Unit 1 the channel 11 monitor is located adjacent to the Reactor Building sample station which is located just outside room I-508 and in Unit 2 the channel 11 monitor is located adjacent to the Reactor Building sample station which is inside SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-20 room II-508. In Unit 1, channel 12 monitors the spent fuel hoistway during transfer operations.

This hoistway does not exist in Unit 2.

Circuit Description Sensor/Converter: Each sensor/converter contains all silicon semiconductors in sealed enclosure with a Cooke-Yarborough courtyard circuit which combines a long integrating time constant at low radiation levels with fast overall response at high radiation levels.

Auxiliary Unit: Each auxiliary unit gives instant local readout at the sensor location with a visual alarm. An audible alarm is connected to the auxiliary unit to alert personnel of excessive area radiation.

Indicator and Trip Unit: The indicator and trip unit provides channel control for the area radiation monitoring system. Its circuitry provides an upscale trip that indicates high radiation and a downscale trip that may indicate instrument trouble or loss of power. The module has an analog readout, a low and high trip indicating light, a trip test device, an alarm reset and an output for a multipoint recorder.

Ranges and Sensitivity: Ranges and sensitivities are selected for each location based on the anticipated radiation level as provided by experimental measurements of levels in similar plants and shielding calculations. Refer to Table 12.3-7 for detail. Additional range (10 2 - 10 6 mR) was added for Licensing Commitment to Regulatory Guide 1.97, Rev. 2.

Accuracy: The overall accuracy is such that the actual reading relative to the true reading is within

+/-7.5 percent of equivalent full scale.

12.3.4.1.4 Area Radiation Monitoring Instrumentation

Power Sources: The power source for the area radiation monitoring system is the 120V AC instrument bus and local lighting panels. The area radiation monitor instrumentation is powered by a high and low voltage electrically regulated power supply capable of handling up to 10 channels.

The system has no emergency power supply.

Alarm Setpoints: Alarm setpoints may vary depending on operational considerations and will be determined by measured radiation levels in accordance with controlled station procedures.

Recording Devices: Two multipoint recorders are located in the control room for recording channels pertaining to Unit 1, Unit 2, and channels which are common to both units. This data is also stored in computer history files and can be retrieved and printed using the PMS Historical Recording service program.

Location of Devices: Refer to details in Table 12.3-7.

Readouts and Alarms: Readouts, visual and audible alarms are provided locally for each monitoring channel. The accident range monitors, Channels 48 through 57, serve as indicating channels only and do not have audible alarms. The normal range channels in the same locations serve as the alarm monitors. Readouts and visual alarms are provided by each indicator/trip unit in the Control Structure (Upper Relay Room). Multipoint recorders, visual alarms and PMS displays SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-21 are provided in the Control Structure (Control Room), with the exception of the three Technical Support Center channels (43, 44, 45). The following annunciators are located in the main control room to alert the operator:

a) Reactor Building Area High Radiation (Units 1 and 2)

b) Turbine Building Area High Radiation (Units 1 and 2)

c) Radwaste Building Area High Radiation

d) Refueling Floor Area High Radiation (Units 1 and 2)

e) Spent Fuel Pool Area High Radiation (Units 1 and 2)

f) Reactor Building Common Area High Radiation

g) Administration Building Area High Radiation

h) Control Structure Area High Radiation i) Area Radiation Monitoring Downscale (ganged for all channels)

12.3.4.1.5 Safety Evaluation The area radiation monitoring system is designed to operate unattended for extended periods and is designed for high reliability. Failure of one monitor has no effect on any other.

The system is not essential for safe shutdown of the plant, and serves no active emergency function during operation. The system is not safety related and is constructed to Quality Group D Requirements.

12.3.4.1.6 Calibration Method and Testability Facilities for calibrating these monitor units are provided, which include a test unit designed for use in the adjustment procedure for the area radiation monitor sensor and converter unit. These provide several gamma radiation levels between 1 and 250 mrem/hr.

A cavity in the calibration unit receives the sensor and converter unit. A window through which radiation from the source emanates is located on the back wall of the cylindrical lower half of the cavity. A chart on each calibration unit indicates the radiation levels available from the unit for the various control settings.

An internal trip test circuit, adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip unit input so that a meter reading is provided in addition to a real trip. All trip circuits are the latching type and must be manually reset in the Upper Relay Room.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-22 The radiation monitors will be calibrated at regular time intervals in accordance with station procedures.

12.3.4.2 Airborne Radioactivity Monitoring

Refer to Subsections 12.5.2.2.3 and 12.5.3.5.4 for information on air borne radioactivity monitoring.

12.

3.5 REFERENCES

12.3-1 J. J. Martin and P. H. Blichert-Toft, Nuclear Data Tables "Radioactive Atoms, Auger Electrons, and X-Ray Data," Academic Press, October, 1970.

12.3-2 J. J. Martin, Radioactive Atoms Supplement 1, ORNL 4923, August, 1973.

12.3-3 W. W. Bowman and K. W. MacMurdo, Atomic Data and Nuclear Data Tables, "Radioactive Decay's Ordered by Energy and Nuclide," Academic Press, February, 1970.

12.3-4 M. E. Meek and R. S. Gilbert, "Summary of X-Ray and Gamma Energy and Intensity Data," NEDO-12037, January, 1970.

12.3-5 C. M. Lederer, et al, Table of Isotopes, Lawrence Radiation Laboratory, University of California, March, 1968.

12.3-6 G. W. Goldstein, X-Ray Attenuation Coefficients from 10 KeV to 100 MeV, National Bureau of Standards Circular 583 (Issued April 30, 1957).

12.3-7 D. K. Trubey, "A Survey of Empirical Functions Used to Fit Gamma-Ray Buildup Factors," ORNL-RSIC-10, February, 1966.

12.3-8 W. W. Engle, Jr., "A User's Manual for ANISN: A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," Union Carbide Corporation, Report No. K-1693, 1967.

12.3-9 R. E. Malenfant, QAD, A Series of Point-Kernel General-Purpose Shielding Programs, Los Alamos Scientific Laboratory, LA 3573, October, 1966.

12.3-10 D. Arnold and B. F. Maskewitz, "SDC, A Shielding-Design Calculation for Fuel-Handling Facilities," ORNL-3041, March, 1966.

12.3-11 R. E. Malenfant, "G 3: A General Purpose Gamma-Ray Scattering Program," Los Alamos Scientific Laboratory, LA 5176, June, 1973.

12.3-12 W. E. Selph, "Neutron and Gamma Ray Albedos," ORNL-RSIC-21, February, 1968.

12.3-13 D. S. Duncan and A. B. Spear, Grace I - An IBM 704-709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields, Atomics International (June, 1959).

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-23 12.3-14 D. S. Duncan and A. B. Spear, Grace II - An IBM 709 Program for Computing Gamma Ray Attenuation and Heating in Cylindrical and Spherical Geometries, Atomics International (November, 1959).

12.3-15 D. A. Klopp, NAP - Multigroup Time-Dependent Neutron Activation Prediction Code , IITRI-A6088-21 (January, 1966).

12.3-16 E. A. Straker, P. N. Stevens, D. C. Irving, and V. R. Cain, MORSE - A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code, ORNL-4585 (September, 1970).

12.3-17 W. A. Rhoades and F. R. Mynatt, The DOT III Two-Dimensional Discrete Ordinates Transport Code, ORNL-TM-4280 (1973).

12.3-18 U.S. Nuclear Regulatory Commission, Regulatory Guide 8.8 (July, 1973).

12.3-19 M. J. Bell, "ORIGEN - The ORNL Generation and Depletion Code," Oak Ridge National Laboratory, ORNL-4628 (May, 1973).

12.3-20 ORNL RSIC Computer Code Collection DLC-23, CASK -

40 Group Neutron and Gamma-Ray Cross Section Data.

12.3-21 R. G. Jaeger, et al, Engineering Compendium on Radiation Shielding, Volume I, Springer - Verlag, New York Inc., 1968.

12.3-22 C. A. Negin and G. Worku, Microshield , Grove Engineering, Inc., Rockville, Md.

12.3-23 MICRO Skyshine, Framatome Technologies, Inc., d.b.a. Grove Engineering, Rockville, MD.

12.3-24 Cobalt Reduction Guidelines, Electric Power Research Institute, Palo Alto, CA, March 1990, NP-6737.

THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 12.3-7, Rev. 48 AutoCAD Figure 12_3_7doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-511, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-8 replaced by dwg.

A-511, Sh. 1 FIGURE 12.3-8, Rev. 55 AutoCAD Figure 12_3_8.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-512, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-9 replaced by dwg.

A-512, Sh. 1 FIGURE 12.3-9, Rev. 49 AutoCAD Figure 12_3_9.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-513, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-10 replaced by dwg.

A-513, Sh. 1 FIGURE 12.3-10, Rev. 57 AutoCAD Figure 12_3_10.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-514, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-11 replaced by dwg.

A-514, Sh. 1 FIGURE 12.3-11, Rev. 56 AutoCAD Figure 12_3_11.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-515, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-12 replaced by dwg.

A-515, Sh. 1 FIGURE 12.3-12, Rev. 57 AutoCAD Figure 12_3_12.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-516, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-13 replaced by dwg.

A-516, Sh. 1 FIGURE 12.3-13, Rev. 56 AutoCAD Figure 12_3_13.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-517, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-14 replaced by dwg.

A-517, Sh. 1 FIGURE 12.3-14, Rev. 57 AutoCAD Figure 12_3_14.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-518, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-15 replaced by dwg.

A-518, Sh. 1 FIGURE 12.3-15, Rev. 57 AutoCAD Figure 12_3_15.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-519, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-16 replaced by dwg.

A-519, Sh. 1 FIGURE 12.3-16, Rev. 56 AutoCAD Figure 12_3_16.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-520, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-17 replaced by dwg.

A-520, Sh. 1 FIGURE 12.3-17, Rev. 56 AutoCAD Figure 12_3_17.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-521, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-18 replaced by dwg.

A-521, Sh. 1 FIGURE 12.3-18, Rev. 56 AutoCAD Figure 12_3_18.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-522, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-19 replaced by dwg.

A-522, Sh. 1 FIGURE 12.3-19, Rev. 57 AutoCAD Figure 12_3_19.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-523, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-20 replaced by dwg.

A-523, Sh. 1 FIGURE 12.3-20, Rev. 57 AutoCAD Figure 12_3_20.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-524, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-21 replaced by dwg.

A-524, Sh. 1 FIGURE 12.3-21, Rev. 57 AutoCAD Figure 12_3_21.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-525, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-22 replaced by dwg.

A-525, Sh. 1 FIGURE 12.3-22, Rev. 56 AutoCAD Figure 12_3_22.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-526, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-23 replaced by dwg.

A-526, Sh. 1 FIGURE 12.3-23, Rev. 57 AutoCAD Figure 12_3_23.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-527, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-24 replaced by dwg.

A-527, Sh. 1 FIGURE 12.3-24, Rev. 57 AutoCAD Figure 12_3_24.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-528, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-25 replaced by dwg.

A-528, Sh. 1 FIGURE 12.3-25, Rev. 57 AutoCAD Figure 12_3_25.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-529, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-26 replaced by dwg.

A-529, Sh. 1 FIGURE 12.3-26, Rev. 57 AutoCAD Figure 12_3_26.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-530, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-27 replaced by dwg.

A-530, Sh. 1 FIGURE 12.3-27, Rev. 48 AutoCAD Figure 12_3_27.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

BIN-100 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-28 replaced by dwg.

BIN-100 FIGURE 12.3-28, Rev. 55 AutoCAD Figure 12_3_28.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-137, Sh. 1FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-30-1 replaced by dwg.

M-137, Sh.

1FIGURE 12.3-30-1, Rev. 56 AutoCAD Figure 12_3_30_1.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-137, Sh. 2FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-30-2 replaced by dwg.

M-137, Sh.

2FIGURE 12.3-30-2, Rev. 56 AutoCAD Figure 12_3_30_2.doc

Radiation Protection Manager PC-F12.5-1.vsd FSAR Rev. 68 Figure Rev. 58 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2FINAL SAFETY ANALYSIS REPORT HEALTH PHYSICS ORGANIZATION FSAR FIGURE 12.5-1Radiological Operations SupervisorProgramsRadiological Operations SupervisorOperationsTechnical Staff

- RAM Shipping

- Dosimetry

- InstrumentsHP ForemenHP TechniciansRadiologicalSupervisor__________

ALARARadiologicalOperationsSupervisor___________

ALARA ALARA Staff - Radiological Work Schedulers

- ALARA Specialists

- Radiological Engineer SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA)

12.1.1 POLICY CONSIDERATIONS

12.1.1.1 Management Policy It is the policy of PP&L to maintain occupational radiation exposure As Low As Reasonably Achievable (ALARA) at the Susquehanna SES. This includes maintaining the annual dose to individuals working at the station ALARA, and keeping the annual integrated dose to station personnel ALARA. The management of this Company is firmly committed to performing all reasonable actions to ensure that radiation exposures are maintained ALARA.

Subsection 12.1.2 and Section 12.3 discuss the ALARA considerations that have been incorporated into the design of the Susquehanna SES.

Susquehanna SES will be operated and maintained in such a manner as to ensure occupational radiation exposures (ORE) are ALARA. The operational ALARA program is described in Section 12.5. Training programs will be established to assure personnel understand both why and how occupational radiation exposures will be maintained ALARA. A Station ALARA Committee has been established to ensure implementation of ALARA policy by various program

reviews.

12.1.1.2 Management Responsibilities

Figures 17.2-2 and 13.1-2 exhibit the management organizational structure for the

Susquehanna SES.

The Vice President-Nuclear Operations has the corporate responsibility for the ALARA program. The responsibility for the coordination and administration of the ALARA program is assigned to the General Manager-Nuclear Engineering and the General Manager-Susquehanna SES and their reports. They are responsible to ensure the policies and commitments contained in the PP&L ALARA Program are being properly implemented and maintained.

During the design and construction phase, the Susquehanna SES Project Manager is responsible to ensure that the design and construction of the facility is such that occupational exposures will be ALARA. This will include ensuring that, to the extent practicable:

a. Design concepts and station features reflect consideration of the activities of station personnel that might be anticipated and that might lead to personnel exposure to substantial sources of radiation and that station design features have been provided to reduce the anticipated exposures of station personnel to these sources of radiation.

b. Specifications for equipment reflect the objectives of ALARA, including among others, considerations of reliability, serviceability and limitations of internal accumulations of radioactive material.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-2 During the operational phase, the General Manager-Nuclear Engineering is responsible to ensure that the Station design remains in compliance with all applicable radiation safety standards found in 10CFR20, 10CFR50, 10CFR50.67, 40CFR190 and applicable regulatory guidance documents.

The Manager-Nuclear Modifications is responsible for ensuring the PP&L ALARA program is incorporated into the design of plant modifications and new facilities related to the SSES.

The Manager-Nuclear Technology is responsible to ensure that a radiation protection staff with health physics and radiological expertise is adequately maintained to support Nuclear Engineering, Susquehanna SES plant staff and other functional group activities as appropriate.

During the startup and operation phase, the General Manager-Susquehanna SES is responsible for ensuring radiation exposure is controlled in a manner consistent with ALARA requirements and is specifically responsible for the onsite radiation protection program. Additionally he is responsible for:

a. Ensuring support from all station personnel for the implementation of the Station ALARA program

b. Providing management oversight of the accumulation of personnel exposure at SSES and reviewing and concurring with annual personnel exposure goals

c. Ensuring resources needed to achieve ALARA goals and objectives are made available.

Additional General Manager-Susquehanna SES responsibilities are implemented through the Health Physics Supervisor.

Major ALARA responsibilities of the Health Physics Supervisor or designee, include the

following:

a. Participating in reviews of design changes for facilities and equipment that can affect potential radiation exposures;

b. Identifying locations, operations, and conditions, that have the potential for causing significant exposures to radiation;

c. Initiating and implementing an exposure control program which includes the establishment of manrem goals,

d. Developing plans, procedures, and methods for keeping radiation exposures of station personnel ALARA,

e. Reviewing, commenting on, and recommending changes in applicable procedures to maintain exposures ALARA;

f. Developing or participating in the development of appropriate Health Physics training programs related to work in radiation areas or involving radioactive material;

g. Supervising the radiation surveillance program to maintain data on exposures of and doses to station personnel by specific job functions and type of work; SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-3 h. Supervising the collection, analysis, and evaluation of data and information attained from radiological surveys and monitoring activities;

i. Supervising, training, and qualifying the radiation protection staff of the station; and

j. Ensuring that adequate radiation protection coverage is provided for station personnel during all working hours.

Chapter 13 provides additional information concerning responsibilities and reporting relationships at the Susquehanna SES.

12.1.1.3 Policy Implementation The management ALARA policy is implemented at the Susquehanna SES by the Health Physics Staff under the direction of the General Manager-Susquehanna SES and the Health Physics Supervisor. The policy implementation is formalized by the incorporation of ALARA philosophy and considerations into permanent plant procedures dealing specifically with ALARA concerns. The operational ALARA considerations identified in Subsections 12.1.3 and 12.5.3.2 are implemented by these procedures.

Subsection 12.5.3.7 describes the training program established to give appropriate station personnel the necessary knowledge to understand why and how they should maintain their ORE ALARA.

The Station ALARA Committee has been established to review the implementation of the Company ALARA Program. Specific responsibilities of the Station ALARA Committee include:

a. Assuring the effectiveness of the ALARA program as implemented at the Susquehanna SES.

b. Assuring that high exposure maintenance and modification tasks receive proper management attention ensuring they are planned in accordance with sound ALARA principles.

c. Reviewing, prioritizing and recommending potential action items for inclusion into the Nuclear Department long-term exposure reduction plan and monitoring status of action items included in the plan.

d. Administering the Employee ALARA Concerns program.

e. Assuring Station activities are conducted in an ALARA manner maintaining the balance between cost, schedule and personnel exposure.

f. Reviewing exposure goals, monitoring performance against these goals and taking action as appropriate when goals are jeopardized.

g. Monitoring individual personnel exposures to ensure they are minimized to the extent possible while maintaining overall collective exposures ALARA.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-4 12.1.2 DESIGN CONSIDERATIONS This subsection discusses the methods and features by which the policy considerations of Subsection 12.1.1 are applied. Provisions and designs for maintaining personnel exposures as low as reasonably achievable are presented in Subsections 12.3.1, 12.3.2 and 12.5.3.

Experiences and data from operating plants are evaluated to decide if and how equipment or facility designs could be improved to reduce overall plant personnel exposures. During plant design, operating reports and data such as that given in WASH 1311, NUREG-75/032, NUREG-109 and Compilation and Analysis of Data on occupational Radiation Exposure Experienced at Operating Nuclear Power Plants, AIF, September 1974, References 12.1-1, through 12.1-4 respectively, were reviewed to determine which operations, procedures or types of equipment were most significant in producing personnel exposures. Methods to mitigate such exposures were implemented wherever possible and practicable.

General design considerations and methods employed to keep in-plant radiation exposures ALARA have two objectives:

a) Minimizing the necessity for the amount of personnel time spent to radiation areas; and b) Minimizing radiation levels in routinely occupied plant areas and in the vicinity of plant equipment expected to require personnel attention.

Both equipment and facility designs are considered in keeping exposures ALARA during plant operations including normal operation, maintenance and repairs, refueling operations and fuel storage, in-service inspection and calibrations, radioactive waste handling and disposal, and other events of moderate frequency. The actual design features used are described in

Subsection 12.3.1.

12.1.2.2 Equipment General Design Considerations for ALARA The following equipment general design considerations to minimize the necessity for and amount of personnel time spent in a radiation area include, where practicable:

a) Reliability, durability, construction, and design features of equipment, components, and materials to reduce or eliminate the need for repair or preventive maintenance;

b) Servicing convenience including ease of disassembly and modular ization of components for replacement or removal to a lower radiation area for repair; c) Provisions, where practicable, to remotely or mechanically operate, repair, service, monitor, or inspect equipment; and d) Redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high and when no feasible method is available to reduce

radiation levels.

The following equipment general design considerations directed toward minimizing radiation levels proximate to equipment or components requiring personnel attention include, where practicable:

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-5 a) Provision for draining, flushing, or, if necessary, remote cleaning of equipment containing radioactive material;

b) Design of equipment, to minimize the buildup of radioactive material and to facilitate flushing of crud traps;

c) Utilization of high quality valves, valve packings, and gaskets to minimize leakage and spillage of radioactive materials;

d) Provisions for minimizing the spread of contamination into equipment service areas; and

e) Provisions for isolating equipment from radioactive process fluids.

12.1.2.3 Facility Layout General Design Considerations for ALARA Facility general design considerations to minimize the amount of personnel time spent in radiation areas include where practicable:

a) Locating equipment and instruments, which will require routine maintenance, calibration, or inspection for ease of access and a minimum of required occupancy time in radiation

fields; b) Arranging plant areas to allow remote or mechanical operation, service, monitoring, or inspection of highly radioactive equipment; and

c) Providing, for transportation of equipment or components requiring service to a lower radiation area.

Facility general design considerations directed toward minimizing radiation levels in plant access areas and in the vicinity of equipment requiring personnel attention include, where practicable:

a) Separating radiation sources and occupied areas (e.g., pipes containing potentially highly radioactive fluids do not pass through normally occupied areas);

b) Providing adequate shielding between radiation sources and access and service areas;

c) Locating appropriate equipment, instruments, and sampling sites in the lowest practicable radiation zone;

d) Providing means and adequate space for using movable shielding for sources within the service area when required; and

e) Providing means (e.g., curbing, drains and flush) to control contamination and to facilitate decontamination of potentially contaminated areas.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-6 12.1.2.4 ALARA Design Review During the design phase, Bechtel Power Corporation, as agents for PP&L, were given the basic responsibility for the performance of the ALARA design review. PP&L provided overall coordination of and input to this review. ALARA design reviews were completed on all required systems and areas. Recommended design modifications were made. In addition to intensive system/area ALARA design reviews, field routed small piping drawings were continually reviewed, often resulting in changes in routing, valve and operator types, and connection points.

During the operational phase, ALARA considerations are included in the scoping and design phase of modifications and changes to the design of the facility. The General Manager-Nuclear Engineering has the responsibility to:

a. Ensure that engineering personnel are adequately trained in ALARA design and engineering principles so that radiation exposure with respect to installation, operation and maintenance of plant modifications and new facilities is considered in each design.

b. Ensure an integrated level of involvement within the modifications organizations to effectively implement the PP&L ALARA program.

The Manager-Nuclear Modifications is responsible to:

a. Ensure ALARA/dose reduction opportunities are identified for inclusion into the design of modifications and new facilities.

b. Ensure the PP&L ALARA program is incorporated into the design of plant modifications and new facilities related to the Susquehanna SES.

The Supervisor-Operations Technology has the responsibility to determine the requirements of the ALARA design process and, in concert with the Health Physics Supervisor, ensure that ALARA design reviews are performed by qualified radiological personnel.

At the scoping phase, potential radiological impacts such as occupational and offsite exposure impacts, and radioactive waste generation are identified and considered during the design phase. Radiological personnel evaluate the proposed action and provide input into the scoping document. During the design phase, ALARA considerations consistent with the guidance contained in Regulatory Guidance 8.8 is ensured via the use of an ALARA design checklist. This checklist is completed by the design engineer and reviewed and approved by designated radiological personnel.

Procedures have been developed describing the training programs and requirements for personnel involved in the ALARA design process.

12.1.3 OPERATIONAL CONSIDERATIONS

To assure that the occupational radiation exposures are maintained as low as reasonably achievable (ALARA) during the operation of Susquehanna SES specific activities will be implemented.

SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-7 12.1.3.1 Procedure Development Station procedures will be prepared, reviewed, and approved in accordance with Section 13.5.

12.1.3.1.1 ALARA Procedures

To assure adequate emphasis on the necessity to minimize personnel exposures, ALARA procedures will be prepared. These procedures implement considerations of such topics as ALARA review of applicable Radiation Work Permits (RWP), worker feedback, special task training and evaluation of proposed changes in applicable facilities or equipment.

12.1.3.1.2 Station Procedures

Administrative requirements will be implemented to assure that applicable procedures developed by other plant disciplines have adequately incorporated the principle of minimizing personnel exposure. Station administrative documents will describe the criteria for selection of those procedures and revisions that will be reviewed by Health Physics. Recommendations made by Health Physics will normally be resolved with the appropriate plant discipline prior to submission for final review and approval.

12.1.3.2 Station Organization As described in Subsection 12.5.1, the Station organization provides the Health Physics Supervisor direct access to the Vice President-Nuclear Operations to assure uniform support of Health Physics and ALARA requirements. This organization will allow the Vice President-Nuclear Operations direct involvement in the review and approval of specific ALARA goals and objectives as well as review of data and dissemination of information related to the ALARA program.

The organization also provides a Health Physics Specialist-ALARA who is normally free from routine Health Physics activities to implement the Station ALARA program. This individual is primarily responsible for coordination of Station ALARA activities and will routinely interface with first line supervision in radiation work planning and post job review.

12.1.3.3 Operating Experience

The Radiation Work Permit process described in Subsection 12.5.3.2 will provide a mechanism for collection and evaluation of data relating to personnel exposure. Information collated by systems and/or components and job function will assist in evaluating design or procedure changes intended to minimize future radiation exposures.

12.1.3.4 Exposure Reduction

Specific exposure reduction techniques that will be employed at Susquehanna SES are described in Subsection 12.5.3.2. Procedures will assure that applicable station activities are completed with adequate preparation and planning; work is performed with appropriate Health SSES-FSAR Table Rev. 54 FSAR Rev. 64 12.1-8 Physics recommendations and support; and results of post job data evaluation are applied to implement improvements.

In addition, the Health Physics staff, will at all times be vigilant for ways to reduce exposures by soliciting employee suggestions, evaluating origins of plant exposures, investigating unusual exposures, and assuring that adequate supplies and instrumentation are available.

PP&L management will perform periodic reviews of station programs to assure workers are receiving adequate instruction in ALARA and Health Physics requirements. Implementation of the Health Physics program, selected procedures, and past exposure records will also be reviewed. Management will perform formal re views of the Susquehanna SES Health Physics program at least once every three years and results will be forwarded to the Vice

President-Nuclear Operations and appropriate members of corporate management. The results of management reviews may also include recommendations on mechanisms which may reduce personnel exposure. The Vice President-Nuclear Operations will respond to noted recommendations or deficiencies and corrective action or improvements will be verified during subsequent reviews.

12.

1.4 REFERENCES

12.1-1 T. D. Murphy, WASH-1311, UC-78, "A Compilation of Occupational Radiation Exposure from Light Water Cooled Nuclear Power Plants 1969-1973," USNRC Radiological Assessment Branch, May 1974.

12.1-2 T. D. Murphy, et. al., NUREG-75/032, "Occupational Radiation Exposure at Light Water Cooled Power Reactors 1969-1974," USNRC Radiological Assessment Branch, June 1975.

12.1-3 T. D. Murphy, et. al., NUREG-0109, "Occupational Radiation Exposure at Light Water Cooled Power Reactors 1969-1975," USNRC Radiological Assessment

Branch, August 1976.

12.1-4 C. A. Pelletier, et. al., National Environmental Studies Project, "Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants," Atomic Industrial Forum, September 1974.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-1 12.2 RADIATION SOURCES In this section the sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity required for the design of personnel protective measures and for dose assessment are discussed and identified.

12.2.1 CONTAINED SOURCES The shielding design source terms are based on a noble gas fission product release rate of 0.1 Ci/sec (after 30 minutes decay) and the corresponding fission, activation, and corrosion product concentrations in the primary coolant. The sources in the primary coolant are discussed in Section 11.1 and listed in Tables 11.1-1 through 11.1-5. Throughout most of the primary coolant system, activation products, principally nitrogen-16, are the primary radiation sources for shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while at the same time providing for daughter product formation.

Basic reactor data and core region description used for this section are listed in Tables 12.2-1 through 12.2-5.

The data contained in these Tables is the original design basis for the plant and have not been revised for power uprate. Although doses from the reactor core will increase due to power uprate, these increases have no effect on normal operating doses to plant personnel or on equipment qualification. The increased doses from the core resulting from power uprate meet radiation shielding and zoning requirements inside containment.

In this subsection the design sources are presented by building location and system. General locations of the equipment discussed in this section are shown on the shielding and zoning drawings, Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, and A-530, Sh. 1. The layout of equipment in Unit 2 is not significantly different from the layout in Unit 1, therefore the shielding and zoning requirements shown on these drawings are applicable to both units. Detailed data on source descriptions for each shielded plant area are presented in Tables 12.2-38 through 12.2-40.

Shielding source terms presented in this section and associated tables are based on conservative assumptions regarding system and equipment operations and characteristics to provide reasonably conservative radioactivity concentrations for shielding design. Therefore, SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-2 the shielding source terms are not intended to approximate the actual system design radioactivity concentrations.

12.2.1.1 Drywell 12.2.1.1.1 Reactor Core The primary radiations within the drywell during full power operation are neutron and gamma radiation resulting from the fission process in the core. Tables 12.2-4 and 12.2-5 list the multigroup neutron and gamma ray fluxes at the outside surfaces of the reactor pressure vessel and the primary shield at the core midplane. The gamma fluxes include those resulting from capture or inelastic scattering of neutrons within the reactor pressure vessel and core shroud and the gamma radiation resulting from prompt fission and fission product decay.

The data contained in Tables 12.2-4 and 12.2-5 is the original design basis for the plant and have not been revised for power uprate. Although doses from the reactor core will increase due to power uprate, these increases have no effect on normal operating doses to plant personnel or on equipment qualification. The increased doses from the core resulting from power uprate meet radiation shielding and zoning requirements inside containment.

The largest radiation sources after reactor shutdown are the decaying fission products in the fuel. Tables 12.2-9A1 and 12.2-9A2 list the core gamma sources as a function of shutdown time. Secondary sources are the structural material activation of the RPV, its internals, and the piping and equipment located in the primary containment and also the activated corrosion products accumulated or deposited in the internals of the RPV, the primary coolant piping, and other process system piping in the primary containment.

12.2.1.1.2 Reactor Coolant System Sources of radiation in the reactor coolant system are fission products estimated to be released from fuel and activation and corrosion products that are circulated in the reactor coolant. These sources are listed in Tables 11.1-1 through 11.1-5 and their bases are discussed in Section 11.1. The nitrogen-16 concentration in the reactor coolant is assumed to be 50

µCi/gm of coolant at the reactor recirculation outlet nozzle.

12.2.1.1.3 Primary Steam System Radiation sources in the primary steam system piping include activation gases, principally nitrogen-16, and the corrosion and fission products carried over to the steam system.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-3 The nitrogen-16 concentration in the main steam is assumed to be 250

µCi/gm of steam leaving the reactor vessel at the main steam outlet nozzle. Fission product activity corresponds to an offgas release rate of 100,000

µCi/sec at 30 minutes delay from the reactor steam nozzle. Partition fractions for activity into the steam system are 100 percent for gases, 8percent by weight for halogens, and 0.1 percent by weight for particulates. These partition factors are applied to the reactor water concentrations as given in Table 11.1-2 through 11.1-5.

12.2.1.2 Reactor Building

12.2.1.2.1 Reactor Water Cleanup System

Radiation sources in the RWCU system consist of those radioisotopes carried in the reactor water. Nitrogen-16 is the predominant radiation source in the regenerative and nonregenerative heat exchangers and RWCU pumps and piping. The inventory of N-16 is based upon component transit times, as shown in Table 12.2-6. The main sources for the RWCU filter demineralizers, holding pumps, and the RWCU backwash receiving tank are the accumulated corrosion and fission products, based on the inlet reactor water concentrations given in Section 11.1. Table 12.2-7 provides the inventory of the accumulated isotopes in the filter

demineralizer, and Table 12.2-8 provides the inventory of isotopes in the RWCU backwash receiving tank.

12.2.1.2.2 Spent Fuel Handling and Transfer

The spent fuel assemblies are the predominant source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish near fission product buildup equilibrium for the reactor at rated power is used in determining the source strength. Shielding requirements for spent fuel transfer are based on the fission product activity present 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown. Source terms for spent fuel are discussed in Subsection 12.2.1.3.1 and are listed in Tables 12.2-9A1 and 12.2-9A2.

12.2.1.2.3 Residual Heat Removal System

The pumps, heat exchangers, and associated piping of the Residual Heat Removal (RHR)

System are potential carriers of radioactive materials. For plant shutdown, the RHR pumps and heat exchanger sources result from the radioactive isotopes carried in the reactor coolant, discussed in Subsection 12.2.1.1.2, after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown. The radioactive isotopic concentrations are listed in Table 12.2-10.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-4 12.2.1.2.4 Reactor Core Isolation Cooling System Components of the Reactor Core Isolation Cooling (RCIC) System that are potential radiation sources are the RCIC turbine and steam inlet and exhaust piping. Radioactivity in the turbine and piping is that present in the driving steam that has been extracted from the main steam system. The steam activity as discussed in Subsection 12.2.1.1.3, decayed for the appropriate transit time to the RCIC turbine, is used for the shielding calculations for this system, and is listed in Table 12.2-11.

12.2.1.2.5 High Pressure Coolant Injection System The radiation sources for the High Pressure Coolant Injection System are the HPCI turbine and the steam inlet and exhaust piping. The steam activity, as discussed in Subsections 12.2.1.1.3, decayed for the appropriate transit times is used for the shielding of this system as shown in Table 12.2-11.

12.2.1.2.6 Core Spray Systems Because the core spray, when testing, uses condensate from the condensate storage tank with very low radioactivity concentrations, no shielding is required.

12.2.1.3 Refueling Facilities 12.2.1.3.1 Spent Fuel Storage and Transfer The predominant radiation sources in the spent fuel storage and transfer areas are the spent fuel assemblies. Spent fuel assembly sources are discussed in Subsection 12.2.1.2.2. For shielding design, the spent fuel pool is assumed to contain the design maximum of 2840 fuel assemblies (Section 9.1) with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay. The activity of fission product radionuclides which serve as the basis for the spent fuel source term are shown in Tables 12.2-9A1 and 12.2-9A2.

Spent fuel assemblies are also a radiation source at the Independent Spent Fuel Storage Installation (ISFSI). For shielding design and dose rate determinations each Dry Shielded Canister (DSC) is assumed to contain 61 fuel assemblies with a minimum 5 year decay cooling in the spent fuel pool. Refer to Section 11.7.3 for ISFSI Source Terms.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-5 12.2.1.3.2 Fuel Pool Cooling and Cleanup System Sources in the Fuel Pool Cooling and Cleanup (FPCC) System are primarily a result of transfer of radioactive isotopes from the reactor coolant into the spent fuel pool during refueling operations. The total gross activity in the Fuel Pool water is assumed to be reactor coolant activities for fission, corrosion, and activation products (Tables 11.1-1 through 11.1-5), normalized to 1.0E-2

µCi/cc for shutdown conditions and 1.00E-04 µ Ci/cc for normal operations. This activity then undergoes subsequent decay and accumulation on the FPCC filter demineralizers (see Table 12.2-13). The FPCC filter demineralizer resins are back washed periodically into a backwash receiving tank. Shielding source terms for the backwash receiving tank are shown in Table 12.2-14.

12.2.1.4 Turbine Building 12.2.1.4.1 Primary Steam and Power Conversion Systems

Radiation sources for piping and equipment which contain primary steam are based on the radioactivity carried over into the steam from the reactor coolant and include fission product gases and halogens, corrosion and fission products, and gaseous activation products as discussed in Subsection 12.2.1.1.3. Steam density variations and the steam transit times through equipment and pipes are factored into the source term evaluation to account for volumetric dilution effects, radiological decay, and daughter product generation.

12.2.1.4.2 Condensate System

The sources in the condensate system are based on decayed main steam activities (Subsection 12.2.1.1.3). Eighty percent of the N-16 and 100 percent of the noble gases are assumed to be removed from the condensate system by the main condenser evacuation system. The gaseous activities are minor in the hotwell and negligible in the remainder of the condensate system. The hotwell has a minimum of three minutes holdup of condensate and therefore N-16 activity at the condenser outlet is negligible. Fission products, activated corrosion products, and the daughter products from the decay of fission product gases in transit through the turbine are the inlet sources to the condensate system.

These sources, as shown in Table 12.2-15, are present in the condensate pumps and piping and accumulate on the condensate filters and demineralizer resins. When spent, the condensate filters and demineralizer resins are discharged to the solid waste management system for packaging and offsite disposal. Table 12.2-16 provides the isotopic inventory utilized as the shielding source terms for the condensate demineralizers. Tables 12.2.46, SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-6 12.2.47, and 12.2.48 provide the shielding source terms for the condensate filter vessel, condensate backwash solution, and the condensate filter backwash receiving tank, respectively.

12.2.1.4.3 Offgas System Recombiner

Radioactive sources in the gas treatment system originate with the noble gases and non-condensible gases removed from the main condenser, and the activity entering with the

extraction driving steam to the main condenser evacuation system. The activity removed from the main condenser is based on the primary steam activity as described in Subsection 12.2.1.1.3, decayed for the total transit time to the steam jet air ejector. Eighty percent of the N-16 and 100 percent of the noble gases are assumed to be removed by the air ejector.

Activity in the extraction driving steam to the air ejector is the primary steam activity as described in Subsection 12.2.1.1.3, decayed by the transit time to the air ejector. The total quantity of activity in the offgas pipe and recombiner and source term assumptions are shown in Tables 12.2-17 and 12.2-18.

12.2.1.5 Radwaste Building

12.2.1.5.1 Liquid Waste Management System

The liquid waste management system (LWMS) sources are radioisotopes, including fission and activation products, present in the reactor coolant. The components of the LWMS contain varying degrees of radioactivity depending on the detailed system and equipment design.

The concentrations of radionuclides present in the process fluids at various locations in the system such as pipes, tanks, filters, and demineralizers are discussed in Section 11.2 and are listed in Tables 11.2-5 through 11.2-7. Shielding and associated radiation zoning for components of the LWMS are based on the design basis radioactivity concentrations given in Sections 11.1 and 11.2.

12.2.1.5.2 Solid Waste Management System

Wet and dry radioactive wastes are collected, treated, and stored in the solid radwaste facilities as discussed in Section 11.4. The radioactive wastes are processed by filtration, decanting, and ion-exchange treatment. The resultant volume reduced products (e.g., filter cakes, depleted resins) are dewatered for storage and offsite shipment. Processed liquids may be analyzed for reuse as condensate make-up, processed as radioactive waste, or diluted and discharged.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-7 Radioactive waste may be solidified in containers specified in the Solid Radioactive Waste Process Control Program, then washed to minimize external surface contaminants, and shipped or stored in concrete shielded compartments. The aforementioned operations may be accomplished utilizing remote container loading, transfer, capping facilities, and an overhead crane. Shielding and radiation zone designations of the solid radioactive waste areas are based on the maximum activity sources at zero decay presented in Tables 11.2-6 and 11.4-6, without any consideration of external container shielding credit.

12.2.1.5.3 Ambient Charcoal Offgas Treatment System

The charcoal offgas system as described in Section 11.3 is located in the radwaste building and primarily adsorbs the noble gases and daughter products remaining in the noncondensible gases removed from the main condenser after treatment in the recombiner offgas system.

The shielding of the components is based on the transit times for formation and accumulation of noble gas daughter products and the remaining xenon and krypton gases on the carbon beds.

The gases, after charcoal treatment, pass th rough a post HEPA filter where remaining particulates are trapped prior to exhausting. The concentration of the activity on the piping, equipment, and particulate and charcoal filters for shield design is shown in Tables 12.2-19

through 12.2-24.

12.2.1.6 Sources Resulting from Design Basis Accidents

Radiation sources and shielding requirements resulting from design basis accidents are presented and evaluated in Section 18.1.20 Plant Shielding (II.B.2).

12.2.1.7 Site Boundary N-16 Shine Dose The N-16 present in the reactor steam in the primary steam lines, turbines, and moisture separator can contribute to the site boundary dose as a result of high energy gamma emission.

The turbine shielding was designed to minimize shine dose. The N-16 shine dose rate at the site boundary was calculated based on the final turbine shielding design. The turbine operating floor component N-16 inventories are listed in Table 12.2-11.

12.2.1.8 Stored Radioactivity

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-8 Normally the only sources of activity not stored inside the plant structures are the refueling water storage tank (RWST) and the condensate storage tank (CST). A berm is provided around these tanks to retain leakage and provide for Zone 1 (<0.5 mrem/hr) access. Provisions have been made to recycle the water from both the condensate and refueling water storage tanks to the condensate demineralizer.

Under normal conditions the condensate storage tank contains concentrations of radionuclides that yield a surface exposure rate of less than 1. 5 mr/hr. The condensate storage tank isotopic inventory is shown in Table 12.2-29.

The refueling water storage tank is also expected to have a maximum contact exposure rate of less than1.5 mr/hr.

All spent fuel is stored in the spent fuel pool until it is placed in the spent fuel shipping cask for offsite transport or transferred to the Independent Spent Fuel Storage Installation (ISFSI) described in Section 11.7. Storage space is provided in the radwaste enclosure for interim storage of packaged solid waste materials. Radioactive wastes stored inside the plant structures are shielded in place such that there is normally Zone I access outside the structure.

A separate Steam Dryer Storage Facility (SDSF) is provided within the plant protected area for the storage shielding and radioactive decay of replaced Reactor steam dryers. The steam dryers are cut in half and packaged into steel containers for storage in the SDSF. They are not considered as radioactive waste but are treated as irradiated plant equipment. The SDSF is a

reinforced concrete vault with removable roof slab access only, meeting 10CFR20 dose limits. Low Level Radwaste is stored in the Low Level Radwaste Holding Facility (LLRWHF) and is described in Section 11.6.

12.2.1.9 Special Sources Special materials used in the radiochemistry laboratory and sealed sources used for calibration purposes are of the low activity level and are handled in accordance with station health physics procedures. Unsealed sources and radiochemistry samples are handled in hoods that exhaust to the ventilation system.

The radiation source for the Transverse Incore Probe System (TIP) is provided in Tables 12.2-41, 12.2-42, 12.2-43, and 12.2-44. The radiation source is based upon location within the core and residence time. As indicated in the tables, the TIP system consists of three components for shielding calculations, the fissionable material, non-fissionable material, and the cable. Sources are provided for each component as a function of irradiation and decay times.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-9 The reactor startup source is shipped to the site in a special cask designed for shielding. The source is transferred under water while in the cask and loaded into beryllium containers. This is then loaded into the reactor while remaining under water. The source is stored in the Spent Fuel Pool. Thus, no unique shielding requirements after reactor operation are required.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES

12.2.2.1 Sources of Airborne Radioactivity

The sources of airborne radioactivity are found in the various confined areas of the plant facility and are primarily from the process leakage of the systems carrying radioactive gases, steam, and liquids. Depending on the type of the system and its physical condition, such as system pressures and temperatures, the leakage will be as a gas, steam, liquid, or a mixture of these.

12.2.2.2 Production of Airborne Materials

Radioactive materials become airborne through a number of mechanisms. The primary production mechanisms are spraying, splashing, flashing, evaporation, and diffusion.

12.2.2.3 Locations of Sources of Airborne Radioactivity

The primary sources of airborne radioactivity are found in the reactor, turbine, and radwaste buildings. Within these structures, the radioactivity may be released in equipment cubicles, system compartments, valve and piping galleries, sampling stations, radwaste handling areas, cleaning and decontamination areas and repair shops.

12.2.2.4 Control of Airborne Radioactivity Ventilation is an effective means of controlling airborne radioactive materials. Ventilation flow paths are designed such that air from low potential airborne areas flows toward the higher potential airborne areas. This flow pattern will ensure that activity released in the above mentioned source locations, which usually have low personnel access requirements, will have little chance to escape to areas with a high personnel occupancy such as corridors, working aisles and operating floors.

12.2.2.5 Methodology for Estimating the Concentration of Airborne Radioactive Material Within the Plant

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-10 In order to estimate the airborne radioactive material concentrations at locations within the plant, the following methodology was used:

(1) Estimate the total airborne releases (in Curies per year) for each of the buildings of the plant; (2) Estimate a distribution for these releases among the various equipment areas of each building based on operating data and engineering judgment; (3) Determine the annual exhaust flow from each equipment area, (4) Calculate the resultant airborne radionuclide concentration (µCi/cc) in each equipment area based on the release distribution (Ci/yr) and exhaust flow rate (cc/yr).

The following subsections discuss each step in the above procedure in more detail.

12.2.2.6 Estimation of Total Airborne Releases Within the Plant

The estimated quantities of untreated airborne radioactive material produced in the buildings of the plant are given in Table 12.2-30. These releases were based upon NUREG-0016 Revision 1, "Calculation of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors." The quantities in Table 12.2-30 were generated as follows:

- All turbine building releases were originally reduced by a factor of five to take credit for the leakage collection system installed for valves in lines 2 1/2" and larger.

Valves in the turbine building were originally provided with valve stem packing leakoff connections. Research and testing has shown that improved packing provides an effective seal to prevent leakage into the Turbine Building. As a result, these leakoff connections are in the process of being removed and packing configurations changed, as appropriate, to conform with the new requirements. As part of this effort, leakoff isolation valves and piping will be removed (or abandoned in place) and the leakoff collection header piping will be removed or abandoned in place.

(NOTE: Releases assigned to the turbine building are assumed to include any control structure

releases.)

- The reactor building releases were taken to be the sum of releases listed in NUREG-0016 Revision 1 for the auxiliary building and containment building.

SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-11 - The radwaste building releases are "per reactor" and consequently were doubled for Susquehanna SES.

12.2.2.7 Distribution of Airborne Releases Within the Plant

The approach taken to determine the anticipated distribution of gaseous effluents assumed that all untreated airborne radioactive material originates only within the equipment areas of the plant. It was further assumed that a major percentage of the release is generated within a few specific areas of each building. These ar eas are identified as per NUREG-0016 Revision 1, Section 2.2.4, "Gaseous Releases From Building Ventilation Systems", as follows:

For the Reactor Building, 90% of the normal power operation releases are due to the reactor water cleanup (RWCU) pumps and filter/demineralizers (F/Ds), and the emergency core cooling systems (ECCS) and 10% of the releases are due to the fuel handling and drywell areas.

During shutdown, 10% of the releases are from the RWCU and ECCS Systems and 90% of the releases are from the fuel handling and drywell areas.

For the Turbine Building, 85% of the normal power operation releases are from the main condenser area and 15% of the releases are from miscellaneous areas such as the steam jet air ejector (SJAE) room, the turbine operating floor, the feedwater pump room, and the mechanical vacuum pump room. During shutdown, 50% of the releases are from the main condenser area and 50% of the releases are from miscellaneous areas.

For the Radwaste Building, during normal power operation and shutdown, 10% of the releases are from the solid waste handling areas and 90% of the releases are from the liquid waste handling areas.

Releases were assumed to be generated continuously throughout the year.

The selection and relative contributions of the major areas was determined from NUREG-0016 Revision 1, based on EPRI NP-495 and NRC studies. These studies provide data on the

important sources of Iodine-131 at operating BWR's and determined the relative release rate from each source. The relative release rates for all airborne radionuclides were then assumed to be directly proportional to the Iodine-131 release rates.

Table 12.2-31 lists the major airborne contributors in each building and the percentage of the total building release assigned to each. Tables 12.2-32 through 12.2-34 provide the specific equipment areas of the plant associated with the major contributors and the applicable exhaust SSES-FSAR Text Rev. 59 FSAR Rev. 65 12.2-12 air flow rates. Note that only those equipment areas which have a significant potential for airborne radioactive material releases were included in the "other equipment areas" category.

12.2.2.8 Estimated Airborne Radioactive Material Concentrations Within the Plant

The airborne radionuclide concentrations for each equipment area were calculated using the following methodology. For a specific area, the appropriate building release (Table 12.2-30) was multiplied by the applicable release percentage for the area (Table 12.2-31) and divided by the area annual exhaust flow (Table 12.2-32, 12.2-33, or 12.2-34). The resultant concentrations are presented in Tables 12.2-35 through 12.2-37 and are compared to 10CFR20 Appendix B

requirements.

12.2.2.9 Changes to Source Data Since PSAR Source data has been updated to reflect changes in contained sources due to the implementation of power uprate, 24-month fuel cycle, design and operational changes, hydrogen water chemistry, condensate filtration, and NUREG-0016, Revision 1.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-9 THIS TABLE HAS BEEN INTENTIONALLY LEFT BLANK

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-11 STEAM N-16 SHIELDING SOURCE TERMS FOR TURBINE AND REACTOR BUILDING EQUIPMENT (1)(2) Equipment Decay Time (2) (Seconds)

N-16 (1) µCi/gm Steam Density gm/cc Reactor Pressure Vessel Nozzle 0 250 3.51E-02 (3) 24" Main Steam Line 0 250 3.51E-02 Main Steam Stop Valves 1.87 227.3 3.51E-02 High Pressure Turbine Inlet 2.30 218.0 3.51E-02 High Pressure Turbine Outlet 2.41 215.6 8.67E-03 Moisture Separator Inlet 2.41 215.6 8.67E-03 Moisture Separator Outlet 3.11 201.4 7.22E-03 Cross Around Piping 3.11 201.4 7.22E-03 Combined Intermediate Valve 3.11 201.4 7.22E-03 Low Pressure Turbine Inlet 3.11 201.4 7.22E-03 Low Pressure Turbine Outlet 3.36 196.6 9.28E-05 Steam to Reactor Feedwater Turbine 4" Turbine 10" 0.37 5.24 250 163.7 3.51E-02 7.30E-03 Reactor Feedwater Turbine Exhaust 5.24 163.7 1.07E-04 Feedwater Heaters

E-103 E-104 E-105 3.04 3.15 2.42 202.9 200.6 215.4 2.75E-03 4.71E-03 1.11E-02 Steam Seal Evaporator (Inlet Steam) 5.51 159.5 2.58E-03 RCIC Turbine (Inlet Steam) 1.0 247.2 3.51E-02 HPCI Turbine (Inlet Steam) 1.0 247.2 3.51E-02 Notes: 1. Values presented are based on plant operation with hydrogen water chemistry.

2. Decay time is based on a conservative estimate of the transit time at a Main Steam flow rate of 1.65E+07 lbs/hr.

3. 3.51E-02 = 3.51 X 10

-2

SSES-FSAR Table 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-15 CONDENSATE SHIELDING SOURCE TERMS(1)(2)(4) Isotope µCi/cc Isotope µCi/cc Isotope µCi/cc Br-83 4.75E-04(3) Ba-139 1.72E-04 Kr-83m 3.74E-06 Br-84 1.14E-03 Ba-140 9.34E-06 La-144 1.44E-06 Br-85 7.35E-04 Ba-141 3.32E-04 Cs-141 2.72E-04 I-131 3.38E-04 Ba-142 2.14E-04 Cs-140 2.17E-03 I-132 3.81E-03 Np-289 2.40E-04 Cs-139 8.01E-04 I-133 2.34E-03 Na-24 2.00E-06 Rb-90m 1.76E-04 I-134 9.34E-03 Cr-51 5.00E-07 Rb-88 3.30E-05 I-135 3.63E-03 Mn-56 4.96E-05 Sr-94 9.91E-06 Sr-89 3.10E-06 Co-58 5.00E-06 Y-94 1.31E-06 Sr-90 2.30E-07 Co-60 5.00E-07 Sr-95 1.05E-06 Sr-91 8.10E-05 Ni-65 2.97E-07 Y-95 1.20E-06 Sr-92 1.39E-04 W-187 3.00E-06 Sr-93 1.23E-04 Mo-99 2.20E-05 La-142 3.41E-06 Y-93 2.88E-07 Tc-99m 2.79E-04 La-141 1.77E-06 Y-93m 4.80E-05 Tc-101 1.27E-04 Y-92 9.04E-07 Rb-91 2.17E-03 Te-132 4.90E-05 Y-91m 1.20E-06 Rb-90 1.74E-03 Cs-134 1.60E-07 Xe-135 5.60E-06 Rb-89 2.51E-04 Cs-136 1.10E-07 Xe-135m 3.41E-05 N-13 6.09E-03 Cs-137 2.40E-07 Xe-133 3.20E-07 N-16 4.29E-04 Cs-138 2.68E-04 Kr-85m 1.38E-06 Notes:

(1) Basis for condensate isotopic sources:

a. Daughter products formed from deca y of noble gases in transit through the turbine and main condenser.

b. Coolant and noncoolant particulate activation and fission product carryover fraction of 1X10

-3 per unit weight into steam. c. Reaction water fission product halogen carryover fraction of 8X10

-2 per unit weight into steam.

(2) Isotopes with activity concentrations less than 10

-7 µCi/cc are not shown.

(3) 4.75E-04 = 4.75 X 10

-4 (4) Values shown are conservatively based on 2 minutes holdup. Minimum retention time in the Hotwell is estimated at 3 minutes.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-17 OFFGAS LINE FROM SJAE TO RECOMBINER SHIELDING SOURCE TERMS (1)(3) Isotope Curies Isotope Curies Isotope Curies Xe-133 3.49E-03(2) Ba-139 1.14E-02 Sr-91 1.00E-02 Xe-135m 1.13E-02 Cs-138 3.91E-03 Y-91 1.00E-02 Xe-135 9.10E-03 Cs-137 3.90E-03 Y-91m 5.93E-03 Xe-137 6.26E-02 Kr-83m 1.44E-03 Rb-90 1.12E-02 Xe-138 3.78E-02 Kr-85m 2.58E-03 Sr-90 7.01E-03 Xe-139 1.10E-01 Kr-87 8.33E-03 Y-90 7.01E-03 Xe-140 1.00E-01 Kr-88 8.34E-03 Rb-89 5.62E-03 Xe-141 1.94E-02 Kr-89 5.41E-02 Sr-89 5.60E-03 Xe-142 4.82E-03 Kr-90 1.07E-01 N-13 2.11E-02 Cs-142 1.00E-03 Kr-91 9.70E-02 N-16 7.79E+01 Cs-141 2.23E-03 Kr-92 4.79E-02 N-17 2.85E-02 Ba-141 2.02E-03 Kr-93 9.39E-03 O-19 5.27E-01 La-141 2.01E-03 Kr-94 1.39E-03 F-18 3.03E-03 Ce-141 2.01E-03 Rb-93 1.41E-03 Cs-140 1.08E-02 Rb-92 7.30E-03 Ba-140 1.04E-02 Sr-92 4.98E-03 La-140 1.04E-02 Y-92 4.98E-03 Cs-139 1.14E-02 Rb-91 1.04E-02

Notes:

(1) Bases: a. 100,000

µCi/sec noble gases (after 30 min. decay). b. Decay time from Reactor Nozzl e to Preheater Inlet, 5.23 seconds. c. Main Steam flow 1.65E+07 lbs/ hr. d. All particulates generated plateout on pipe surfaces. e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation. (2) 3.49E-03 = 3.49 X 10

-3 (3) Isotopes with activity less than 10

-3 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-18 OFF-GAS RECOMBINER SH IELDING SOURCE TERMS(1)(2) Isotope Curies Isotope Curies Isotope Curies Xe-133 3.44E-03(3) Cs-139 2.30E-01 Rb-92 8.28E-02 Xe-135m 1.12E-02 Ba-139 2.30E-01 Sr-92 8.59E-02 Xe-135 8.97E-03 Cs-138 7.98E-02 Y-92 8.59E-02 Xe-137 6.15E-02 Cs-137 7.96E-02 Rb-91 1.94E-01 Xe-138 3.72E-02 Kr-83m 1.42E-03 Sr-91 1.95E-01 Xe-139 1.07E-01 Kr-85m 2.54E-03 Y-91 1.95E-01 Xe-140 9.48E-02 Kr-87 8.21E-03 Y-91m 1.15E-01 Xe-141 1.32E-02 Kr-88 8.22E-03 Rb-90 2.22E-01 Xe-142 3.14E-03 Kr-89 5.32E-02 Sr-90 1.41E-01 Cs-142 6.85E-03 Kr-90 1.03E-01 Y-90 1.41E-01 Ba-142 7.46E-03 Kr-91 8.93E-02 Rb-89 1.14E-01 La-142 7.46E-03 Kr-92 3.81E-02 Sr-89 1.14E-01 Cs-141 3.07E-02 Kr-93 6.74E-03 Rb-88 1.76E-02 Ba-141 3. 09E-02 Rb-94 1.97E-03 N-13 2.08E-02 La-141 3.09E-02 Sr-94 2.11E-03 N-16 7.03E+01 Ce-141 3.09E-02 Y-94 2.11E-03 N-17 2.39E-02 Ba-140 2.05E-01 Rb-93 1.51E-02 O-19 5.07E-01 Cs-140 2.05E-01 Sr-93 1.56E-02 F-18 2.99E-03 La-140 2.05E-01 Y-93 1.56E-02 NOTES: (1) Bases a) 100,000

µCi/sec noble gases (after 30 minute decay). b) Decay time from Reactor Nozzle to entrance of recombiner, 6.22 seconds. c) Particulate daughter products accumu late for 40 years in Recombiner. d) Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell. e) Main steam flow 1.65E+07 lbs/hr. f) Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation. (2) Isotopes with activity less then 10

-3 Ci are not shown. (3) 3.44E-03 = 3.44 X 10

-3 SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-19 OFF-GAS LINE FROM RECOMBINER TO CHARCOAL SYSTEM SHIELDING SOURCE TERMS(1),(2) Isotope Curies Isotope Curies Isotope Curies Xe-133m 1.32E-02(3) Cs-138 4.48E-01 Y-91 4.45E-02 Xe-133 3.79E-01 Cs-137 3.24E-01 Y-91m 2.65E-02Xe-135m 1.16E+00 Kr-83m 1.55E-01 Rb-90 7.20E-01Xe-135 9.90E-01 Kr-85m 2.79E-01 Sr-90 1.90E-01Xe-137 5.34E+00 Kr-87 8.95E-01 Y-90 1.90E-01Xe-138 3.88E+00 Kr-88 9.01E-01 Rb-89 5.75E-01Xe-139 3.96E+00 Kr-89 4.39E+00 Sr-89 4.42E-01Xe-140 1.07E+00 Kr-90 2.9 E+00 Rb-88 1.15E-01Cs-140 4.27E-01 Kr-91 4.31E-01 N-13 2.08E+00Ba-140 1.09E-01 Kr-92 1.20E-02 N-16 1.80E+02La-140 1.09E-01 Rb-92 1.07E-02 N-17 1.76E-02 Cs-139 5.93E-01 Rb-91 1.67E-01 F-18 3.27E-01 Ba-139 4.02E-01 Sr-91 4.46E-02 NOTES:

(1) Bases a) 100,000

µCi/sec noble gases (after 30 minute decay). b) Main steam flow 1.65E+07 lbs/hr. c) Decay time from Reactor Nozzle to offgas line entrance, 45.5 seconds. d) Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation. e) Nitrogen partition factors: 80 percent for non-condensible gases; 20 percent to main condenser hotwell. (2) Isotopes with activity less then 10

-2 Ci are not shown. (3) 1.32E-02 = 1.32 X 10

-2

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-21 OFFGAS AMBIENT CHARCOAL FREON CHILLER SHIELDING SOURCE TERMS (1) (3)

Isotope Curies Isotope Curies Kr-87 1.62E-01(2) Rb-89 3.81E-01 Kr-88 1.66E-01 Sr-89 3.81E-01 Kr-89 4.07E-01 Rb-88 1.54E-01 Xe-135m 1.87E-01 N-13 3.11E-01 Xe-135 1.84E-01 Xe-137 5.64E-01 Xe-138 6.37E-01 Cs-138 5.94E-01 Cs-137 3.18E-01

NOTES:

(1) Bases: a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to inlet of Freon Chiller, 472 seconds. c. Particulate daughter products accumulate for 40 years in equipment. d. Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell. e. Main Steam flow 1.65E+07 lbs /hr. f. Normal water chemistry source terms for N-13, N-16, and N-17 are increased by a factor of five due to hydrogen water chemistry operation.

(2) 1.62E-01 = 1.62 X 10

-1 (3) Isotopes with activities less than10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-22 OFFGAS CHARCOAL GUARD BED SHIELDING SOURCE TERMS (1) (3) Isotope Curies Isotope Curies Kr-87 1.13E-01(2) Rb-89 3.64E-01 Kr-88 1.15E-01 Sr-89 3.64E-01 Kr-89 2.77E-01 Rb-88 1.51E-01 Xe-135m 1.30E-01 N-13 2.15E-01 Xe-135 1.28E-01 Xe-137 3.85E-01 Xe-138 4.42E-01 Cs-138 5.78E-01 Cs-137 3.05E-01

NOTES:

(1) Bases: a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to Charcoal Guard Bed, 486 seconds. c. Particulate daughter products accumulate for 40 years in equipment. d. Nitrogen partition factors: 80 percent noncondensible gases, 20 percent to main condenser hotwell.

e. Main Steam flow 1.65E+07 lbs/hr. f. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 1.13E-01 = 1.13 X 10

-1.

(3) Isotopes with activity less than 10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-23 OFFGAS CHARCOAL BEDS SHIELDING SOURCE TERMS (1)(3)(4)(5)

Isotope Curies Isotope Curies F-18dd 3.87E+00 Kr-83m 3.10E+01 Xe-131m 4.79E+00 Kr-85m 1.29E+02 Xe-133m 4.75E+01(2) Kr-85 8.05E+00 Xe-133 2.08E+03 Kr-87 1.21E+02 Xe-135m 2.89E+01 Kr-88 2.70E+02 Xe-135 9.36E+02 Kr-89 1.24E+01 Xe-137 2.13E+01 Rb-89 1.26E+01 Xe-138 1.11E+02 Sr-89 1.25E+01 Xe-139 1.82E-01 Rb-88 2.70E+02 Cs-139 1.97E-01 N-13 1.25E+01 Ba-139 1.97E-01 Cs-137 1.29E+01 Cs-138 1.11E+02 NOTES:

(1) Bases:

a. 100,000

µCi/sec noble gas release rate after 30 min. decay.

b. Decay time from Reactor Nozzle to inlet of Main Charcoal Beds (1 st bed), 503 seconds.

c. Particulate daughter products accumulate for 40 years in equipment.

d. Main Steam flow 1.65E+07 lbs/hr.

e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 4.75E+01 = 4.75 X 10

+1.

(3) Activity per charcoal bed (1 only).

(4) Representative of each of the five charcoal beds.

(5) Isotopes with activity less than 10

-1 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-24 OFFGAS POST HEPA FILTER SHIELDING SOURCE TERMS (1)(3) Isotope Curies Xe-131m 1.25E-05(2) Xe-133 2.12E-03 Kr-83m 2.08E-05 Kr-85m 5.21E-04 Kr-85 5.62E-05 Kr-87 2.82E-05 Kr-88 4.72E-04 Rb-88 3.06E-03 N-13 3.01E-03 F-18 1.29E-02

NOTES:

(1) Bases:

a. 100,000

µCi/sec noble gas release rate after 30 min. decay. b. Decay time from Reactor Nozzle to outlet of ambient Charcoal Absorbers: Kryptons - 1.16E+05 seconds Xenons - 2.05E+06 seconds

c. Particulate daughter products accumulate for 40 years in equipment.

d. Main Steam flow 1.65E+07 lbs /hr. e. Normal water chemistry source terms fo r N-13, N-16, and N-17 are increased by a factor of five due to hydr ogen water chemistry operation.

(2) 1.25E-05 = 1.25 X 10

-5.

(3) Isotopes with activity less than 10

-5 Ci are not shown.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 2 of 2 TABLE 12.2-30 ESTIMATED AIRBORNE RADIOACTIVE RELEASES PRIOR TO TREATMENT (CURIES /YEAR)

(1) Notes:

1. Based on NUREG-0016, Revision 1, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water

Reactors", (BWR - GALE Code), Section 2.2.4.

2. Turbine Building releases reduced by a factor of 5 due to installation of the Process Valve Stem Leakoff Collection System.

3. Reactor Building is the sum of th e Containment & Auxiliary Building.

4. 4.00E+00 = 4.00 X 10

0.

SSES-FSAR Table Rev. 55 FSAR Rev. 64 Page 1 of 1 TABLE 12.2-32 TURBINE BUILDING AIRBORNE SOURCE DESCRIPTIONS AREA DESIGNATION ROOM NUMBERS AND AREAS INCLUDED TOTAL EXHAUST FLOW (CFM) TOTAL ANNUAL (1) EXHAUST (cc/yr) Condenser 30, 31, 32, 34, 37, 38, 39, 113, 211 23,100 3.44E+14(2) Miscellaneous Areas C-10, C-12, C-130, C212 C220, C301, C-400, C-900, C-912, 33, 35, 42, 43, 45, 53, 59, 111, 112, 114,115, 116, 117, 118, 119, 120, 121, 122, 123, 124, 125, 126, 128, 130, 210, 212, 214, 215, 216, 217, 300, 415, 416, 417, 419, 421 183,900 2.74E+15 Notes:

1. Based on continuous release of 365 days per year.

2. 3.44E+14 = 3.44 x 10 14

SSES-FSARNIMSRev.60FSARRev.65Page1of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable1646'Offgaslinesfromrecombinertoradwaste8"GBC-1068"diam.cyl.1.17x10

-30.32212.2-19 30 31 656'656'Condensatepumpsandlines1P-102A,B,C,D48",30"and18"diam.cyl.1.00.37512.2-15 32 34 656'656'RecombinerRecombiner1S-1255'diam.x6'-2"ht.cyl.1.21x10-30.7512.2-1833656'SteamPackingexhauster1E-11036"diam.cyl.30"diam.cyl.1.00.37511.1-1,-2,-3,-455656'Vacuumpump1P-10512"diam.cyl.1.4x10

-30.40611.1-1,-2,-3,-456656'RecombinerpipevalvemanifoldfromSJAE8"diam.and10"diam.cyl.1.21x10-30.32212.2-17C-11656'ChemdraintankandpumpsOT-114OP-132A,B3'-6"diam.6'-4"ht.cyl.1.00.18711.2-642656'CondensatedemineralizerpipewayVariousresinlines30",24",14",3"diam.cyl.1.00.21612.2-1612.2-15152656'Regenerationwastesurgetanks1T-106A,B7'diam.13'highcyl.1.00.512.2-16151656'Neutralizingpumps1P130A,B2"diam.cyl.1.00.15412.2-1643656'Chemicalwaste1T130A,B12'diam.1.00.19712.2-16Neutralizingtanks20'length37656'RFPturbine27"x75"RFT3'diam.cyl.2.04x10

-40.2512.2-1138656'exhaustductduct11.1-1,-2,-3,-439656' SSES-FSARNIMSRev.60FSARRev.65Page2of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable36656'MaincondenserandsteampipingIE-108A,B,C,RFWductcondensatelines3'diam.36"diam.cyl.2.04x10-41.00.2511.1-1,-2,-3,-412.2-11 12.2-1559656'CondensateFiltrationBackwashReceivingTank1T-1879'diam.x24'1cyl1.00.2512.2-48117676'Deepbed1F-106G11'diam.cyl.1.01.512.2-6118676'Condensatedemin.1F-106F3'deep1.01.512.2-16119676'1F-106E1.01.512.2-16120676'1F-106D1.01.512.2-16121676'1F-106C1.01.512.2-16123676'1F-106B1.01.512.2-16124676'1F-106A1.01.512.2-16125676'1F-106H1.01.512.2-16114676'RFPturbine105C4"diam.cyl.0.0330.33712.2-11115676'RFPturbine105B10"diam.cyl.0.00610.36511.1-1,-2,-3,-4116676'RFPturbine105A111676'SJAE1E-1098"diam.cyl.1.21x10

-30.32212.2-17112676'FutureSJAE3'-6"diam.x12'hcyl.1.21x10-31.08113676'Maincondensersteampiping1E-108A,B,C42",28",26",16"diam.cycl..00623,.036.00216,0.6250.37511.1-1,-2,-3,-4,12.2-11.00786125676'Anionandcation1T-1574'6"øx14'hcyl.1.00.2512.2-16regeneration1T-158,1596'6"x14'hcyl.

SSES-FSARNIMSRev.60FSARRev.65Page3of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable110676'Turbinebldg.1C-132A,B3/8"SStubingVaries-11.1-1,-2,-3,-4samplestation130676'CondensateFilterBundles1F135A,B,C,D,E,F6'diam.x4'-2"hcyl.0.9751.58412.2-46133(1)676'CondensateFilterBundles2F125G6'diam.x4'-2"hcyl.0.9751.58412.2-46 130,133(1),C-100676'BackwashPiping12"HCD-107(130)12"HCD-207(133,C-100)12'diam.cyl.1.00.2512.2-47211699'Maincondenserandsteampiping1E108A,B,C,24"DBB-101102,203,10424"diam.cyl.28"diam.cyl.

42"diam.cyl..036.0328 0.006740.940.375 0.62511.1-1,-2,-3,-412.2-11212699'Feedwaterheaters1E1036'øx45'cyl.0.7149/1611.1-1,-2,-3,-4214699'Feedwaterheaters1E1046'øx40'cyl.0.8159/1612.2-11215699'Feedwaterheaters1E1056'øx36'cyl.0.8539/16300714'Mainsteamtunnel24"-DBB-101,24"diam.cyl..0360.9711.1-1,-2,-3,-4102,103,10412.2-11411729'Mainsteamtunnel24"-DBB-101,24"diam.cyl..0360.9711.1-1,-2,-3,-4102,103,10412.2-11416729'Steamseal1E-12851/2'øx34'cyl.0.4475/811.1-1,-2,-3,-4evaporator12-EBB-11312"diam.cyl.2.25-30.68712.2-11419729'Moistureseparator1T-104A,B10'-8"x67.3'cyl.0.7680.7511.1-1,-2,-3,-4421729'and42"cross-oversteamlines42"-GESupplied42"diam.cyl.0.00630.62512.2-11 SSES-FSARNIMSRev.60FSARRev.65Page4of4TABLE12.2-38TURBINEBUILDINGSHIELDINGDESIGNRADIATIONSOURCEDESCRIPTION RoomNo.E1DominantRadiationSourceSourceLocationorIdentificationGeometryEffectiveSourceDensitygm/ccEquipmentSelfShielding(Steel)Thicknessin.)FSARSourceTable420729'HP&LPturbines1G-10128"diam.cyl..03281.411.1-1,-2,-3,-428",42"GESupplied42"diam.cyl..006230.62512.2-11531762'H&VSystem1F-156,1572'x2'x1'(63units)0.490.25N/A532762'(HEPA&charcoal)1F-158315ft 3(carbon)Note:1.Room133onlyappliestoUnit2.

SSES-FSAR Table Rev. 54 FSAR Rev. 65 Page 1 of 1 Table 12.2-45 This Table Has Been Deleted

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 1 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr H 3 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.57E+04 7.55E+04 7.47E+04 7.37E+04 7.17E+04 6.40E+04 Na 24 2.73E+02 2.73E+02 2.66E+02 2.60E+02 1.86E+02 8.79E+01 2.91E+00 4.46E-13 0.00E+00 Cr 51 5.65E+06 5.65E+06 5.65E+06 5.65E+06 5.61E+06 5.52E+06 5.11E+06 2.67E+06 5.95E+05 6.26E+04 6.08E+02 7.04E-06 Mn 54 4.27E+05 4.27E+05 4.27E+05 4.27E+05 4.27E+05 4.26E+05 4.23E+05 4.00E+05 3.50E+05 2.87E+05 1.90E+05 3.75E+04 Fe 55 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.86E+06 1.79E+06 1.68E+06 1.48E+06 8.86E+05 Mn 56 1.21E+07 1.21E+07 1.05E+07 9.17E+06 1.40E+06 1.90E+04 7.47E-05 0.00E+00 Co 58 5.91E+05 5.91E+05 5.91E+05 5.91E+05 5.90E+05 5.86E+05 5.68E+05 4.41E+05 2.45E+05 1.02E+05 1.67E+04 1.32E+01 Fe 59 1.28E+05 1.28E+05 1.28E+05 1.28E+05 1.28E+05 1.27E+05 1.21E+05 3.02E+04 3.16E+04 7.79E+03 4.35E+02 4.97E+03 Ni 59 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 Co 60 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.16E+05 3.09E+05 2.99E+05 2.80E+05 2.35E+05 Ni 63 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.54E+04 3.54E+04 3.53E+04 3.48E+04 Cu 64 2.65E+02 2.65E+02 2.58E+02 2.51E+02 1.72E+02 7.17E+01 1.41E+00 2.28E-15 0.00E+00 Ni 65 5.57E+04 5.57E+04 4.86E+04 4.23E+04 6.17E+03 7.57E+01 1.89E-07 0.00E+00 Ga 73 4.97E+03 4.97E+03 4.64E+03 4.32E+03 1.59E+03 1.63E+02 5.65E-03 0.00E+00 Ge 73m 4.91E+03 4.91E+03 4.58E+03 4.26E+03 1.57E+03 1.60E+02 5.58E-03 0.00E+00 As 76 2.56E+03 2.56E+03 2.53E+03 2.50E+03 2.08E+03 1.36E+03 2.05E+02 1.49E-05 5.05E-22 0.00E+00 Ge 77 7.31E+04 7.31E+04 7.10E+04 6.88E+04 4.48E+04 1.68E+04 2.02E+02 4.83E-15 0.00E+00 As 77 2.26E+05 2.26E+05 2.25E+05 2.23E+05 2.04E+05 1.60E+05 4.61E+04 6.72E-01 4.61E-12 8.71E-29 0.00E+00 Ge 78 7.22E+05 7.22E+05 5.70E+05 4.50E+05 1.65E+04 8.56E+00 1.43E-14 0.00E+00 As 78 7.32E+05 7.32E+05 7.13E+05 6.70E+05 8.25E+04 1.25E+02 1.39E-12 0.00E+00 Br 82 3.81E+05 3.81E+05 3.78E+05 3.75E+05 3.27E+05 2.38E+05 5.81E+04 2.77E-01 1.46E-13 0.00E+00 Br 83 1.29E+07 1.29E+07 1.18E+07 1.04E+07 1.40E+06 1.38E+04 1.28E-05 0.00E+00 Kr 83m 1.31E+07 1.31E+07 1.29E+07 1.26E+07 3.64E+06 5.27E+04 5.39E-05 0.00E+00 Kr 85 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.48E+06 1.47E+06 1.46E+06 1.44E+06 1.39E+06 1.22E+06 Kr 85m 2.68E+07 2.68E+07 2.51E+07 2.32E+07 7.87E+06 6.62E+05 9.63E+00 0.00E+00 Rb 86 2.17E+05 2.17E+05 2.16E+05 2.16E+05 2.14E+05 2.09E+05 1.86E+05 7.10E+04 7.62E+03 2.68E+02 2.73E-01 4.31E-13 Kr 87 5.37E+07 5.37E+07 4.13E+07 3.15E+07 6.94E+05 1.13E+02 1.03E-15 0.00E+00 Kr 88 7.45E+07 7.45E+07 6.60E+07 5.84E+07 1.05E+07 2.12E+05 4.93E-03 0.00E+00 Rb 88 7.64E+07 7.64E+07 7.16E+07 6.46E+07 1.18E+07 2.38E+05 5.51E-03 0.00E+00 Zr 89 5.52E+04 5.52E+04 5.50E+04 5.48E+04 5.15E+04 4.47E+04 2.37E+04 9.55E+01 2.84E-04 1.46E-12 1.74E-29 0.00E+00 Sr 89 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.02E+08 9.78E+07 6.85E+07 3.01E+07 8.79E+06 6.91E+05 3.09E+01

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 2 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Y 89m 1.68E+05 1.63E+05 6.43E+04 6.40E+04 6.07E+04 5.39E+04 3.25E+04 6.46E+03 2.80E+03 8.18E+02 6.43E+01 2.87E-03 Sr 90 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Y 90 1.36E+07 1.36E+07 1.36E+07 1.36E+07 1.35E+07 1.34E+07 1.32E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Sr 91 1.31E+08 1.31E+08 1.26E+08 1.21E+08 7.30E+07 2.28E+07 1.21E+05 2.23E-15 0.00E+00 Y 91 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.29E+08 9.48E+07 4.65E+07 1.60E+07 1.78E+06 3.11E+02 Y 91m 7.57E+07 7.57E+07 7.53E+07 7.40E+07 4.64E+07 1.44E+07 7.64E+04 1.41E-15 0.00E+00 Sr 92 1.39E+08 1.39E+08 1.22E+08 1.08E+08 1.80E+07 2.99E+05 3.01E-03 0.00E+00 Y 92 1.40E+08 1.40E+08 1.39E+08 1.37E+08 6.53E+07 4.42E+06 4.04E+00 0.00E+00 Zr 93 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 Y 93 1.07E+08 1.07E+08 1.04E+08 1.01E+08 6.23E+07 2.08E+07 1.48E+05 3.74E-14 0.00E+00 Zr 95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.91E+08 1.89E+08 1.84E+08 1.39E+08 7.23E+07 2.73E+07 3.67E+06 1.35E+03 Nb 95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.79E+08 1.21E+08 5.36E+07 7.92E+06 2.98E+03 Nb 95m 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.12E+06 2.10E+06 1.63E+06 8.50E+05 3.21E+05 4.32E+04 1.59E+01 Nb 96 3.12E+05 3.12E+05 3.07E+05 3.02E+05 2.45E+05 1.53E+05 1.80E+04 1.63E-04 4.44E-23 0.00E+00 Zr 97 1.90E+08 1.90E+08 1.86E+08 1.82E+08 1.36E+08 7.09E+07 3.70E+06 2.84E-05 0.00E+00 Nb 97 1.91E+08 1.91E+08 1.90E+08 1.88E+08 1.46E+08 7.13E+07 3.72E+06 3.06E-05 0.00E+00 Nb 97m 1.80E+08 1.80E+08 1.76E+08 1.73E+08 1.30E+08 6.73E+07 3.51E+06 2.70E-05 0.00E+00 Mo 99 2.02E+08 2.02E+08 2.01E+08 2.00E+08 1.86E+08 1.57E+08 7.37E+07 1.05E+05 2.79E-02 3.84E-12 0.00E+00 Tc 99 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.21E+03 2.22E+03 2.22E+03 2.22E+03 2.22E+03 2.22E+03 Tc 99m 1.79E+08 1.79E+08 1.79E+08 1.79E+08 1.73E+08 1.51E+08 7.14E+07 1.01E+05 2.70E-02 3.72E-12 0.00E+00 Ru103 1.72E+08 1.72E+08 1.72E+08 1.71E+08 1.70E+08 1.69E+08 1.60E+08 1.01E+08 3.50E+07 7.15E+06 2.71E+05 6.80E-01 Rh103m 1.71E+08 1.71E+08 1.71E+08 1.71E+08 1.70E+08 1.68E+08 1.60E+08 1.01E+08 3.50E+07 7.14E+06 2.71E+05 6.79E-01 Ru105 1.19E+08 1.19E+08 1.13E+08 1.05E+08 3.51E+07 2.89E+06 3.78E+01 0.00E+00 Rh105 1.11E+08 1.11E+08 1.12E+08 1.12E+08 1.05E+08 8.02E+07 1.96E+07 9.55E+01 5.26E-11 1.74E-29 0.00E+00 Rh105m 3.38E+07 3.38E+07 3.22E+07 2.99E+07 1.00E+07 8.25E+05 1.08E+01 0.00E+00 Ru106 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rh106 7.38E+07 7.37E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rh106m 2.39E+06 2.39E+06 2.04E+06 1.74E+06 1.85E+05 1.11E+j03 1.10E-07 0.00E+00 Ag109m 4.14E+07 4.14E+07 4.04E+07 3.94E+07 2.77E+07 1.23E+07 3.22E+05 7.79E-02 7.12E-02 6.22E-02 4.71E-02 1.57E-02 Ag110 1.87E+07 1.82E+07 6.23E+03 6.23E+03 6.23E+03 6.22E+03 6.17E+03 5.74E+03 4.86E+03 3.78E+03 2.26E+03 2.98E+02 Ag110m 4.58E+05 4.58E+05 4.58E+05 4.58E+05 4.58E+05 4.57E+05 4.53E+05 4.22E+05 3.57E+05 2.78E+05 1.67E+05 2.19E+04

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 3 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Pd111 6.66E+06 6.66E+06 2.88E+06 1.29E+06 8.10E+04 1.08E+04 1.24E+00 0.00E+00 Pd111m 2.83E+05 2.83E+05 2.65E+05 2.49E+05 1.03E+05 1.38E+04 1.57E+00 0.00E+00 Ag111 6.72E+06 6.72E+06 6.71E+06 6.70E+06 6.52E+06 6.13E+06 4.65E+06 4.13E+05 1.56E+03 3.59E-01 1.18E-08 0.00E+00 Ag111m 6.71E+06 6.71E+06 3.04E+06 1.39E+06 1.01E+05 1.34E+04 1.54E+00 0.00E+00 Pd112 3.02E+06 3.02E+06 2.97E+06 2.92E+06 2.32E+06 1.37E+06 1.28E+05 1.51E-04 3.82E-25 0.00E+00 Ag112 3.03E+06 3.03E+06 3.03E+06 3.02E+06 2.64E+06 1.60E+06 1.51E+05 1.78E-04 4.49E-25 0.00E+00 In113m 3.47E+05 3.47E+05 3.47E+05 3.47E+05 3.46E+05 3.45E+05 3.39E+05 2.90E+05 2.02E+05 1.18E+05 3.85E+04 4.73E+02 Sn113 3.47E+05 3.47E+05 3.47E+05 3.47E+05 3.46E+05 3.45E+05 3.38E+05 2.90E+05 2.02E+05 1.18E+05 3.84E+04 4.72E+02 Ag113 1.69E+06 1.69E+06 1.60E+06 1.50E+06 6.05E+05 7.64E+04 7.05E+00 0.00E+00 Cd113m 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.32E+03 4.29E+03 4.23E+03 4.13E+03 3.74E+03 In114 5.63E+04 5.62E+04 3.54E+04 3.54E+04 3.53E+04 3.50E+04 3.35E+04 2.33E+04 1.01E+04 2.86E+03 2.13E+02 7.72E-03 In114m 3.71E+04 3.71E+04 3.71E+04 3.71E+04 3.69E+04 3.65E+04 3.51E+04 2.44E+04 1.05E+04 2.98E+03 2.23E+02 8.10E-03 Cd115 9.02E+05 9.02E+05 8.94E+05 8.94E+05 8.17E+05 6.62E+05 2.61E+05 8.02E+01 6.26E-07 4.32E-19 0.00E+00 Cd115m 4.20E+04 4.20E+04 4.19E+04 4.19E+04 4.18E+04 4.13E+04 3.95E+04 2.64E+04 1.04E+04 2.56E+03 1.44E+02 1.68E-03 In115m 9.02E+05 9.02E+05 9.02E+05 9.02E+05 8.63E+05 7.21E+05 2.84E+05 9.02E+01 1.15E+00 2.83E-01 1.59E-02 1.86E-07 Cd117 8.86E+05 8.86E+05 7.79E+05 6.76E+05 9.63E+04 1.12E+03 2.22E-06 0.00E+00 Cd117m 2.03E+05 2.03E+05 1.83E+05 1.66E+05 3.90E+04 1.44E+03 5.13E-04 0.00E+00 In117 6.62E+05 6.62E+05 6.56E+05 6.38E+05 2.02E+05 4.12E+03 6.63E-04 0.00E+00 In117m 8.10E+05 8.10E+05 8.02E+05 7.79E+05 2.34E+05 4.13E+03 2.74E-05 0.00E+00 Sn117m 2.49E+06 2.49E+06 2.49E+06 2.48E+06 2.44E+06 2.36E+06 2.03E+06 5.39E+05 2.53E+04 2.58E+02 2.04E-02 1.38E-18 Sn119m 2.46E+06 2.46E+06 2.46E+06 2.46E+06 2.46E+06 2.45E+06 2.43E+06 2.29E+06 1.99E+06 1.60E+06 1.04E+06 1.84E+05 Sn121 1.54E+06 1.54E+06 1.52E+06 1.51E+06 1.26E+06 8.35E+05 1.32E+05 4.17E+02 4.16E+02 4.16E+02 4.12E+02 4.03E+02 Sn121m 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.91E+02 1.91E+02 1.89E+02 1.85E+02 Sb122 2.13E+05 2.13E+05 2.12E+05 2.12E+05 1.96E+05 1.65E+05 7.65E+04 9.66E+01 1.98E-05 1.83E-15 0.00E+00 Te123m 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.63E+02 9.47E+02 8.17E+02 5.76E+02 3.42E+02 1.17E+02 1.70E+00 Sn123 1.36E+05 1.36E+05 1.36E+05 1.36E+05 1.36E+05 1.35E+05 1.33E+05 1.16E+05 8.40E+04 5.18E+04 1.92E+04 3.80E+02 Sn123m 9.25E+05 9.25E+05 5.53E+05 3.29E+05 2.31E+02 1.42E-05 0.00E+00 Sb124 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.64E+04 9.31E+04 6.89E+04 3.45E+04 1.22E+04 1.45E+03 3.22E-01 Sn125 5.57E+05 5.57E+05 5.56E+05 5.55E+05 5.44E+05 5.18E+05 4.18E+05 6.45E+04 8.61E+02 1.33E+00 2.19E-06 0.00E+00 Sn125m 2.07E+06 2.07E+06 2.35E+05 2.65E+04 1.39E-09 0.00E+00 0.00E+00 Sb125 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.35E+06 1.29E+06 1.21E+06 1.07E+06 6.43E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 4 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Te125m 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.04E+05 3.06E+05 3.04E+05 2.93E+05 2.61E+05 1.57E+05 Sb126 4.64E+04 4.64E+04 4.64E+04 4.64E+04 4.56E+04 4.39E+04 3.71E+04 8.68E+03 3.15E+02 1.36E+01 1.16E+01 1.16E+01 Sb126m 5.45E+04 5.45E+04 1.83E+04 6.18E+03 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 Sn127 3.78E+06 3.78E+06 3.21E+06 2.72E+06 2.70E+05 1.38E+03 6.56E-08 0.00E+00 Sb127 9.40E+06 9.40E+06 9.40E+06 9.32E+06 8.94E+06 7.95E+06 4.61E+06 4.28E+04 8.71E-01 7.95E-08 2.60E-22 0.00E+00 Te127 9.32E+06 9.32E+06 9.32E+06 9.32E+06 9.24E+06 8.63E+06 5.78E+06 1.38E+06 9.09E+05 5.13E+05 1.58E+05 1.52E+03 Te127m 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.57E+06 1.36E+06 9.32E+05 5.25E+05 1.61E+05 1.55E+03 Sb128 1.61E+06 1.61E+06 1.57E+06 1.53E+06 9.17E+05 2.67E+05 1.05E+03 1.50E-18 0.00E+00 Sb129 3.47E+07 3.47E+07 3.25E+07 3.00E+07 9.93E+06 8.02E+05 9.47E+00 0.00E+00 0.00E+00 Te129 3.29E+07 3.29E+07 3.26E+07 3.19E+07 1.53E+07 5.10E+06 3.95E+06 2.31E+06 6.71E+05 1.05E+05 2.29E+03 6.54E-04 Te129m 6.66E+06 6.66E+06 6.66E+06 6.66E+06 6.64E+06 6.56E+06 6.17E+06 3.61E+06 1.05E+06 1.63E+05 3.58E+03 1.02E-03 Xe129m 5.16E+03 5.16E+03 5.15E+03 5.14E+03 5.03E+03 4.78E+03 3.77E+03 4.97E+02 4.62E+00 4.15E-03 2.21E-09 0.00E+00 I130 2.64E+06 2.64E+06 2.58E+06 2.51E+06 1.70E+06 6.91E+05 1.22E+04 7.79E-12 0.00E+00 Te131 9.09E+07 9.09E+07 6.98E+07 4.42E+07 4.03E+06 2.79E+06 5.28E+05 2.90E-01 1.02E-15 0.00E+00 Te131m 2.15E+07 2.15E+07 2.12E+07 2.11E+07 1.80E+07 1.24E+07 2.35E+06 1.28E+00 4.56E-15 0.00E+00 I131 1.07E+08 1.07E+08 1.07E+08 1.07E+08 1.05E+08 1.00E+08 7.87E+07 8.40E+06 4.76E+04 2.03E+01 2.35E-06 0.00E+00 Xe131m 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.45E+06 1.38E+06 5.05E+05 2.00E+04 1.11E+02 2.29E-03 7.54E-22 Te132 1.54E+08 1.54E+08 1.54E+08 1.53E+08 1.44E+08 1.25E+08 6.59E+07 2.61E+05 7.46E-01 3.61E-09 2.76E-26 0.00E+00 I132 1.57E+08 1.57E+08 1.57E+08 1.56E+08 1.48E+08 1.28E+08 6.78E+07 2.69E+05 7.72E-01 3.71E-09 2.84E-26 0.00E+00 Cs132 4.35E+03 4.35E+03 4.35E+03 4.33E+03 4.20E+03 3.91E+03 2.83E+03 1.76E+02 2.87E-01 1.89E-05 4.67E-14 0.00E+00 Te133m 9.86E+07 9.86E+07 6.82E+07 4.69E+07 2.44E+05 1.49E+00 0.00E+00 I133 2.22E+08 2.22E+08 2.20E+08 2.18E+08 1.74E+08 1.02E+08 9.24E+06 8.63E-03 1.25E-23 0.00E+00 Xe133 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.06E+08 1.51E+08 4.99E+06 1.80E+03 1.22E-02 2.82E-13 0.00E+00 Xe133m 6.99E+06 6.99E+06 6.99E+06 6.98E+06 6.86E+06 6.29E+06 3.00E+06 8.48E+02 4.78E-06 2.03E-18 0.00E+00 I134 2.45E+08 2.45E+08 2.15E+08 1.75E+08 1.53E+06 5.71E+00 0.00E+00 Cs134 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.29E+07 2.24E+07 2.12E+07 1.95E+07 1.64E+07 8.40E+06 Cs134m 4.81E+06 4.81E+06 4.27E+06 3.79E+06 7.16E+05 1.59E+04 5.69E-04 0.00E+00 I135 2.11E+08 2.11E+08 2.00E+08 1.90E+08 9.09E+07 1.68E+07 8.40E+03 0.00E+00 Xe135 7.03E+07 7.03E+07 7.56E+07 8.02E+07 1.01E+08 5.62E+07 4.00E+05 1.18E-15 0.00E+00 Xe135m 4.63E+07 4.63E+07 3.58E+07 3.18E+07 1.48E+07 2.74E+06 1.38E+03 0.00E+00 Ba135m 4.19E+04 4.19E+04 4.13E+04 4.09E+04 3.45E+04 2.35E+04 4.12E+03 1.17E-03 9.17E-19 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 5 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Cs136 7.34E+06 7.34E+06 7.33E+06 7.33E+06 7.21E+06 6.96E+06 5.94E+06 1.51E+06 6.41E+04 5.60E+02 3.24E-02 6.31E-19 Ba136m 8.40E+05 8.25E+05 8.17E+05 8.17E+05 8.10E+05 7.79E+05 6.66E+05 1.70E+05 7.18E+03 6.27E+01 3.63E-03 7.07E-20 Cs137 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 172E+07 1.70E+07 1.62E+07 Ba137m 1.65E+07 1.65E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.63E+07 1.63E+07 1.62E+07 1.60E+07 1.53E+07 Cs138 2.05E+08 2.05E+08 1.51E+08 8.94E+07 1.15E+04 1.21E-05 0.00E+00 Ba139 1.95E+08 1.95E+08 1.70E+08 1.34E+08 4.32E+06 1.66E+03 7.15E-13 0.00E+00 Ba140 1.96E+08 1.96E+08 1.96E+08 1.96E+08 1.93E+08 1.86E+08 1.57E+08 3.84E+07 1.47E+06 1.11E+04 4.69E-01 2.69E-18 La140 2.09E+08 2.09E+08 2.09E+08 2.09E+08 2.08E+08 2.03E+08 1.78E+08 4.42E+07 1.70E+06 1.28E+04 5.40E-01 3.10E-18 La141 1.78E+08 1.78E+08 1.72E+08 1.60E+08 4.70E+07 2.77E+06 8.17E+00 0.00E+00 Ce141 1.80E+08 1.80E+08 1.80E+08 1.80E+08 1.79E+08 1.76E+08 1.66E+08 9.55E+07 2.64E+07 3.88E+06 7.46E+04 1.28E-02 La142 1.74+08 1.74E+08 1.54E+08 1.24E+08 5.10E+06 3.42E+03 1.82E-11 0.00E+00 Pr142 7.33E+06 7.33E+06 7.20E+06 7.07E+06 5.49E+06 3.07E+06 2.26E+05 3.38E-05 0.00E+00 Ce143 1.67E+08 1.67E+08 1.66E+08 1.64E+08 1.42E+08 1.02E+08 2.24E+07 4.55E+01 3.32E-12 0.00E+00 Pr143 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.60E+08 1.44E+08 3.90E+07 1.82E+06 1.83E+04 1.41E+00 8.71E-17 Ce144 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.50E+08 1.41E+08 1.21E+08 9.78E+07 6.22E+07 1.05E+07 Pr144 1.52E+08 1.52E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.50E+08 1.41E+08 1.21E+08 9.78E+07 6.22E+07 1.05E+07 Pr144m 2.12E+06 2.12E+06 2.12E+06 2.12E+06 2.12E+06 2.11E+06 2.09E+06 1.97E+06 1.70E+06 1.37E+06 8.71E+05 1.47E+05 Pr145 1.13E+08 1.13E+08 1.08E+08 1.02E+08 4.53E+07 7.11E+06 1.70E+03 0.00E+00 Nd147 7.24E+07 7.24E+07 7.23E+07 7.22E+07 7.09E+07 6.80E+07 5.62E+07 1.09E+07 2.47E+05 8.40E+02 7.02E-03 6.59E-23 Pm147 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.52E+07 2.52E+07 2.43E+07 2.28E+07 1.99E+07 1.18E+07 Pm148 2.02E+07 2.02E+07 2.02E+07 2.01E+07 1.94E+07 1.78E+07 1.21E+07 5.38E+05 4.55E+04 1.00E+04 4.46E+02 2.11E-03 Pm148m 3.88E+06 3.88E+06 3.88E+06 3.87E+06 3.86E+06 3.81E+06 3.63E+06 2.35E+06 8.56E+05 1.89E+05 8.40E+03 4.00E-02 Nd149 4.16E+07 4.16E+07 3.47E+07 2.84E+07 1.70E+06 2.75E+03 7.49E-10 0.00E+00 Pm149 6.46E+07 6.46E+07 6.44E+07 6.42E+07 5.94E+07 4.82E+07 1.89E+07 5.45E+03 3.72E-05 2.09E-17 0.00E+00 Pm150 5.53E+05 5.53E+05 4.87E+05 4.27E+05 6.99E+04 1.12E+03 9.09E-06 0.00E+00 Pm151 2.17E+07 2.17E+07 2.15E+07 2.13E+07 1.80E+07 1.21E+07 2.10E+06 5.07E-01 2.73E-16 0.00E+00 Sm151 6.80E+04 6.80E+04 6.80E+04 6.80E+04 6.81E+04 6.83E+04 6.87E+04 6.87E+04 6.86E+04 6.85E+04 6.82E+04 6.72E+04 Eu152m 2.48E+04 2.48E+04 2.38E+04 2.30E+04 1.37E+04 4.15E+03 1.96E+01 1.37E-19 0.00E+00 Sm153 5.31E+07 5.31E+07 5.27E+07 5.23E+07 4.71E+07 3.71E+07 1.26E+07 1.10E+03 4.73E-07 4.21E-21 0.00E+00 Gd153 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.05E+05 7.49E+05 6.30E+05 4.87E+05 2.86E+05 3.52E+04 Eu154 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.04E+06 1.02E+06 9.77E+05 8.29E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 6 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Eu155 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.34E+05 4.29E+05 4.18E+05 4.04E+05 3.75E+05 2.78E+05 Sm156 2.57E+06 2.57E+06 2.48E+06 2.38E+06 1.42E+06 4.38E+05 2.16E+03 2.25E-17 0.00E+00 Eu156 2.74E+07 2.74E+07 2.74E+07 2.73E+07 2.70E+07 2.62E+07 2.29E+07 6.98E+06 4.51E+05 7.41E+03 1.58E+00 5.20E-15 Gd159 2.74E+07 2.74E+07 2.69E+07 2.65E+07 2.04E+07 1.12E+07 7.61E+05 5.77E-05 0.00E+00 Tb160 7.20E+06 7.20E+06 7.19E+06 7.19E+06 7.17E+06 7.13E+06 6.92E+06 5.40E+06 3.03E+06 1.28E+06 2.17E+05 1.97E+02 Tb161 4.33E+06 4.33E+06 4.32E+06 4.31E+06 4.19E+06 3.92E+06 2.90E+06 2.13E+05 5.14E+02 6.09E-02 5.04E-10 0.00E+00 Dy165 5.38E+05 5.38E+05 4.66E+05 4.02E+05 5.03E+04 4.34E+02 2.25E-07 0.00E+00 Dy166 2.09E+03 2.09E+03 2.09E+03 2.08E+03 1.96E+03 1.71E+03 9.24E+02 4.63E+00 2.26E-05 2.44E-13 1.74E-29 0.00E+00 Ho166 8.58E+04 8.58E+04 8.41E+04 8.33E+04 6.99E+04 4.69E+04 8.43E+03 8.10E+00 3.95E-05 4.26E-13 1.74E-29 0.00E+00 Hf175 3.62E+03 3.62E+03 3.62E+03 3.62E+03 3.61E+03 3.58E+03 3.48E+03 2.69E+03 1.49E+03 6.10E+02 9.78E+01 7.05E-02 Lu176m 1.64E+03 1.64E+03 1.50E+03 1.36E+03 3.58E+02 1.70E+01 1.86E-05 0.00E+00 Lu177 6.84E+02 6.84E+02 6.82E+02 6.81E+02 6.60E+02 6.17E+02 4.52E+02 3.12E+01 3.46E-01 1.93E-01 8.63E-02 3.72E-03 Hf180m 1.12E+04 1.12E+04 1.05E+04 9.86E+03 4.10E+03 5.45E+02 6.24E-02 0.00E+00 Hf181 2.38E+05 2.38E+05 2.38E+05 2.38E+05 2.37E+05 2.35E+05 2.23E+05 1.46E+05 5.47E+04 1.25E+04 6.07E+02 3.93E-03 W181 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.41E+03 1.39E+03 1.20E+03 8.48E+02 5.09E+02 1.76E+02 2.70E+00 Ta182 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.57E+04 2.53E+04 2.16E+04 1.51E+04 8.79E+03 2.87E+03 3.51E+01 Ta183 5.46E+04 5.46E+04 5.45E+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 0.00E+00 W183m 6.40E+04 6.29E+04 5.45E+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 0.00E+00 W185 3.52E+04 3.52E+04 3.52E+04 3.52E+04 3.51E+04 3.49E+04 3.39E+04 2.67E+04 1.54E+04 6.69E+03 1.21E+03 1.43E+00 Re186 2.58E+04 2.58E+04 2.57E+04 2.56E+04 2.43E+04 2.15E+04 1.24E+04 1.05E+02 1.73E-03 1.16E-10 1.99E-25 0.00E+00 W187 4.42E+05 4.42E+05 4.36E+05 4.30E+05 3.51E+05 2.21E+05 2.74E+04 3.77E-04 2.75E-22 0.00E+00 W188 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.94E+03 1.89E+03 1.45E+03 8.02E+02 3.25E+02 5.12E+01 3.49E-02 Re188 1.73E+05 1.73E+05 1.72E+05 1.69E+05 1.28E+05 6.75E+04 5.37E+03 1.47E+03 8.10E+02 3.29E+02 5.17E+01 3.52E-02 Os191 7.72E+01 7.72E+01 7.64E+01 7.64E+01 7.63E+01 7.49E+01 6.59E+01 2.05E+01 1.38E+00 2.40E-02 5.75E-06 3.06E-20 Os191m 5.82E+01 5.82E+01 5.67E+01 5.52E+01 3.81E+01 1.63E+01 3.62E-01 1.66E-15 0.00E+00 Ir192 3.01E+01 3.01E+01 3.01E+01 3.01E+01 3.00E+01 2.98E+01 2.90E+01 2.27E+01 1.29E+01 5.55E+00 9.78E-01 1.06E-03 Np236m 4.84E+02 4.84E+02 4.77E+02 4.70E+02 3.78E+02 2.31E+02 2.51E+01 1.13E-07 6.12E-27 0.00E+00 U237 1.00E+08 1.00E+08 1.00E+08 9.93E+07 9.70E+07 9.02E+07 6.64E+07 4.60E+06 1.02E+04 4.48E+02 4.37E+02 3.97E+02 Pu237 6.71E+02 6.71E+02 6.70E+02 6.70E+02 6.67E+02 6.60E+02 6.30E+02 4.23E+02 1.69E+02 4.23E+01 2.47E+00 3.35E-05 Np238 4.70E+07 4.70E+07 4.67E+07 4.64E+07 4.22E+07 3.39E+07 1.27E+07 2.55E+03 7.44E+00 7.43E+00 7.41E+00 7.34E+00 Pu238 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.57E+05 4.58E+05 4.63E+05 4.69E+05 4.75E+05 4.82E+05 4.81E+05 SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 7 of 7 Table 12.2-9A1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Np239 2.12E+09 2.12E+09 2.12E+09 2.11E+09 1.93E+09 1.59E+09 6.59E+08 3.16E+05 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Pu239 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.84E+04 4.87E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 Np240 3.93E+06 3.93E+06 2.81E+06 2.01E+06 1.82E+04 3.90E-01 6.20E-16 7.40E-16 1.01E-15 1.41E-15 2.25E-15 5.53E-15 Pu240 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 Pu241 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.91E+07 1.89E+07 1.87E+07 1.83E+07 1.66E+07 Am241 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.55E+04 2.57E+04 2.79E+04 3.29E+04 4.02E+04 5.51E+04 1.11E+05 Am242m 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.65E+03 1.65E+03 1.65E+03 1.63E+03 Am242 1.14E+07 1.14E+07 1.12E+07 1.09E+07 8.10E+06 4.04E+06 1.81E+05 1.65E+03 1.64E+03 1.64E+03 1.64E+03 1.62E+03 Cm242 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.66E+06 6.59E+06 5.91E+06 4.58E+06 3.12E+06 1.42E+06 6.47E+04 Pu243 4.19E+07 4.19E+07 3.90E+07 3.64E+07 1.37E+07 1.46E+06 6.17E+01 3.93E-05 3.93E-05 3.93E-05 3.93E-05 3.93E-05 Am243 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Cm243 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.86E+03 2.84E+03 2.80E+03 2.67E+03 Am244 1.38E+07 1.38E+07 1.33E+07 1.29E+07 7.95E+06 2.66E+06 1.90E+04 4.79E-15 0.00E+00 Cm244 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.89E+05 3.87E+05 3.83E+05 3.76E+05 3.48E+05

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 1 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr H 3 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.41E+02 1.39E+02 1.35E+02 1.21E+02 Na 24 4.16E-01 4.16E-01 4.07E-01 3.97E-01 2.85E-01 1.34E-01 4.45E-02 6.81E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cr 51 9.20E+03 9.20E+03 9.19E+03 9.19E+03 9.12E+03 8.97E+03 8.32E+03 4.34E+03 9.68E+02 1.02E+02 9.90E-01 1.15E-08 Mn 54 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.64E+02 6.59E+02 6.22E+02 5.45E+02 4.46E+02 2.96E+02 5.84E+01 Fe 55 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.64E+03 3.58E+03 3.43E+03 3.22E+02 2.84E+03 1.71E+03 Mn 56 1.95E+04 1.95E+04 1.70E+04 1.49E+04 2.27E+03 3.07E+01 1.21E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Co 58 8.15E+02 8.15E+02 8.14E+02 8.14E+02 8.12E+02 8.07E+02 7.83E+02 6.07E+02 3.38E+02 1.40E+02 2.29E+01 1.82E-02 Fe 59 2.18E+02 2.18E+02 2.18E+02 2.18E+02 2.17E+02 2.15E+02 2.05E+02 1.37E+02 5.37E+01 1.32E+01 7.38E-01 8.44E-06 Ni 59 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 Co 60 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.26E+02 6.21E+02 6.07E+02 5.88E+02 5.50E+02 4.23E+02 Ni 63 7.67E+01 7.67E+01 7.67E+01 7.67E+01 7.66E+01 7.66E+01 7.66E+01 7.66E+01 7.65E+01 7.64E+01 7.61E+01 7.51E+01 Cu 64 1.01E+00 1.01E+00 9.85E-01 9.58E-01 6.54E-01 2.73E-01 5.37E-03 8.69E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ni 65 9.17E+01 9.17E+01 7.99E+01 6.97E+01 1.02E+01 1.25E-01 3.12E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ga 73 6.64E+00 6.64E+00 6.20E+00 5.77E+00 2.13E+00 2.17E-01 7.55E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ge 73m 6.58E+00 6.56E+00 6.12E+00 5.69E+00 2.10E+00 2.14E-01 7.46E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 76 5.54E+00 5.54E+00 5.47E+00 5.40E+00 4.49E+00 2.95E+00 4.42E-01 3.23E-08 1.09E-24 0.00E+00 0.00E+00 0.00E+00 Ge 77 8.69E+01 8.69E+01 8.44E+01 8.18E+01 5.32E+-01 2.00E+01 2.41E-01 5.74E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 77 2.64E+02 2.64E+02 2.63E+02 2.61E+02 2.38E+02 1.87E+02 5.40E+01 7.87E-04 5.41E-15 9.13E-32 0.00E+00 0.00E+00 Ge 78 9.12E+02 9.12E+02 7.20E+02 5.69E+02 2.08E+01 1.08E-02 1.80E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 78 9.26E+02 9.26E+02 9.02E+02 8.48E+02 1.04E+02 1.57E-01 1.75E-15 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br 82 7.95E+02 7.95E+02 7.89E+02 7.81E+02 6.81E+02 4.97E+02 1.21E+02 5.78E-04 3.05E-16 0.00E+00 0.00E+00 0.00E+00 Br 83 1.45E+04 1.45E+04 1.32E+04 1.17E+04 1.57E+03 1.55E+01 1.44E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 83m 1.47E+04 1.47E+04 1.46E+04 1.42E+04 4.10E+03 5.93E+01 6.08E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 85 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.48E+03 2.45E+03 2.41E+03 2.33E+03 2.05E+03 Kr 85m 2.88E+04 2.88E+04 2.69E+04 2.49E+04 8.44E+03 7.10E+02 1.03E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 86 4.64E+02 4.64E+02 4.63E+02 4.63E+02 4.58E+02 4.47E+02 4.00E+02 1.52E+02 1.63E+01 5.73E-01 5.83E-04 9.22E-16 Kr 87 5.70E+04 5.70E+04 4.38E+04 3.34E+04 7.35E+02 1.20E-01 1.10E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 88 7.80E+04 7.80E+04 6.91E+04 6.11E+04 1.11E+04 2.23E+02 5.16E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 2 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Rb 88 8.02E+04 8.02E+04 7.50E+04 6.76E+04 1.24E+04 2.48E+02 5.76E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 89 8.71E+01 8.71E+01 8.67E+01 8.64E+01 8.12E+01 7.05E+01 3.73E+01 1.50E-01 4.48E-07 2.30E-15 2.28E-32 0.00E+00 Sr 89 1.07E+05 1.07E+05 1.07E+05 1.07E+05 1.06E+05 1.05E+05 1.01E+05 7.07E+04 3.11E+04 9.04E+03 7.13E+02 3.19E-02 Y 89m 3.10E+02 3.01E+02 9.62E+01 9.58E+01 9.06E+01 7.98E+01 4.64E+01 6.73E+00 2.89E+00 8.41E-01 6.63E-02 2.96E-06 Sr 90 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.23E+04 2.23E+04 2.21E+04 2.18E+04 2.08E+04 Y 90 2.36E+04 2.36E+04 2.36E+04 2.36E+04 2.35E+04 2.33E+04 2.28E+04 2.24E+04 2.23E+04 2.21E+04 2.19E+04 2.08E+04 Sr 91 1.41E+05 1.41E+05 1.36E+05 1.31E+05 7.86E+04 2.45E+04 1.30E-02 2.40E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 91 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.43E+05 1.39E+05 1.02E+05 5.00E+04 1.72E+04 1.92E+03 3.35E-01 Y 91m 8.15E+04 8.15E+04 8.10E+04 7.97E+04 4.99E+04 1.56E+04 8.23E+01 1.52E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr 92 1.55E+05 1.55E+05 1.36E+05 1.20E+05 2.00E+04 3.34E+02 3.36E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 92 1.56E+05 1.56E+05 1.55E+05 1.52E+05 7.28E+04 4.93E+03 4.51E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 93 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 Y 93 1.23E+05 1.23E+05 1.20E+05 1.16E+05 7.19E+04 2.40E+04 1.71E+02 4.32E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 95 2.33E+05 2.33E+05 2.33E+05 2.33E+05 2.32E+05 2.30E+05 2.23E+05 1.68E+05 8.79E+04 3.32E+04 4.46E+03 1.64E+00 Nb 95 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.33E+05 2.17E+05 1.47E+05 6.51E+04 9.62E+03 3.61E+00 Nb 95m 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.55E+03 1.98E+03 1.03E+03 3.90E+02 5.25E+01 1.93E-02 Nb 96 4.78E+02 4.78E+02 4.71E+02 4.64E+02 3.78E+02 2.35E+02 2.77E+01 2.50E+07 6.80E-26 0.00E+00 0.00E+00 0.00E+00 Zr 97 2.43E+05 2.43E+05 2.38E+05 2.34E+05 1.75E+05 9.09E+04 4.74E+03 3.63E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb 97 2.45E+05 2.45E+05 2.44E+05 2.41E+05 1.88E+05 9.13E+04 4.76E+03 3.92E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb 97m 2.30E+05 2.30E+05 2.26E+05 2.22E+05 1.66E+05 8.62E+04 4.50E+03 3.46E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Mo 99 2.61E+05 2.61E+05 2.59E+05 2.58E+05 2.40E+-5 2.02E+05 9.50E+04 1.35E+02 3.60E-05 4.96E-15 0.00E+00 0.00E+00 Tc 99 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 Tc 99m 2.32E+05 2.32E+05 2.32E+05 2.31E+05 2.24E+05 1.95E+05 9.20E+04 1.30E+02 3.48E-05 4.80E-15 0.00E+00 0.00E+00 Ru103 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.47E+05 2.44E+05 2.31E+05 1.46E+05 5.07E+04 1.03E+04 3.93E+02 9.83E-04 Rh103m 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.46E+05 2.44E+05 2.31E+05 1.46E+05 5.06E+04 1.03E+04 3.92E+02 9.82E-04 Ru105 1.89E+05 1.89E+05 1.80E+05 1.66E+-5 5.58E+04 4.59E+03 6.01E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rh105 1.74E+05 1.74E+05 1.74E+05 1.74E+05 1.64E+05 1.25E+05 3.07E+04 1.50E-01 8.25E-14 2.28E-32 0.00E+00 0.00E+00 Rh105m 5.36E+04 5.36E+04 5.12E+04 4.74E+04 1.59E+04 1.31E+03 1.71E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 3 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Ru106 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.25E+05 1.19E+05 1.06E+05 8.99E+04 6.36E+04 1.63E+04 Rh106 1.36E+05 1.36E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.25E+05 1.19E+05 1.06E+05 8.99E+04 6.36E+04 1.63E+04 Rh106m 4.69E+03 4.69E+03 4.00E+03 3.41E+03 3.63E+02 2.17E+00 2.16E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag109m 7.74E+04 7.74E+04 7.57E+04 7.38E+04 5.18E+04 2.30E+04 6.03E+02 3.55E-04 3.24E-04 2.83E-04 2.15E-04 7.18E-05 Ag110 4.66E+04 4.53E+04 1.74E+01 1.74E+01 1.74E+01 1.74E+01 1.73+01 1.60E+01 1.36E+01 1.06E+01 6.33E+00 8.34E-01 Ag110m 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.27+03 1.18E+03 9.99E+02 7.78E+02 4.65E+02 6.13E+01 Pd111 1.14E+04 1.14E+-04 4.94E+03 2.21E+03 1.35E+02 1.80E+01 2.06E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pd111m 4.71E+02 4.71E+02 4.43E+02 4.16E+02 1.72E+02 2.29E+01 2.62E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag111 1.16E+04 1.16E+04 1.16E+04 1.16E+04 1.13+04 1.06E+04 8.01E+03 7.13E+02 2.68E+00 6.20E-04 2.03E-11 0.00E+00 Ag111m 1.16E+04 1.16E+04 5.22E+03 2.38E+03 1.68E+02 2.24E+01 2.57E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pd112 5.01E+03 5.01E+03 4.93E+03 4.85E+03 3.85E+03 2.27E+03 2.12E+02 2.51E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag112 5.03E+03 5.03E+03 5.02E+03 5.01E+03 4.38E+03 2.67E+03 2.49E+02 2.95E-07 7.45E-28 0.00E+00 0.00E+00 0.00E+00 In113m 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.24E+02 5.14E+02 4.40E+02 3.06E+02 1.78E+02 5.84E+01 7.17E-01 Sn113 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.23E+02 5.14E+02 4.40E+02 3.06E+02 1.78E+02 5,84E+01 7.17E-01 Ag113 2.74E+03 2.74E+03 2.58E+03 2.42E+03 9.81E+02 1.24E+02 1.14E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cd113m 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.11E+01 1.10E+01 1.07E+01 9.69E+00 In114 1.36E+02 1.35E+02 8.61E+01 8.61E+01 8.57E+01 8.49E+01 8.14E+01 5.66E+01 2.44E+01 6.93E+00 5.18E-01 1.88E-05 In114m 9.00E+01 9.00E+01 8.99E+01 8.99E+01 8.95E+01 8.87E+01 8.51E+01 5.91E+01 2.55E+01 7.24E+00 5.41E-01 1.96E-05 Cd115 1.43E+03 1.43E+03 1.42E+03 1.41E+03 1.29E+03 1.05E+03 4.13E+02 1.27E-01 9.90E-10 6.85E-22 0.00E+00 0.00E+00 Cd115m 7.03E+01 7.03E+01 7.03E+01 7.03E+01 7.00E+01 6.93E+01 6.61E+01 4.41E+01 1.74E+01 4.29E+00 2.41E-01 2.82E-06 In-115m 1.43E+03 1.43E+03 1.43E+03 1.43E+03 1.37E+03 1.14E+03 4.50E+02 1.43EE-01 1.92E-03 4.74E-04 2.66E-05 3.12E-10 Cd117 1.34E+03 1.34E+03 1.17E+03 1.02E+03 1.46E+02 1.69E+00 3.34E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cd117m 3.09E+02 3.09E+02 2.79E+02 2.52E+02 5.95E+01 2.19E+00 7.81E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 In117 1.00E+03 1.00E+03 9.94E+02 9.67E+02 3.06E+02 6.25E+00 1.01E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 In117m 1.23E+03 1.23E+03 1.21+03 1.18E+03 3.53E+02 6.24E+00 4.16E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sn117m 3.80E+03 3.80E+03 3.80E+03 3.79E+03 3.74E+03 3.61E+03 3.11E+03 8.24E+02 3.87E+01 3.94E-01 3.13E-05 2.12E-21 Sn119m 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.84E+03 3.81E+03 3.58E+03 3.11E+03 2.52E+03 1.62E+03 2.88E+02 Sn121 2.31E+03 2.31E+03 2.28E+03 2.25E+03 1.89E+03 1.25E+03 1.99E+02 8.96E-01 8.93E-01 8.91E-01 8.86E-01 8.63E-01

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 4 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Sn121m 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.96E-01 3.95E-01 3.92E-01 3.83E-01 Sb122 5.06E+02 5.06E+02 5.02E+02 5.00E+02 4.64E+02 3.91E+02 1.81E+02 2.28E-01 4.68E-08 4.33E-18 0.00E+00 0.00E+00 Te123m 8.01E+00 8.01E+00 8.00E+00 8.00E+00 8.00E+00 7.96E+00 7.82E+00 6.74E+00 4.76E+00 2.83E+00 9.65E-01 1.40E-02 Sn123 1.97E+02 1.97E+02 1.97E+02 1.97E+02 1.96E+02 1.95E+02 1.92E+02 1.67E+02 1.21E+02 7.48E+01 2.76E+01 5.49E-01 Sn123m 1.34E+03 1.34E+03 8.03E+02 4.78E+-2 3.35E-01 2.07E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb124 2.39E+02 2.39E+02 2.39E+02 2.39E+02 2.38E+02 2.36E+02 2.29E+02 1.69E+02 8.48E+01 3.01E+01 3.56E+00 7.92E-04 Sn125 8.09E+02 8.09E+02 8.08E+02 8.07E+02 7.89E+02 7.52E+02 6.07E+02 9.36E+01 1.25E+00 1.94E-03 3.18E-09 0.00E+00 Sn125m 3.08E+03 3.08E+03 3.49E+02 3.93E+01 2.06E-12 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb125 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.47E+03 2.43E+03 2.33E+03 2.19E+03 1.92E+03 1.16E+03 Te125m 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.55E+02 5.30E+02 4.69E+02 2.83E+02 Sb126 8.31E+01 8.31E+01 8.31E+01 8.29E+01 8.16E+01 7.86E+01 6.65E+01 1.56E+01 5.67E-01 2.88E-02 2.53E-02 2.53E-02 Sb126m 9.43E+01 9.42E+01 3.17E+01 1.07E+01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 Sn127 5.62E+03 5.62E+03 4.77E+03 4.04E+03 4.01E+02 2.04E+00 9.74E-11 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb127 1.39E+04 1.39E+04 1.38E+04 1.38E+04 1.32E+04 1.17E+04 6.81E+03 6.31E+01 1.28E-03 1.18E-10 3.84E-25 0.00E+00 Te127 1.38E+04 1.38E+04 1.38E+04 1.38E+04 1.36E+04 1.28E+04 8.54E+03 2.04E+03 1.35E+03 7.63E+02 2.35E+02 2.26E+00 Te127m 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.36E+03 2.34E+03 2.02E+03 1.38E+03 7.79E+02 2.40E+02 2.31E+00 Sb128 2.32E+03 2.32E+03 2.26E+03 2.20E+03 1.31E+03 3.84E+02 1.51E+00 2.15E-21 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb129 4.77E+04 4.77E+04 4.46E+04 4.12E+04 1.37E+04 1.10E+03 1.31E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te129 4.54E+04 4.54E+04 4.50E+04 4.39E+04 2.10E+04 7.03E+03 5.45E+03 3.19E+03 9.25E+02 1.45E+02 3.16E+00 9.02E-07 Te129m 9.19E+03 9.19E+03 9.19E+03 9.19E+03 9.17E+03 9.05E+03 8.51E+03 4.98E+03 1.44E+03 2.25E+02 4.94E+00 1.41E-06 Xe129m 2.23E+01 2.23E+01 2.22E+01 2.22E+01 2.17E+01 2.06E+01 1.63E+01 2.15E+00 2.00E-02 1.79E-05 9.55E-12 0.00E+00 I130 6.42E+03 6.42E+03 6.27E+03 6.10E+03 4.12E+03 1.68E+03 2.97E+01 1.89E-14 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te131 1.18E+05 1.18E+05 9.07E+04 5.76E+04 5.70E+03 3.94E+03 7.46E+02 4.09E-04 1.45E-18 0.00E+00 0.00E+00 0.00E+00 Te131m 3.04E+04 3.04E+04 3.01E+04 2.98E+04 2.53E+04 1.75E+04 3.32E+03 1.82E-03 6.45E-18 0.00E+00 0.00E+00 0.00E+00 I131 1.41E+05 1.41E+05 1.41E+05 1.41E+05 1.38E+05 1.32E+05 1.04E+05 1.11E+04 6.28E+01 2.68E-02 3.11E-09 0.00E+00 Xe131m 2.15E+03 2.15E+03 2.15E+03 2.15E+03 2.13E+03 2.11E+03 1.97E+03 7.02E+02 2.75E+01 1.52E-01 3.14E-06 1.03E-24 Te132 2.01E+05 2.01E+05 2.01E+05 2.00E+05 1.88E+05 1.63E+05 8.61E+04 3.41E+02 9.75E-04 4.71E-12 3.61E-29 0.00E+00 I132 2.06E+05 2.06E+05 2.05E+05 2.05E+05 1.93E+05 1.68E+05 8.87E+04 3.51E+02 1.00E-03 4.86E-12 3.72E-29 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 5 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Cs132 9.34E+00 9.34E+00 9.32E+00 9.30E+00 9.01E+00 8.39E+00 6.09E+00 3.77E-01 6.15E-04 4.05E-08 1.00E-16 0.00E+00 Te133m 1.24E+05 1.24E+05 8.59E+04 5.90E+04 3.08E+02 1.87E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I133 2.85E+05 2.85E+05 2.83E+05 2.80E+05 2.24E+05 1.32E+05 1.19E-05 1.11E-05 1.60E-26 0.00E+00 0.00E+00 0.00E+00 Xe133 2.74E+05 2.74E+05 2.74E+05 2.74E+05 2.73E+05 2.65E+05 1.95E+05 6.43E+03 2.31E+00 1.57E-05 3.63E-16 0.00E+00 Xe133m 9.22E+03 9.22E+03 9.21E+03 9.20E+03 9.03E+03 8.25E+03 3.93E+03 1.11E+00 6.25E-09 2.65E-21 0.00E+00 0.00E+00 I134 3.12E+05 3.12E+05 2.72E+05 2.20E+05 1.90E+03 7.08E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs134 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.56E+04 5.43E+04 5.14E+04 4.73E+04 3.99E+04 2.04E+04 CS134m 9.91E+03 9.91E+03 8.80E+03 7.81E+03 1.47E+03 3.27E+01 1.17E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I135 2.74E+05 2.74E+05 2.60E+05 2.46E+05 1.18E+05 2.18E+04 1.09E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe135 7.50E+04 7.50E+04 8.26E+04 8.90E+04 1.22E+05 7.02E+04 5.08E+02 1.50E-18 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe135m 6.25E+04 6.25E+04 4.70E+04 4.14E+04 1.92E+04 3.55E+03 1.78E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba135m 1.82E+02 1.82E+02 1.80E+02 1.78E+02 1.50E+02 1.02E+02 1.79E+01 5.10E-06 3.98E-21 0.00E+00 0.00E+00 0.00E+00 Cs136 1.54E+04 1.54E+04 1.54E+04 1.53E+04 1.51E+04 1.46E+04 1.25E+04 3.17E+03 1.34E+02 1.17E+00 6.79E-05 1.32E-21 Ba136m 1.77E+03 1.73E+03 1.72E+03 1.72E+03 1.69E+03 1.63E+03 1.40E+03 3.55E+02 1.50E+01 1.31E-01 7.61E-06 1.48E-22 Cs137 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.28E+04 3.27E+04 3.23E+04 3.08E+04 Ba137m 3.14E+04 3.14E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.11E+04 3.10E+04 3.08E+04 3.05E+04 2.91E+04 Cs138 2.58E+05 2.58E+05 1.90E+05 1.12E+05 1.44E+01 1.53E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba139 2.46E+05 2.46E+05 2.13E+05 1.69E+05 5.43E+05 2.09E+00 9.00E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba140 2.47E+05 2.47E+05 2.46E+05 2.46E+05 2.42E+05 2.34E+05 1.99E+05 4.83E+04 1.85E+03 1.39E+01 5.91E-04 3.39E-21 La140 2.70E+05 2.70E+05 2.70E+05 2.70E+05 2.67E+05 2.60E+05 2.26E+05 5.57E+04 2.13E+03 1.60E+01 6.81E-04 3.91E-21 La141 2.24E+05 2.24E+05 2.16E+05 2.01E+05 5.90E+04 3.48E+03 1.03E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ce141 2.25E+05 2.25E+05 2.25E+05 2.25E+05 2.24E+05 2.21E+05 2.08E+05 1.19E+05 3.31E+04 4.86E+03 9.35E+01 1.60E-05 La142 2.17E+05 2.17E+05 1.90E+05 1.54E+05 6.33E+03 4.25E+00 2.26E-14 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pr142 1.72E+04 1.72E+04 1.69E+04 1.66E+04 1.29E+04 7.20E+03 5.29E+02 7.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ce143 2.04E+05 2.04E+05 2.03E+05 2.01E+05 1.73E+05 1.24E+05 2.73E+04 5.55E-02 4.06E-15 0.00E+00 0.00E+00 0.00E+00 Pr143 1.97E+05 1.97E+05 1.97E+05 1.97E+05 1.96E+05 1.95E+05 1.76E+05 4.74E+04 2.21E+03 2.22E+01 1.72E-03 1.06E-19 Ce144 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E+05 1.76E+05 1.52E+05 1.22E+05 7.77E+04 1.31E+04 Pr144 1.90E+05 1.90E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E+05 1.76E+05 1.52E+05 1.22E+05 7.77E+04 1.31E+04

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 6 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0d 1 yr 3 yr Pr144m 2.65E+03 2.65E+03 2.64E+03 2.64E+03 2.64E+03 2.64E+03 2.62E+03 2.46E+03 2.12E+03 1.71E+03 1.09E+03 1.84E+02 Pr145 1.40E+05 1.40E+05 1.33E+05 1.26E+05 5.59E+04 8.75E+03 2.09E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nd147 9.29E+04 9.29E+04 9.28E+04 9.27E+04 9.10E+04 8.73E+04 7.22E+04 1.40E+04 3.17E+02 1.08E+00 9.01E-06 8.47E-26 Pm147 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.15E+04 3.16E+04 3.16E+04 3.05E+04 2.85E+04 2.50E+04 1.47E+04 Pm148 2.88E+04 2.88E+04 2.87E+04 2.87E+04 2.76E+04 2.54E+04 1.73E+04 7.46E+02 5.61E+01 1.23E+01 5.50E-01 2.60E-06 Pm148m 4.78E+03 4.78E+03 4.78E+03 4.78E+03 4.75E+03 4.70E+03 4.47E+03 2.89E+03 1.05E+03 2.33E+02 1.04E+01 4.92E-05 Nd149 5.70E+04 5.70E+04 4.76E+04 3.90E+04 2.34E+03 3.77E+00 1.03E-12 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm149 9.22E+04 9.22E+04 9.19E+04 9.16E+04 8.47E+04 6.88E+04 2.69E+04 7.77E+00 5.30E-08 2.99E-20 0.00E+00 0.00E+00 Pm150 1.00E+03 1.00E+03 8.81E+02 7.74E+02 1.27E+02 2.02E+00 1.65E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm151 3.15E+04 3.15E+04 3.13E+04 3.10E+04 2.61E+04 1.77E+04 3.05E+03 7.37E-04 3.96E-19 0.00E+00 0.00E+00 0.00E+00 Sm151 9.85E+01 9.85E+01 9.85E+01 9.86E+01 9.87E+01 9.90E+01 9.95E+01 9.96E+01 9.95E+01 9.93E+01 9.89E+01 9.74E+01 Eu152m 3.63E+01 3.63E+01 3.50E+01 3.37E+01 2.00E+01 6.10E+00 2.88E-02 2.01E-22 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sm153 1.04E+05 1.04E+05 1.03E+05 1.02E+05 9.22E+04 7.26E+04 2.47E+04 2.15E+00 9.25E-10 8.24E-24 0.00E+00 0.00E+00 Gd153 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.04E+03 9.67E+02 8.13E+02 6.28E+02 3.69E+02 4.54E+01 Eu154 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.39E+03 2.36E+03 2.32E+03 2.22E+03 1.89E+03 Eu155 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.00E+03 9.78E+02 9.42E+02 8.74E+02 6.50E+02 Sm156 4.23E+03 4.23E+03 4.07E+03 3.93E+03 2.34E+03 7.20E+02 3.56E+00 3.70E-20 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu156 7.32E+04 7.32E+04 7.31E+04 7.31E+04 7.22E+04 7.00E+04 6.11E+04 1.87E+04 1.21E+03 1.98E+01 4.21E-03 1.39E-17 Gd159 4.36E+04 4.36E+04 4.28E+04 4.21E+04 3.24E+04 1.78E+04 1.21E+03 9.18E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tb160 1.64E+04 1.64E+04 1.64E+04 1.64E+04 1.63E+04 1.62E+04 1.58E+04 1.23E+04 6.91E+03 2.92E+03 4.95E+02 4.50E-01 Tb161 8.08E+03 8.08E+03 8.06E+03 8.05E+03 7.81E+03 7.30E+03 5.41E+03 3.97E+02 9.57E-01 1.13E-04 9.41E-13 0.00E+00 Dy165 3.06E+03 3.06E+03 2.65E+03 2.29E+03 2.86E+02 2.47E+00 1.28E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Dy166 1.45E+01 1.45E+01 1.45E+01 1.44E+01 1.36E+01 1.19E+01 6.43E+00 3.21E-02 1.57E-07 1.69E-15 6.85E-32 0.00E+00 Ho166 7.57E+02 7.57E+02 7.48E+02 7.38E+02 6.18E+02 4.13E+02 7.14E+01 5.01E-02 2.44E-07 2.63E-15 9.13E-32 0.00E+00 Hf175 2.86E+00 2.86E+00 2.86E+00 2.86E+00 2.85E+00 2.83E+00 2.75E+00 2.13E+00 1.17E+00 4.82E-01 7.70E-02 5.57E-05 Lu176m 2.60E+00 2.60E+00 2.36E+00 2.14E+00 5.65E-01 2.67E-02 2.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Lu177 1.12E+00 1.12E+00 1.12E+00 1.11E+00 1.08E+00 1.01E+00 7.40E-01 5.10E-02 6.23E-04 3.53E-04 1.59E-04 6.83E-06 Hf180m 1.19E+01 1.19E+01 1.12E+01 1.05E+01 4.35E+00 5.80E-01 6.64E-05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 7 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay Time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Hf181 4.48E+02 4.48E+02 4.48E+02 4.47E+02 4.45E+02 4.40E+02 4.19E+02 2.74E+02 1.03E+02 2.36E+01 1.14E+00 7.39E-06 W181 1.87E+00 1.87E+00 1.87E+00 1.87E+00 1.87E+00 1.86E+00 1.83E+00 1.58E+00 1.12E+00 6.68E-01 2.32E-01 3.55E-03 Ta182 5.61E+01 5.61E+01 5.61E+01 5.61E+01 5.60E+01 5.58E+01 5.48E+01 4.68E+01 3.26E+01 1.90E+01 6.21E+00 7.60E-02 Ta183 1.54E+02 1.54E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 0.00E+00 W183m 1.69E+02 1.67E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 0.00E+00 W185 5.96E+01 5.96E+01 5.96E+01 5.96E+01 5.94E+01 5.91E+01 5.74E+01 4.52E+01 2.60E+01 1.13E-01 2.05E+00 2.42E-03 Re186 5.05E+01 5.05E+01 5.03E+01 5.01E+01 4.75E+01 4.21E+01 2.42E+01 2.05E-01 3.39E-06 2.27E-13 3.89E-28 0.00E+00 W187 5.17E+02 5.17E+02 5.10E+02 5.02E+02 4.10E+02 2.58E+02 3.19E+01 4.41E-07 3.22E-25 0.00E+00 0.00E+00 0.00E+00 W188 2.72E+00 2.72E+00 2.71E+00 2.71E+00 2.71E+00 2.69E+00 2.61E+00 2.01E+00 1.11E+00 4.50E-01 7.09E-02 4.83E-05 Re188 3.39E+02 3.39E+02 3.36E+02 3.31E+02 2.50E+02 1.31E+02 9.44E+00 2.03E+00 1.12E+00 4.55E-01 7.16E-02 4.88E-05 Os191 6.33E-01 6.33E-01 6.33E-01 6.33E-01 6.29E-01 6.17E-01 5.43E-01 1.69E-01 1.13E-02 1.98E-04 4.74E-08 2.52E-22 Os191m 4.70E-01 4.70E-01 4.58E-01 4.46E-01 3.08E-01 1.32E-01 2.93E-03 1.34E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ir192 3.14E-01 3.14E-01 3.14E-01 3.14E-01 3.13E-01 3.12E-01 3.03E-01 2.37E-01 1.35E-01 5.80E-02 1.02E-02 1.12E-05 Np236m 1.15E+00 1.15E+00 1.14E+00 1.12E+00 9.01E-01 5.51E-01 5.99E-02 2.69E-10 1.46E-29 0.00E+00 0.00E+00 0.00E+00 U237 1.67E+05 1.67E+05 1.66E+05 1.66E+05 1.61E+05 1.50E+05 1.11E+05 7.65E+03 1.69E+01 7.36E-01 7.16E-01 6.50E-01 Pu237 2.52E+00 2.52E+00 2.52E+00 2.52E+00 2.51E+00 2.48E+00 2.37E+00 1.59E+00 6.33E-01 1.59E-01 9.28E-03 1.26E-07 Np238 1.11E+05 1.11E+05 1.11E+05 1.10E+05 9.99E+04 8.03E+04 3.01E+04 6.05E+00 1.35E-02 1.35E-02 1.35E-02 1.33E-02 Pu238 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.38E+03 1.39E+03 1.40E+03 1.41E+03 1.43E+03 1.45E+03 1.44E+03 Np239 3.26E+06 3.26E+06 3.26E+06 3.24E+06 2.98E+06 2.45E+06 1.01E+06 4.94E+02 1.23E+01 1.23E+01 1.23E+01 1.23E+01 Pu239 6.13E+01 6.13E+01 6.13E+01 6.13E+01 6.14E+01 6.16E+01 6.19E+01 6.22E+01 6.22E+01 6.22E+01 6.22E+01 6.22E+01 Np240 7.18E+03 7.18E+03 5.13E+03 3.67E+03 3.33E+01 7.13E-04 4.03E-17 4.55E-17 5.72E-17 7.48E-17 1.11E-16 2.54E-16 Pu240 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.33E+02 Pu241 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.13E+04 3.10E+04 3.07E+04 2.99E+04 2.72E+04 Am241 4.63E+01 4.63E+01 4.63E+01 4.63E+01 4.64E+01 4.65E+01 4.69E+01 5.05E+01 5.86E+01 7.08E+01 9.54E+01 1.86E+02 Am242m 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.01E+00 3.00E+00 3.00E+00 3.00E+00 2.99E+00 2.96E+00 Am242 2.46E+04 2.46E+04 2.40E+04 2.35E+04 1.74E+04 8.70E+03 3.88E+02 2.99E+00 2.99E+00 2.98E+00 2.98E+00 2.95E+00 Cm242 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.71E+04 1.69E+04 1.52E+04 1.17E+04 8.01E+03 3.64E+03 1.65E+02 Pu243 1.22E+05 1.22E+05 1.14E+05 1.06E+05 4.00E+04 4.26E+03 1.80E-01 9.63E-07 9.63E-07 9.63E-07 9.63E-07 9.63E-07

SSES-FSAR Table Rev. 0 FSAR Rev. 64 Page 8 of 8 Table 12.2-9A2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay Time Following Burnup to 58 GWd/MTU Nuclide T = 0 1 sec 30 min 1 hr 8 hr 1.0 d 4.0 d 30.0 d 90.0 d 180.0 d 1 yr 3 yr Am243 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 1.23E+01 Cm243 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.05E+01 1.04E+01 1.03E+01 9.80E+00 Am244 6.13E+04 6.13E+04 5.92E+04 5.72E+04 3.54E+04 1.18E+04 8.44E+01 2.13E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cm244 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.62E+03 2.60E+03 2.58E+03 2.53E+03 2.34E+03

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES

Specific design features for maintaining personnel exposures ALARA are discussed in this subsection.

12.3.1.1 Common Equipment and Component Designs for ALARA

This subsection describes the design features used for several general classes of equipment or components. These classes of equipment are common to many of the plant systems; thus, the features employed for each system to maintain minimum exposures are similar and are discussed by equipment class in the following paragraphs.

Filters: Whenever practicable, filters that accumulate radioactive material are supplied with the means either to backflush the filter remotely or to perform cartridge replacement with semi-remote tools (i.e., long handled tools). For cartridge filters, adequate space is provided to allow removing, cask loading, and transporting the cartridge to the solid radwaste area.

Demineralizers: Demineralizers for radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin tanks prior to dewatering or solidification and that fresh resin can be loaded into the demineralizer remotely. Underdrains and downstream strainers are designed for full system pressure drop. The demineralizers and piping are designed with provisions for being flushed.

Evaporators: Evaporators are provided with chemical addition connections to allow the use of chemicals for descaling operations. Space is provided to allow uncomplicated removal of heating tube bundles. To the extent practicable, the more radioactive components are separated from those that are less radioactive by a shield wall.

Pumps: Wherever practicable, pumps, in radioactive areas are purchased with mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space for access to the pumps for servicing. Small pumps are installed in a manner that allows easy removal if necessary. All pumps in radioactive waste systems are provided with flanged connections for ease in removal. Generally, pump casings are provided with drain connections for draining the pump for maintenance.

Tanks: Whenever practicable, tanks are provided with sloped bottoms and bottom outlet connections. Overflow lines are directed to the waste collection system in order to control contamination within plant structures.

Heat Exchangers: Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to-tube sheet joints welded to minimize leakage. Impact baffles are provided and tube side and shell side velocities are limited to minimize erosive effects.

Instruments: Instrument devices are located in low radiation zones and away from radiation sources when practicable. Where practicable primary instrument devices, which for functional SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-2 reasons are located in high radiation zones, are designed for easy removal to a lower radiation zone for calibration. Where practicable transmitters and readout devices are located in low radiation zones, such as corridors and the control room, for servicing. Some instruments in high radiation zones are provided in duplicate to reduce access and service time required.

Seals are provided on instrument sensing lines on process piping that may contain highly radioactive materials to reduce the servicing time required to keep the lines free of solids.

Instrument and sensing line connections are located in such a way as to avoid corrosion product and radioactive gas buildup.

Valves: To minimize personnel exposures from valve operations, motor-operated, air operated, or other remotely actuated valves are used to the maximum extent practicable.

When practicable, valves are located in valve galleries so that they are shielded separately from the major components that accumulate radioactivity. Long runs of exposed piping are minimized in valve galleries. In areas where manual valves are used on frequently operated process lines, either remote valve operators or shielding is normally provided to minimize personnel exposure.

For equipment located in Zone V areas, remote actuators are provided for frequently operated valves associated with system operation. All other valve operations are either infrequent or performed with equipment in the shutdown mode. To the maximum practicable extent, simple straight reach rods will be used to allow operators to retain the feel of whether the valves are tightly closed or not. Where practicable valves with reach rods are installed with their stems horizontal so that the reach rods are also horizontal but above the heads of personnel to allow ready access.

For valves in radiation areas, provisions are made to drain adjacent radioactive components where practicable.

Wherever practicable, valves for clean, non-radioactive systems are separated from radioactive sources and are located in readily accessible areas.

Manually operated valves in the filter/demineralizer valve compartments required for normal operation and shutdown are equipped with reach rods extending through or over the valve gallery wall. Personnel are not required to enter the valve gallery during flushing operations. The valve gallery shield walls are designed for maximum expected filter backflush activities during flushing operations.

Full ported valves are used in systems expected to contain radioactive solids.

Special valve designs with minimum internal crevices are normally used where crud trapping could become a problem, especially for piping carrying spent resin or evaporator bottoms.

Piping: The piping in pipe chases is designed for the lifetime of the unit. The number of valves or instrumentation in the pipe chases has been reduced to maximum extent practicable. Where radioactive piping is routed through areas that require routine maintenance, pipe chases are normally provided to reduce the radiation contribution from these pipes. Wherever practicable, piping containing radioactive material is routed to minimize radiation exposure to the unit personnel.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-3 Floor Drains: Floor drains and properly sloped floors are provided for each room or cubicle, within which are serviceable components containing radioactive liquids. If a radioactive drain line must pass through a zone lower than that at which it will terminate, proper shielding is provided. Local gas traps or porous seals are not used on radwaste floor drains. Gas traps are provided at the common sump or tank.

Lighting: Multiple electric lights are provided for each cell or room containing highly radioactive components so that the burnout of a single lamp will not require entry and immediate replacement of the defective lamp. Normally, incandescent lights that require less time for servicing are provided to minimize personnel exposure. The fluorescent lights used in some areas do not require frequent service because of the increased life of the tubes.

HVAC: The HVAC system design provides for rapid replacement of the filter elements and housings.

Sample Stations: Sample stations for routine sampling of process fluids are located in accessible areas. Shielding is provided at the sample stations as required to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling. Ventilation, drains or other means of contamination control are provided where necessary. The counting room and laboratory facilities are described in Section 12.5.

Clean Services: Whenever practicable, active components of clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways.

12.3.1.2 Common Facility and Layout Designs for ALARA

This subsection describes the design features used for standard type plant processes and layout situations. These features are used in conjunction with the general equipment designs described in Subsection 12.3.1.1 and include the features discussed in the following paragraphs.

Valve Galleries: Valve galleries are provided with shielded entrances for personnel protection. In many cases the valve galleries are divided by shielding or distance into subcompartments that service only two or three components and are further subdivided by fin walls so that personnel are only exposed to the valves and piping associated with one component at any given location. Threshold berms and floor drains are provided to control radioactive leakage. To facilitate decontamination in valve galleries, concrete surfaces are covered with a smooth surfaced coating which will allow easy decontamination.

Piping: Pipes carrying radioactive materials are routed through controlled access areas zoned for that level of activity. Each piping run is analyzed to determine the potential radioactivity level and surface dose rate. Where radioactive piping must be routed through corridors or other low radiation zone areas, shielded pipeways are normally provided. Whenever practicable, valves and instruments are not placed in radioactive pipeways. Whenever practicable, equipment compartments contain only piping associated with equipment in that compartment.

Where practicable piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be eliminated. Where possible, thermal expansion loops are raised rather than dropped. In radioactive systems, the use of non-removable SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-4 backing rings in the piping joints is minimized to eliminate a potential crud trap for radioactive materials. Piping carrying resin slurries or evaporator bottoms is run vertically as much as possible.

Whenever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe.

Field Run Piping: All routing of radioactive process piping, large and small, is reviewed by the design engineering office.

Penetrations: To minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between the source and the accessible areas. If offsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If these two methods are not used, then baffle shield walls or grouting the area around the penetration are provided.

Contamination Control: Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from certain radioactive systems are piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain. All-welded piping systems are used on contaminated systems to the maximum extent practicable to reduce system leakage and crud buildup at joints. The valves in some radioactive systems are provided with leak-off connections piped directly to the collection system.

Decontamination of potentially contaminated areas within the plant is facilitated by the application of suitable smooth surfaced coatings to the concrete floors and walls.

Floor drains with properly sloping floors are provided in potentially contaminated areas of the plant.

In addition, radioactive and potentially radioactive drains are separated from non-radioactive drains. Systems that become highly radioactive, such as the radwaste slurry transport system, are provided with flush and drain connections. Certain systems have provisions for chemical and mechanical cleaning prior to maintenance.

Equipment Layout: In systems where process equipment is a major radiation source (such as fuel pool cleanup, radwaste, condensate demineralizer, etc.), pumps, valves, and instruments are normally separated from the process component. This allows servicing and maintenance of items in reduced radiation zones. Control panels are located in lowest practicable radiation zones.

Major components (such as tanks, demineralizers, and filters) in radioactive systems are isolated in individual shielded compartments insofar as practicable.

Provision is made on some major plant components for removal of these components to lower radiation zones for maintenance.

Labyrinth entrance way shields or shielding doors are provided for compartments from which radiation could stream to access areas and exceed the radiation zone dose limits for those areas.

For potentially high radiation components (such as filters and demineralizers), completely enclosed shielded compartments with hatch openings are used.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-5 Equipment in non-radioactive systems that requires lubrication is located outside radiation areas.

Wherever practicable, lubrication of equipment in radiation areas is achieved with the use of tube type extensions to reduce exposure during maintenance.

Figures 12.3-1, 12.3-2, 12.3-3, 12.3-4, 12.3-5, and 12.3-6 provide layout arrangements for demineralizers, filters, spent resin storage tanks, hydrogen recombiners, and their associated valve compartments or galleries.

Exposure from routine in-plant inspection is controlled by locating, whenever possible, inspection points in shielded low background radiation areas. Radioactive and non-radioactive systems are separated as far as practicable to limit radiation exposure from routine inspection of non-radioactive systems. For radioactive systems, emphasis is placed on space and ease of motion in a shielded inspection area. Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is normally provided. In high radiation areas where routine surveillance is required, remote viewing devices are provided when practicable.

Facilities for Handling Sealed and Unsealed Radioactive Material:

As discussed in Subsection 12.2.1.9, special material used in the radiochemistry laboratory require the design of special handling equipment. For unsealed materials, the following is provided:

a) Exhaust hoods that exhaust to the ventilation system are located in areas such as sample stations and the radiochemistry laboratory.

b) Decontamination facilities, radiochemistry laboratory, controlled zone shop, instrument repair shops and washdown area are situated at various locations in the plant and are described in Subsection 12.5.2.

c) An area for the repair and maintenance of removed control rod drives is provided in the reactor building in close proximity to the control rod drive removal hatch.

12.3.1.3 Radiation Zoning and Access Control Access to areas inside the plant structures and plant yards is regulated and controlled. Each radiation zone defines the radiation level range to which the aggregate of all contributing sources must be attenuated by shielding.

All plant areas are categorized into radiation zones according to expected radiation levels and anticipated personnel occupancy, with consideration given toward maintaining personnel external exposures ALARA and within the standards of 10CFR20. Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal operation and shutdown; for maintenance occupancy requirements, and for general access requirements to determine appropriate zoning. Radiation zone categories used and their descriptions are given in Table 12.3-1 and the specific zoning for each plant area is shown on Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-6 B1N-100, Sh. 1. Note that the radiation zoning for Unit 1 is not significantly different from those for Unit 2, therefore Dwgs. A 511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1 are also representative of Unit

2. Attachment of the new common tool room facility to Unit 2 turbine building (Ref. Dwg. M-231, Sh. 1) represents a unique feature not represented by the radiation zoning for Unit 1 turbine building at elevation 676' (Dwg. A-514, Sh. 1). The radiation zoning for the tool room is Zone II. Where possible, frequently accessed areas, i.e., corridors, are shielded for Zone I and Zone II access.

The control of ingress or egress of plant operating personnel to controlled access areas and procedures employed to ensure that radiation levels and allowable working time are within the limits prescribed by 10CFR20 as described in Section 12.5.

12.3.1.4 Control of Activated Corrosion Products

In order to minimize the radiation exposure associated with the deposition of activated corrosion products in reactor coolant and auxiliary systems, the following steps have been taken:

(1) The reactor coolant system consists mainly of austenitic stainless steel, carbon steel and low alloy steel components. Nickel content of these materials is low, and it is controlled in accordance with applicable ASME material specifications.

A small amount of nickel base material (Inconel 600) is employed in the reactor vessel internal components. Inconel 600 is required where components are attached to the reactor vessel shell and the coefficient of expansion must match the thermal expansion characteristics of the low alloy vessel steel. Inconel 600 was selected because it provides the proper thermal expansion characteristics, adequate corrosion resistance and can be readily fabricated and welded.

(2) Materials employed in the reactor coolant system are purchased to ASME material specification requirements. No special controls on levels of cobalt impurities are specified.

(3) Hardfacing and wear materials having a high percentage of cobalt are restricted to applications where no satisfactory alternate materials are available. The EPRI cobalt reduction guidelines (Ref. 12.3-24) are utilized to the ex tent practical.

(4) A high temperature filtration system was not employed in the Reactor Water Clean-up System. The reasons for this included:

a) Lack of quantitative data on the removal efficiency for insoluble cobalt by the high temperature filter; b) Uncertainty in the deposition model including the relative effectiveness of cobalt removal on deposition rate;

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-7 c) Doubtful cost-effectiveness in an area where other methods under study (such as decontamination) may prove better at reducing dose rates while also being more cost-effective.

(5) Items 1, 2, and 3 above also apply to valve materials in contact with reactor coolant. Valve packing materials are selected primarily for their properties in the particular environment.

(6) Subsections 12.1.2.2, 12.3.1.1, and 12.3.1.2 describe the various flushing, draining, testing, and chemical addition connections which have been incorporated into the design of piping and equipment which handle radioactive materials. If decontamination is to be performed, these connections would be used for that purpose.

(7) The plant is designed with a powdered resin, pressure precoat clean-up system for the primary coolant in the reactor and a full flow deep bed condensate demineralizer and filter vessel system for the feedwater. See Dwgs. M-116, Sh. 1, M-116, Sh. 2, M-116, Sh. 3, M-144, Sh. 1, M-144, Sh. 2, M1-G33-16, Sh. 1, M1-G33-18, Sh. 1 and M-145, Sh. 1.

(8) A chemistry control program has been developed and implemented at SSES to reduce crud buildup.

12.3.2 SHIELDING

In this subsection the bases for the nuclear radiation shielding and the shielding configurations are

discussed.

12.3.2.1 Design Objectives

The basic objective of the plant radiation shielding is to reduce personnel exposures, in conjunction with a program of controlled personnel access to and occupancy of radiation areas, to levels that are within the dose regulations of 10CFR50 and are as low as reasonably achievable (ALARA) within the dose regulations of 10CFR20. Shielding and equipment layout and design are considered in ensuring that exposures are kept ALARA during anticipated personnel activities in areas of the plant containing radioactive materials.

Basic plant conditions considered in the nuclear radiation shielding design are normal operation at full-power, and plant shutdown.

The shielding design objectives for the plant during normal operation, including anticipated operational occurrences, and for Shutdown operations are:

a. To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10CFR20 b. To ensure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-related operations required for each plant equipment and instrumentation area

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-8 c. To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.

d. To sufficiently shield the control room so that the total dose from all post-accident sources (Rem-TEDE).(calculated in Chapter 15) in the event of design basis accidents will not exceed the limits of 10CFR50.67.

12.3.2.2 General Shielding Design Shielding is provided to attenuate direct radiation through walls and penetrations and scattered radiation to less than the upper limit of the radiation zone for each area shown in Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1. Since the layout for Unit 2 is not significantly different from that of Unit 1, the minimum shielding requirements (see Subsection 12.3.2.3) indicated on those drawings are applicable to both Units. General locations of the plant areas and equipment discussed in this subsection are also shown on those drawings.

The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 145 lb/ft

3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design ensures protection on an equivalent shielding basis as determined by the characteristics of the concrete block selected. Compliance of concrete radiation shield design with Regulatory Guide 1.69 is discussed in Section 3.13. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.

Special features employed to maintain radiation exposures ALARA in routinely occupied areas such as valve operating stations and sample stations are described in Subsections 12.3.1.1 and 12.3.1.2.

12.3.2.2.1 Reactor Building Shielding Design

During reactor operation, the steel-lined, reinforced concrete drywell wall and the reactor building walls protect personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and associated equipment within the reactor building. The reactor vessel shield wall, drywell wall, and various equipment compartment walls together with the reactor building walls minimize the radiation levels outside the reactor building.

Where personnel and equipment hatches or penetrations pass through the drywell wall, additional shielding is designed to attenuate the radiation to below the required level defined by the radiation zone outside the drywell wall during normal operation and shutdown and to acceptable emergency levels as defined by 10CFR50 during design basis accidents.

12.3.2.2.2 Reactor Building Interior Shielding Design Inside Drywell Structure: Areas within the drywell are designed as Zone V areas and are normally inaccessible during plant operation. The reactor vessel shield provides shielding for access in the SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-9 drywell during shutdown, and reduces the activation of and radiation damage to drywell equipment and materials.

Outside Drywell Structure: The drywell wall is designed to reduce radiation levels in normally occupied areas of the reactor building from sources within the drywell to less than the maximum level for Zone II.

Penetrations and hatch openings in the drywell wall are shielded, as necessary, to meet adjacent area radiation zoning levels. Shielding requirements for the personnel, equipment, and CRD removal hatch openings are shown on Dwg. A-522, Sh. 1 in the areas numbered 412, 413, and 402, respectively. Drywell piping and electrical penetrations are shielded by providing either local shields within the penetration assembly or a shielded penetration room. Shielded piping penetration room locations and bulk shielding requirements are shown on Dwgs. A-521, Sh. 1, A-522, Sh. 1, and A-523, Sh. 1. These rooms, numbered 202, 204, 205; 403, 411, 501, 504, 506, 515; are designated radiation Zone V during reactor power operation and are provided with personnel access controls. Electrical penetrations which are not located within these rooms are provided with supplementary local shielding as needed to meet outside zoning levels.

The components of the reactor water cleanup (RWCU) system described in Section 5.4.8 are located in shielded compartments which are designed as Zone V, restricted access areas. Shielding is provided for each piece of equipment in the RWCU system consistent with its postulated maximum activity Subsection 12.2.1 and with the access and zoning requirements of the adjacent areas. This equipment includes:

a) Regenerative heat exchanger b) Non-regenerative heat exchanger

c) RWCU pumps and piping

d) RWCU filter demineralizers and holdup pumps

e) RWCU backwash receiving tank and piping.

The traversing in-core probe (TIP) system is located inside a shielded compartment to protect personnel from the neutron activated portion of the TIP cable.

Main steamlines are located within shielded structures from the drywell wall to the reactor building

wall.

Spent fuel is a primary source of radiation during refueling. Because of the extremely high activity of the fission products contained in the spent fuel assemblies and the proximity of Zone II areas, shielding is provided for areas surrounding the fuel transfer canal and pool to ensure that radiation levels remain below zone levels specified for adjacent areas.

After reactor shutdown, the Residual Heat Removal (RHR) System pumps and heat exchangers are in operation to remove heat from the reactor water. It is anticipated that the radiation levels in the vicinity of this equipment will temporarily reach Zone V levels due to corrosion and fission products in the reactor water. Shielding is designed to attenuate radiation from RHR equipment SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-10 during shutdown cooling operations to levels consistent with the radiation zoning requirements of adjacent areas.

During functional testing operations of the Reactor Core Isolation Cooling (RCIC) System and the High Pressure Coolant Injection (HPCI), the steam driven turbine and the inlet and exhaust piping are shielded consistent with the maximum steam activities in the lines and the access zone requirements of surrounding areas.

The concrete shield walls surrounding the spent fuel cask loading, storage, and transfer areas, as well as the shield walls surrounding the fuel transfer and storage areas, are designed to provide Zone II maximum dose rates in accessible areas outside of the shield walls.

Water in the spent fuel pool provides shielding above the spent fuel transfer and storage areas.

Direct radiation levels at the fuel handling equipment are calculated to be less than 2.5 mrem/hr from spent fuel during normal operations.

Water is also used as shielding material above the steam dryer and separator storage area.

Concrete walls and water in the pool are designed to provide Zone II dose rates in adjacent accessible areas during storage of the dryer and separator.

The Fuel Pool Cooling and Cleanup (FPCC) System (see Section 9.1.3) shielding is based on the maximum activity discussed in Subsection 12.2.1 and the access and zoning requirements of adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, pumps and piping, filter demineralizers, and backwash receiving tank.

12.3.2.2.3 Radwaste Building Shielding Design

Shielding is provided as necessary around the following equipment in the radwaste building to ensure that the radiation zone and access requirements are met for surrounding areas.

a) Laundry drain tank and pumps b) Chemical waste tank and pumps c) Radwaste evaporators d) Radwaste evaporator tanks and pumps e) Liquid radwaste collection tanks and pumps f) Liquid radwaste surge tanks g) Liquid radwaste sample tanks and pumps h) Reactor water cleanup phase separator and pumps i) Waste sludge phase separator and pumps j) Spent resin tank k) Waste filling and capping station l) Waste liner transfer and storage areas m) Liquid radwaste demineralizer and piping SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-11 n) Waste mixing tanks o) Liquid radwaste filters p) Gaseous radwaste equipment.

12.3.2.2.4 Turbine Building Shielding Design

Radiation shielding is provided around the following equipment in the turbine building to ensure that zone access requirements (Dwgs. A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, and A-518, Sh. 1) are met for the following surrounding areas:

a) Condensate filter demineralizers and piping b) Regeneration waste surge tanks and pumps c) Chemical waste neutralizing tanks and pumps d) Reactor feed pump turbines and piping e) Condensate pumps and piping f) Main condensers and hotwell g) Mechanical vacuum pump h) Recombiners and piping i) Steam packing exhauster j) Condensate demineralizer resin regeneration tanks k) Air ejectors and gland steam condensers l) Feedwater heaters, heater drains, and piping m) Main steam piping n) Steam seal evaporator and drain tank o) Moisture separator and drain tanks p) High pressure and low pressure turbines q) Offgas piping r) Ultrasonic resin cleanser s) Condensate filter vessels t) Condensate filter vessel backwash receiving tank

Areas within most of these shield walls have high radiation levels and limited access.

12.3.2.2.5 Control Room Shielding Design

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-12 "Radiation shielding is provided, as necessary, for the control building and the control structure envelope in order to ensure that the radiation zoning and access requirements as presented on Dwgs. A-511, Sh. 1 and A-512, Sh. 1 are satisfied during normal operation. In addition, shielding is provided to permit access and occupancy of the areas of the control structure envelope in which critical safety functions are performed under post accident conditions with radiation doses limited to 5 rem TEDE from all contributing modes of exposure for the duration of any accident described in Chapter 15 (in accordance with 10CFR50.67).

An isometric drawing of the control and reactor building shielding is provided on Figure 12.3-29 to show the relationship of potential post accident sources to control structure habitability zones. The parameters used in the assessment of control structure envelope habitability during normal and abnormal station operating conditions, including post accident requirements as discussed in Sections 6.4.9.4, and in Appendix 15B."

12.3.2.2.6 Diesel Generator Building Shielding Design

There are no radiation sources in the diesel generator building; therefore, no shielding is required for the building.

12.3.2.2.7 Miscellaneous Plant Areas and Plant Yard Areas

Sufficient shielding and/or radiation protection controls are provided for all plant buildings and designated yard areas containing radiation sources or radioactive material such that radiation levels are maintained within regulatory Units. Some operations, such as loading solidified waste into shield casks, and storage of radiation materials, require access controls. These areas are surrounded by a security fence and closed off from areas accessible to the general public.

12.3.2.2.8 Counting Room Shielding Because the counting room contains sensitive instruments to radioactivity measurements, it is imperative that the background radiation levels are minimized. To accomplish this, no fly ash was used in the concrete mix for the walls and slabs surrounding the counting room. Fly ash normally contains a relatively large amount of slow decaying radioactive isotopes. In addition, the shield walls and slabs were sized to maintain a background radiation level of less than 130 mrem/year for anticipated operational occurrences and 45 mrem/year for normal operation.

12.3.2.3 Shielding Calculational Methods

The shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in Subsection 12.2.1. The thickness of each shield wall surrounding radioactive equipment is determined by approximating as closely as possible the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to energy group sources using data from standard Refs. 12.3-1, 12.3-2, 12.3-3, 12.3-4, and 12.3-5.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-13 The geometric model assumed for shielding evaluation of tanks, heat exchangers, filters, demineralizers, and evaporators is a finite cylindrical volume source. For shielding evaluation of piping, the geometric model is a finite shielded cylinder. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source. Typical computer codes that are used for shielding analysis are listed in Table 12.3-2. Shielding attenuation data used in those codes include gamma class attenuation coefficients (Ref. 12.3-6),

gamma buildup factors (Ref. 12.3-7), neutron-gamma multigroup cross sections (Ref. 12.3-20), and albedos (Ref. 12.3-12).

The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area.

Shielding requirements are evaluated at the point of maximum radiation dose through any wall.

Therefore, the actual anticipated radiation levels in the greater region of each plant area is less than this maximum dose and therefore less than the radiation zone upper limit.

Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are designed using a general purpose gamma-ray scattering code G 3 (Ref. 12.3-11). The mazes are constructed so that the scattered dose rate plus the transmitted dose rate through the shield wall from all contributing sources is below the upper limit of the radiation zone specified for each plant area.

2.3.3 VENTILATION

The plant heating, ventilating, and air conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Detailed HVAC system descriptions are provided in Section 9.4. Control Structure habitability is discussed in Section 6.4.

12.3.3.1 Design Objectives The systems are designed to operate such that the in-plant airborne activity levels for normal operation (including anticipated operational occurrences) in the general personnel access areas are within the limits of 10CFR20. The systems operate to reduce the spread of airborne radioactivity during normal and anticipated abnormal operating conditions.

During post-accident conditions, the ventilation system for the plant control room provides a suitable environment for personnel and equipment and ensures continuous occupancy in this area.

The plant ventilation systems are designed to comply with the airborne radioactivity release limits for offsite areas during normal operation.

12.3.3.2 Design Criteria

Design criteria for the plant HVAC systems include the following:

a) During normal operation and anticipated operational occurrences, airborne radioactivity levels to which plant personnel are exposed is ALARA and within the limits specified in 10CFR20.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-14 b) During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10CFR20 and 10CFR50.

c) The plant siting dose guidelines of 10CFR50.67 will be satisfied following those hypothetical accidents, described in Chapter 15, which involve a release of radioactivity from the plant.

d) The dose to control room personnel shall not exceed the limits specified in 10CFR50.67 following those hypothetical accidents, described in Chapter 15, which involve a release of radioactivity from the plant.

12.3.3.3 Design Guidelines In order to accomplish the design objectives, the following guidelines are followed wherever practicable.

12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity

a) Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.

b) Equipment vents and drains are piped directly to a collection device connected to the collection system instead of allowing any contaminated fluid to flow across the floor to the floor drain.

c) All-welded piping systems are used on contaminated systems to the maximum extent practicable to reduce system leakage. If welded piping systems are not used, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.

d) The valves in some systems are provided with leak-off connections piped directly to the collection system.

e) Suitable coatings are applied to the concrete floors of potentially contaminated areas to facilitate decontamination.

f) Where practicable, metal diaphragm or bellows seat valves are used on those systems where essentially no leakage can be tolerated.

g) Contaminated equipment has design features that minimize the potential for airborne contamination during maintenance operations. These features may include flush connections on pump casings for draining and flushing the pump prior to maintenance or flush connections on piping systems that could become highly radioactive.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-15 h) Exhaust hoods are used in plant areas to facilitate processing of radioactive samples by drawing contaminants away from the personnel breathing areas and into the ventilation and filtering systems.

i) Equipment decontamination facilities are ventilated to ensure control of released contamination and minimize personnel exposure and the spread of contamination.

12.3.3.3.2 Guidelines to Control Airborne Radioactivity a) The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination under normal conditions.

b) In building compartments with a potential for contamination, a greater volumetric flow is exhausted from the area than is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.

c) Floor and equipment drain collector tank vents are piped to a collection header and processed by the tank vent filter system.

d) Exhaust air is routed through a prefilter and HEPA filters or a combination of prefilter, HEPA and charcoal filters where necessary before release to the atmosphere to reduce onsite and offsite airborne concentrations.

e) Air is supplied to each principal building via separate supply intakes and duct systems.

f) Redundant, Seismic Category I systems and components are provided for portions of the ventilation system required for safe shutdown of the reactor and to mitigate the consequences of design basis accidents. Included herein are the plant control room ventilation system, the reactor building recirculation system, the standby gas treatment system, and coolers and selected engineered safety feature equipment rooms.

g) Air being discharged from potentially contaminated areas of the Turbine Building and the Reactor Building is passed through prefilters, HEPA filters and charcoal adsorbing filters.

Air being discharged from the Radwaste Building is passed through prefilters and HEPA filters. Means are provided to isolate these areas upon indication of contamination to prevent the discharge of contaminants to the environment.

h) Suitable containment isolation valves are installed in accordance with General Design Criteria 54 and 56, including valve controls, to ensure that the containment integrity is maintained. See additional discussion in Subsections 3.1.2.5.5, 3.1.2.5.7 and 6.2.4.

12.3.3.3.3 Guidelines to Minimize Personnel Exposure from HVAC Equipment

a) Ventilation fans and filters are provided with adequate access space to permit servicing with minimum personnel radiation exposure. The HVAC system is designed to allow rapid replacement of components.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-16 b) Ventilation ducts are designed to minimize the buildup of radioactive contamination within the ducts to the maximum extent practicable. Welded seams are used to join ductwork segments and internal obstructions are avoided wherever practicable.

c) Access and service of ventilation systems in potentially radioactive areas are provided by component location to minimize operator exposure during maintenance, inspection, and testing as follows:

1) The outside air supply units and building exhaust system components are enclosed in ventilation equipment rooms. These equipment rooms are located in radiation Zone II and are accessible to the operators. Work space will be provided around each unit for anticipated maintenance, testing, and inspection. Filter-adsorber units generally comply with the access and service requirements of Regulatory Guide 1.52. (Refer to response to Regulatory Guide 1.52 in Section 3.13.)

Local cooling equipment, servicing the building requirements, will normally be located in areas of low contamination potential radiation Zones I or II.

d) While the majority of the activity in the filter train is removed by simply removing the contaminated filters, further decontamination of the internal structure is facilitated by the proximity of electrical outlets for operation of decontamination equipment, and water supply for washdown of the interior, if necessary. Drains are provided on the filter housing for removal of contaminated water.

e) Active elements of the atmospheric cleanup systems are designed to permit easy removal.

f) Access to active elements is direct from working platforms to simplify element handling. Space is provided on the platforms for accommodating safe personnel movement during replacement of components, including the use of necessary material-handling facilities and during any in-place testing operations.

g) The clear space for doors is a minimum of 3 ft by 7 ft.

h) The filters are designed with replaceable units that are clamped in place against compression seals. The filter housing is designed, tested, and proven to be airtight with bulkhead type doors that are closed against compression seals.

i) Filter systems in which radioactive materials could accumulate to produce significant radiation fields external to the ductwork are appropriately located and shielded to reduce exposure to personnel and equipment.

j) Filters in all systems are changed based upon the airflow and the pressure drop across the filter bank. Charcoal adsorbers are changed based on the residual adsorption capacity of the bed as measured by the testing of carbon samples taken from the removable canisters located in the carbon bed. The testing of the carbon adsorbers and all other components is described in Table 9.4-1.

12.3.3.4 Design Description

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-17 The ventilation systems serving the following structures are assumed to be potentially radioactive and are discussed in detail in Section 9.4.

a) Reactor Building b) Radwaste Building c) Turbine Building Although the control room is considered to be a non-radioactive area, radiation protection is provided to ensure habitability (see Section 6.4).

Ventilation system design parameters for the four systems are given in Tables 9.4-2, 9.4-3, 9.4-6, and 9.4-7.

A typical layout of a potentially radioactive filter unit is given on Figure 12.3-3.

12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION

12.3.4.1 Area Radiation Monitoring

The area radiation monitoring system supplements the personnel and area radiation survey provisions of the plant health physics program described in Section 12.5 to ensure compliance with the personnel radiation protection guidelines of 10CFR20, 10CFR50, 10CFR70, and Regulatory Guides 1.21, 8.2, 8.8 and 8.12 as discussed below.

The area radiation monitors function to:

a) Alert plant personnel of abnormally high radiation levels which, if unnoticed, could result in inadvertent exposures.

b) Inform the control room operator of the occurrence and approximate location of abnormal radiation level increase.

c) Comply with the requirements of 10CFR50 Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.

d) Assist in the detection of unauthorized or inadvertent movement of radioactive material in the plant.

e) Supplement other systems including process radiation monitoring, leak detection, etc., in detecting abnormal migrations of radioactive material in or from the process streams.

The area radiation monitoring system has no function related to the safe shutdown of the plant, or to the quantitative monitoring of the release of radioactive material to the environment.

The combination of the airborne radioactivity monitoring system in conjunction with administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns throughout the plant, plant equipment layout, lack of significant radioactive airborne SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-18 sources in normally occupied areas (radiation Zones I and II), and administrative control of access into applicable radiation and high radiation areas is sufficient to ensure that airborne radioactivity levels are safe in terms of the required duration of personnel access. A general review of these concepts follows:

a) Significant releases of airborne radioactive materials within the plant are detected by effluent monitors as described in Table 11.5-1 in Section 11.5. Air flow patterns are normally from occupied areas to non-occupied areas and from low airborne radioactive material areas to high airborne radioactive material areas. Readouts and alarms are located locally and in the control room.

b) Additional Continuous Air Monitors (CAMS) with local readout and alarm are located in selected areas of potential airborne concentrations throughout the reactor turbine and radwaste building (s).

c) Administrative controls for limiting exposure to airborne radioactivity concentrations greater than 10 Derived Air Concentration (DAC)-HRS. as specified in 10CFR20 Appendix B to Sections 20.1001 - 20.2401, Table 1 Col. 3 are as follows:

1) Routine airborne radioactivity surveys of various accessible radiation zones within the plant. The routine monitoring schedule and frequency is delineated in Station Health Physics Procedures. These locations may be modified with consideration of plant operating status.

2) Access to airborne radioactivity areas with concentrations greater than 30% of the applicable Derived Air Concentration and/or Derived Air Concentration mixture are controlled via a Radiation Work Permit (RWP). Entry into and/or work in the area are preceded by a survey sufficient to determine the radiological conditions present and protection required for these conditions. This information is specified on the RWP. 3) Access to areas where the potential for high radioactive airborne concentration exist due to work conditions is controlled via the RWP process.

12.3.4.1.1 Criteria for Area Monitor Selection The following design criteria are applicable to the area radiation monitoring system.

Energy Dependence: The detector-indicator and trip unit should be responsive to gamma radiation over an energy range of 80 KeV to 6 MeV. The energy dependence should not exceed

+/-20 percent of point for an exposure rate of approximately 50 mr/hr from 100 KeV to 3 MeV and there should be response from 80 KeV to 6 MeV.

Accuracy: The overall accuracy within the design range of temperature, humidity, line voltage, and line frequency variation should be such that the actual reading relative to the true reading, including susceptibility and energy dependence (100 KeV to 3 MeV), should be within 9.5 percent of equivalent linear full scale recorder output for any decade.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-19 Reproducibility: At design center the reading shall be reproducible within

+/-10 percent of point with constant geometry.

ENVIRONMENTAL CONDITIONS PARAMETER SENSOR LOCATION CONTROL ROOM DESIGN CENTER RANGE DESIGN CENTER RANGE Temperature (degrees C) 25 0 to 60 25 5 to 50 Relative Humidity (Percent) 50 20 to 100 50 20 to 90 12.3.4.1.2 Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel unknowingly to receive high radiation doses (e.g., in excess of 10CFR20 limits) in a short period of time because of system failure or improper personnel action. Any plant area that meets one or more of the following criteria is monitored:

a) Zone I areas which, during normal plant operation, including refueling, could exceed the radiation limit of 0.5 mrem/hr upon system failure or personnel error or which will be continuously occupied following an accident requiring plant shutdown b) Zone II areas where personnel could otherwise unknowingly receive high levels of radiation exposure due to system failure or personnel error c) Area monitors are in accordance with General Design Criterion 63 of 10CFR50 Appendix A. 12.3.4.1.3 System Description (Area Radiation Monitoring)

General The area radiation monitoring system is shown in diagram form in Dwgs. M-137, Sh. 1 and M-137, Sh. 2. Each channel consists of a combined sensor/converter unit, a local auxiliary unit (readout with visual and audible alarm), a combined indicator/trip unit, a shared power supply, and a shared multipoint recorder. The exception to this is that the accident range monitors, channels 48 through 57, do not have audible alarms. The location of each area radiation detector is indicated on the shielding and zoning drawings, Dwgs. A-511, Sh. 1, A-512, Sh. 1, A-513, Sh. 1, A-514, Sh. 1, A-515, Sh. 1, A-516, Sh. 1, A-517, Sh. 1, A-518, Sh. 1, A-519, Sh. 1, A-520, Sh. 1, A-521, Sh. 1, A-522, Sh. 1, A-523, Sh. 1, A-524, Sh. 1, A-525, Sh. 1, A-526, Sh. 1, A-527, Sh. 1, A-528, Sh. 1, A-529, Sh. 1, A-530, Sh. 1, and B1N-100, Sh. 1, and with the exception of the LLRW Holding Facility, is listed in Table 12.3-7. With the exception of channels 11 and 12, the detector locations are the same for both Units 1 and 2. In Unit 1 the channel 11 monitor is located adjacent to the Reactor Building sample station which is located just outside room I-508 and in Unit 2 the channel 11 monitor is located adjacent to the Reactor Building sample station which is inside SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-20 room II-508. In Unit 1, channel 12 monitors the spent fuel hoistway during transfer operations.

This hoistway does not exist in Unit 2.

Circuit Description Sensor/Converter: Each sensor/converter contains all silicon semiconductors in sealed enclosure with a Cooke-Yarborough courtyard circuit which combines a long integrating time constant at low radiation levels with fast overall response at high radiation levels.

Auxiliary Unit: Each auxiliary unit gives instant local readout at the sensor location with a visual alarm. An audible alarm is connected to the auxiliary unit to alert personnel of excessive area radiation.

Indicator and Trip Unit: The indicator and trip unit provides channel control for the area radiation monitoring system. Its circuitry provides an upscale trip that indicates high radiation and a downscale trip that may indicate instrument trouble or loss of power. The module has an analog readout, a low and high trip indicating light, a trip test device, an alarm reset and an output for a multipoint recorder.

Ranges and Sensitivity: Ranges and sensitivities are selected for each location based on the anticipated radiation level as provided by experimental measurements of levels in similar plants and shielding calculations. Refer to Table 12.3-7 for detail. Additional range (10 2 - 10 6 mR) was added for Licensing Commitment to Regulatory Guide 1.97, Rev. 2.

Accuracy: The overall accuracy is such that the actual reading relative to the true reading is within

+/-7.5 percent of equivalent full scale.

12.3.4.1.4 Area Radiation Monitoring Instrumentation

Power Sources: The power source for the area radiation monitoring system is the 120V AC instrument bus and local lighting panels. The area radiation monitor instrumentation is powered by a high and low voltage electrically regulated power supply capable of handling up to 10 channels.

The system has no emergency power supply.

Alarm Setpoints: Alarm setpoints may vary depending on operational considerations and will be determined by measured radiation levels in accordance with controlled station procedures.

Recording Devices: Two multipoint recorders are located in the control room for recording channels pertaining to Unit 1, Unit 2, and channels which are common to both units. This data is also stored in computer history files and can be retrieved and printed using the PMS Historical Recording service program.

Location of Devices: Refer to details in Table 12.3-7.

Readouts and Alarms: Readouts, visual and audible alarms are provided locally for each monitoring channel. The accident range monitors, Channels 48 through 57, serve as indicating channels only and do not have audible alarms. The normal range channels in the same locations serve as the alarm monitors. Readouts and visual alarms are provided by each indicator/trip unit in the Control Structure (Upper Relay Room). Multipoint recorders, visual alarms and PMS displays SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-21 are provided in the Control Structure (Control Room), with the exception of the three Technical Support Center channels (43, 44, 45). The following annunciators are located in the main control room to alert the operator:

a) Reactor Building Area High Radiation (Units 1 and 2)

b) Turbine Building Area High Radiation (Units 1 and 2)

c) Radwaste Building Area High Radiation

d) Refueling Floor Area High Radiation (Units 1 and 2)

e) Spent Fuel Pool Area High Radiation (Units 1 and 2)

f) Reactor Building Common Area High Radiation

g) Administration Building Area High Radiation

h) Control Structure Area High Radiation i) Area Radiation Monitoring Downscale (ganged for all channels)

12.3.4.1.5 Safety Evaluation The area radiation monitoring system is designed to operate unattended for extended periods and is designed for high reliability. Failure of one monitor has no effect on any other.

The system is not essential for safe shutdown of the plant, and serves no active emergency function during operation. The system is not safety related and is constructed to Quality Group D Requirements.

12.3.4.1.6 Calibration Method and Testability Facilities for calibrating these monitor units are provided, which include a test unit designed for use in the adjustment procedure for the area radiation monitor sensor and converter unit. These provide several gamma radiation levels between 1 and 250 mrem/hr.

A cavity in the calibration unit receives the sensor and converter unit. A window through which radiation from the source emanates is located on the back wall of the cylindrical lower half of the cavity. A chart on each calibration unit indicates the radiation levels available from the unit for the various control settings.

An internal trip test circuit, adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip unit input so that a meter reading is provided in addition to a real trip. All trip circuits are the latching type and must be manually reset in the Upper Relay Room.

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-22 The radiation monitors will be calibrated at regular time intervals in accordance with station procedures.

12.3.4.2 Airborne Radioactivity Monitoring

Refer to Subsections 12.5.2.2.3 and 12.5.3.5.4 for information on air borne radioactivity monitoring.

12.

3.5 REFERENCES

12.3-1 J. J. Martin and P. H. Blichert-Toft, Nuclear Data Tables "Radioactive Atoms, Auger Electrons, and X-Ray Data," Academic Press, October, 1970.

12.3-2 J. J. Martin, Radioactive Atoms Supplement 1, ORNL 4923, August, 1973.

12.3-3 W. W. Bowman and K. W. MacMurdo, Atomic Data and Nuclear Data Tables, "Radioactive Decay's Ordered by Energy and Nuclide," Academic Press, February, 1970.

12.3-4 M. E. Meek and R. S. Gilbert, "Summary of X-Ray and Gamma Energy and Intensity Data," NEDO-12037, January, 1970.

12.3-5 C. M. Lederer, et al, Table of Isotopes, Lawrence Radiation Laboratory, University of California, March, 1968.

12.3-6 G. W. Goldstein, X-Ray Attenuation Coefficients from 10 KeV to 100 MeV, National Bureau of Standards Circular 583 (Issued April 30, 1957).

12.3-7 D. K. Trubey, "A Survey of Empirical Functions Used to Fit Gamma-Ray Buildup Factors," ORNL-RSIC-10, February, 1966.

12.3-8 W. W. Engle, Jr., "A User's Manual for ANISN: A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," Union Carbide Corporation, Report No. K-1693, 1967.

12.3-9 R. E. Malenfant, QAD, A Series of Point-Kernel General-Purpose Shielding Programs, Los Alamos Scientific Laboratory, LA 3573, October, 1966.

12.3-10 D. Arnold and B. F. Maskewitz, "SDC, A Shielding-Design Calculation for Fuel-Handling Facilities," ORNL-3041, March, 1966.

12.3-11 R. E. Malenfant, "G 3: A General Purpose Gamma-Ray Scattering Program," Los Alamos Scientific Laboratory, LA 5176, June, 1973.

12.3-12 W. E. Selph, "Neutron and Gamma Ray Albedos," ORNL-RSIC-21, February, 1968.

12.3-13 D. S. Duncan and A. B. Spear, Grace I - An IBM 704-709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields, Atomics International (June, 1959).

SSES-FSAR Text Rev. 59 FSAR Rev. 64 12.3-23 12.3-14 D. S. Duncan and A. B. Spear, Grace II - An IBM 709 Program for Computing Gamma Ray Attenuation and Heating in Cylindrical and Spherical Geometries, Atomics International (November, 1959).

12.3-15 D. A. Klopp, NAP - Multigroup Time-Dependent Neutron Activation Prediction Code , IITRI-A6088-21 (January, 1966).

12.3-16 E. A. Straker, P. N. Stevens, D. C. Irving, and V. R. Cain, MORSE - A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code, ORNL-4585 (September, 1970).

12.3-17 W. A. Rhoades and F. R. Mynatt, The DOT III Two-Dimensional Discrete Ordinates Transport Code, ORNL-TM-4280 (1973).

12.3-18 U.S. Nuclear Regulatory Commission, Regulatory Guide 8.8 (July, 1973).

12.3-19 M. J. Bell, "ORIGEN - The ORNL Generation and Depletion Code," Oak Ridge National Laboratory, ORNL-4628 (May, 1973).

12.3-20 ORNL RSIC Computer Code Collection DLC-23, CASK -

40 Group Neutron and Gamma-Ray Cross Section Data.

12.3-21 R. G. Jaeger, et al, Engineering Compendium on Radiation Shielding, Volume I, Springer - Verlag, New York Inc., 1968.

12.3-22 C. A. Negin and G. Worku, Microshield , Grove Engineering, Inc., Rockville, Md.

12.3-23 MICRO Skyshine, Framatome Technologies, Inc., d.b.a. Grove Engineering, Rockville, MD.

12.3-24 Cobalt Reduction Guidelines, Electric Power Research Institute, Palo Alto, CA, March 1990, NP-6737.

THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 12.3-7, Rev. 48 AutoCAD Figure 12_3_7doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-511, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-8 replaced by dwg.

A-511, Sh. 1 FIGURE 12.3-8, Rev. 55 AutoCAD Figure 12_3_8.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-512, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-9 replaced by dwg.

A-512, Sh. 1 FIGURE 12.3-9, Rev. 49 AutoCAD Figure 12_3_9.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-513, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-10 replaced by dwg.

A-513, Sh. 1 FIGURE 12.3-10, Rev. 57 AutoCAD Figure 12_3_10.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-514, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-11 replaced by dwg.

A-514, Sh. 1 FIGURE 12.3-11, Rev. 56 AutoCAD Figure 12_3_11.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-515, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-12 replaced by dwg.

A-515, Sh. 1 FIGURE 12.3-12, Rev. 57 AutoCAD Figure 12_3_12.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-516, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-13 replaced by dwg.

A-516, Sh. 1 FIGURE 12.3-13, Rev. 56 AutoCAD Figure 12_3_13.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-517, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-14 replaced by dwg.

A-517, Sh. 1 FIGURE 12.3-14, Rev. 57 AutoCAD Figure 12_3_14.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-518, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-15 replaced by dwg.

A-518, Sh. 1 FIGURE 12.3-15, Rev. 57 AutoCAD Figure 12_3_15.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-519, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-16 replaced by dwg.

A-519, Sh. 1 FIGURE 12.3-16, Rev. 56 AutoCAD Figure 12_3_16.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-520, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-17 replaced by dwg.

A-520, Sh. 1 FIGURE 12.3-17, Rev. 56 AutoCAD Figure 12_3_17.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-521, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-18 replaced by dwg.

A-521, Sh. 1 FIGURE 12.3-18, Rev. 56 AutoCAD Figure 12_3_18.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-522, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-19 replaced by dwg.

A-522, Sh. 1 FIGURE 12.3-19, Rev. 57 AutoCAD Figure 12_3_19.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-523, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-20 replaced by dwg.

A-523, Sh. 1 FIGURE 12.3-20, Rev. 57 AutoCAD Figure 12_3_20.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-524, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-21 replaced by dwg.

A-524, Sh. 1 FIGURE 12.3-21, Rev. 57 AutoCAD Figure 12_3_21.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-525, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-22 replaced by dwg.

A-525, Sh. 1 FIGURE 12.3-22, Rev. 56 AutoCAD Figure 12_3_22.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-526, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-23 replaced by dwg.

A-526, Sh. 1 FIGURE 12.3-23, Rev. 57 AutoCAD Figure 12_3_23.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-527, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-24 replaced by dwg.

A-527, Sh. 1 FIGURE 12.3-24, Rev. 57 AutoCAD Figure 12_3_24.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-528, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-25 replaced by dwg.

A-528, Sh. 1 FIGURE 12.3-25, Rev. 57 AutoCAD Figure 12_3_25.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-529, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-26 replaced by dwg.

A-529, Sh. 1 FIGURE 12.3-26, Rev. 57 AutoCAD Figure 12_3_26.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

A-530, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-27 replaced by dwg.

A-530, Sh. 1 FIGURE 12.3-27, Rev. 48 AutoCAD Figure 12_3_27.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

BIN-100 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-28 replaced by dwg.

BIN-100 FIGURE 12.3-28, Rev. 55 AutoCAD Figure 12_3_28.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-137, Sh. 1FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-30-1 replaced by dwg.

M-137, Sh.

1FIGURE 12.3-30-1, Rev. 56 AutoCAD Figure 12_3_30_1.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-137, Sh. 2FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 12.3-30-2 replaced by dwg.

M-137, Sh.

2FIGURE 12.3-30-2, Rev. 56 AutoCAD Figure 12_3_30_2.doc

Radiation Protection Manager PC-F12.5-1.vsd FSAR Rev. 68 Figure Rev. 58 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2FINAL SAFETY ANALYSIS REPORT HEALTH PHYSICS ORGANIZATION FSAR FIGURE 12.5-1Radiological Operations SupervisorProgramsRadiological Operations SupervisorOperationsTechnical Staff

- RAM Shipping

- Dosimetry

- InstrumentsHP ForemenHP TechniciansRadiologicalSupervisor__________

ALARARadiologicalOperationsSupervisor___________

ALARA ALARA Staff - Radiological Work Schedulers

- ALARA Specialists

- Radiological Engineer