ML20049H287

Chapter 12 to Gessar, Radiation Protection

ML20049H287
Person / Time
Site:	05000447
Issue date:	02/12/1982
From:	GENERAL ELECTRIC CO.
To:
References
NUDOCS 8202230068
Download: ML20049H287 (100)

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 12.1 CONTENTS Section Title Page 12.1 ENSURING TilAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACIIIEVABLE (ALARA) 12.1-1 12.1.1 Policy Considerations 12.1-1 12.1.1.1 Design and Construction Policies 12.1-1 12.1.1.2 Operation Policies 12.1-2 12.1.1.3 Compliance with 10CFR20'and Regulatory Guides 8.8, 8.10 and 1.8 12.1-2 12.1.1.3.1 Compliance with Regulatory Guide 8.8 12.1-2 12.1.1.3.2 Compliance with Regulatory Guide 8.10 12.1-3 \\ 12.1.1.3.3 Compliance with Regulatory Guide 1.8 12.1-3 12.1.2 Design Considerations 12.1-3 12.1.2.1 General Design Consideration for ALARA Exposures 12.1-3 12.1.2.2 Equipment Design Considerations for ALARA Exposures 12.1-4 12.1.2.2.1 General Design Criteria 12.1-4 12.1.2.2.2 Equipment Design Considerations to Limit Time Spent in Radiation Areas 12.1-4 12.1.2.2.3 Equipment Design Considerations to Limit Component Radiation Levels 12.1-6 12.1.2.3 Facility Layout General Design Considerations for Maintaining Radiation Exposures ALARA 12.1-6 12.1.2.3.1 Minimizing Personnel Time Spent in Radiation Areas 12.1-6 12.1.2.3.2 Minimizing Radiation Levels in Plant Access Areas and Vicinity of Equipment 12.1-7 12.1.3 Operational Considerations 12.1-9 O 12.1-i/12.1-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 RADIATION PROTECTION 12.1 ENSURING TIIAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACIIIEVABLE ( ALARA) 12.1.1 Policy Considerations Administrative programs and procedures, in conjunction with facility design, ensure that the occupational radiation exposure to personnel will be kept As Low As Reasonably Achievable ( ALARA). 12.1.1.1 Design and Construction Policies The ALARA philosophy was applied during the initial design of the plant and implemented via internal design reviews. The design was reviewed in detail for ALARA considerations and was reviewed, updated and modified as necessary during the design phase as experience was gained from operating plants. Engineers reviewed the plant design and integrated the layout, shielding, ventilation and monitoring instrument designs with traffic control, security, access control and health physics aspects to ensure that the over-all design is conducive to maintaining exposures ALARA. All pipe routings containing radioactive fluids were reviewed as part of the engineering design effort. This ensured that lines expected to contain significant radiation sources are adequately shielded and properly routed to minimize exposure to personnel. Operating plant results were continuously integrated during the design phase of the Nuclear Island. Review by and input from various utilities was used in such areas as TIP drive and control rod drive maintenance, and the incorporation of the suppression pool cleanup system. t 12.1-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.1.1.1 Design and Construction Policies (Continued) (Description on onsite inspections to determine that the design and operation keeps radiation exposures ALARA is, where required, the responsibility of the Applicant.) 12.1.1.2 Operation Policies (Description of operational policies to maintain occupational doses ALARA is the responsibility of the Applicant.) 12.1.1.3 Compliance with 10CFR20 and Regulatory Guides 8.8, 8.10 and 1.8 Compliance of the Nuclear Island desigr, with Title 10 of the Code of Federal Regulations, Part 20 (10CFR20), is ensured by the compliance of the design and operation of the facility within the guidelines of Regulatory Guides 8.8, 8.10 and 1.8. 12.1.1.3.1 Compliance with Regulatory Guide 8.8 The design of the Nuclear Island fully meets the intent of Revision 1 of Regulatory Guide 8.8, and reflects the commitment of General Electric and its subcontractors. Examples of compliance with all items in Section C.3 of the regulatory guide are deline-j ated in Subsection 12.3.1 of this SAR. Design features of the l Nuclear Island allow the Applicant to comply easily with the l l recommendations of Subsection C.4 of the guide. For instance, provisions are made in systems such as the Reactor Water Cleanup System (RWCS) to allow flushing of the piping in shielded cubicles before entry, and to use remote reach rods. Breathing air headers are provided in areas where past experience indicates airborne radioactivity has been a problem. Design provisions allow for remote operation of fuel handling and radwaste cask filling. O 12.1-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (' 12.1.1.3.1 Compliance with Regulatory Guide 8.8 (Continued) (Further discussion of compliance with the regulatory guide, if required, is the responsibility of the Applicant.) 12.1.1.3.2 Compliance with Regulatory Guide 8.10 (Description of compliance with this regulatory guide is the responsibility of the Applicant.) 12.1.1.3.3 Compliance with Regulatory Guide 1.8 (Description of compliance with this regulatory guide is the responsibility of the Applicant.) 12.1.2 Design Considerations ) This subsection discusses the methods and features by which the policy considerations of Subsection 12.1.1 are applied. Provi-sions and designs for maintaining personnel exposures ALARA are presented in detail in Subsections 12.3.1 and 12.3.2. 12.1.2.1 General Design Consideration for ALARA Exposures General design considerations and method employed to maintain inplant radiation exposures ALARA, consistent with the recommenda-tions of Regulatory Guide 8.8, have two objectives: (1) minimizing tle necessity for and amount of personnel time spent in radiation areas, and (2) minimizing radiation levels in routinely occupied plant areas in the vicinity of plant equipment expected to require personnel attention. O 12.1-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.1.2.1 General Design Consideration for ALARA Exposures (Continued) Both equipment and facility designs are considered in maintaining exposures ALARA during plant operations. Events considered include normal operation maintenance and repairs, refueling operations and fuel storage, inservice inspection and calibrations, radioactive waste handling and disposal, etc. 12.1.2.2 Equipment Design Considerations For ALARA Exposures 12.1.2.2.1 General Design Criteria The engineering design procedures require that the component design engineer consider the applicable regulatory guides as a part of the design criteria. This includes Regulatory Guide 8.8. In this way, the radiation problems of a component or system are considered. A summary survey of the components designs was made to determine the factors considered. The following paragraphs cite some exampics of design considerations made to implement ALARA. 12.1.2.2.2 Equipment Design Considerations to Limit Time Spent in Radiation Areas (1) Equipment is designed to be operated and have its instrumentation and controls in accessible areas both during normal and abnormal operating conditions. Equipment such as the Reactor Water Cleanup (RWCS) System and the Fuel Pool Cleanup (FPCCU) System are remotely operated, including the backwashing and precoat operations. Other equipment has been redesigned in order to lengthen service life. For example, seal water is applied to the recirculation pump seals to keep them clean. This increased the maintenance interval from O 12.1-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.1.2.2.2 Equipment Design Considerations to Limit Time Spent \\s,/ in Radiation Areas (Continued) approximately 1 year to more than 3 years. Changes in the LPRM design changed the expected life from 1 to 3 years. (2) Equipment is designed to facilitate maintenance. Equipment such as the RHR heat exchanger is designed with an excess of tubes in order to permit plugging of some tubes. The heat exchanger has drains to allow draining of the shell side water. Some of the valves have stem packing of the cartridge type that can be easily replaced. The recirculation pipes have blocking valves in order to isolate the pump. The BWR/6 Mark III has a CRD maintenance machine that reduces manhours for removal or replacement by a factor of 5. Refueling f 3, tools are designed for drainage and with smooth surfaces \\s-) in order to reduce contamination. Vessel and piping insulation is of an easily removable type. (3) The materials selected for use in the system have been chosen to fulfill the environmental requirements. Valves, for example, use grafoil stem packing to reduce leakage and maintenance. (4) Past experience has been factored into current designs. The steam relief valves have been redesigned as a result of inservice testing. Access for inservice inspection has been changed. l s_- 12.1-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.1.2.2.3 Equipment Design Considerations to Limit Component Radiation Levels (1) Equipment and piping were designed to reduce the accumulation of radioactive materials in the equipment. The piping, where possible, was constructed of seamless pipe as a means to reduce radiation accumulation on the seam. The control rod drive volume tanks are sloped for drainage and flushing capability to reduce accumula-tion of radioactivity. The filter demineralizers in the RWCS and FPCCU Systems are backwashed and flushed prior to maintenance. (2) Equipment designs include provisions for limiting leaks or controlling the fluid that does leak. This includes piping the released fluid to the sumps and the use of drip pans with drains piped to the floor drains. (3) The materials selected for use in the primary coolant system consist mainly of austenitic stainless steel, carbon steel and low alloy steel components. (4) The system design includes a RWCS and a Condensate Domineralizer System on the reactor feedwater. These systems are designed to limit the radioactive isotopes in the reactor system. The systems are described in Subsections 5.4.8 and 10.4.6. 12.1.2.3 Facility Layout General Design Considerations for Maintaining Radiation Exposures ALARA 12.1.2.3.1 Minimizing Personnel Time Spent in Radiation Areas Facility general design considerations to minimize the amount of personnel time spent in radiation areas include the following: O 12.1-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O (A) 12.1.2.3.1 Minimizing Personnel Time Spent in Radiation Areas (Continued) (1) locating equipment, instruments, and sampling stations, which require routine maintenance, calibration, operation, or inspection, for ease of access and minimum of required occupancy time in radiation areas; (2) layir.g out plant areas to allow remote or mechanical operation, service, monitoring, or inspection of highly radioactive equipment; and (3) providing, where practicable, for transportation of equipment or components requiring service to a lower radiation area. 12.1.2.3.2 Minimizing Radiation Levels in Plant Access Areas gw and Vicinity of Equipment V Facility general design considerations directed toward minimizing radiation levels in plant access areas and in the vicinity of equipment rcquiring personnel attention include the following: (1) separating radiation sources and occupied areas where practicable (e.g., pipes or ducts containing potentially high radioactive fluids not passing through occupied areas); (2) providing adequate shielding between radiation sources and access and service areas; (3) locating equipment, instruments, and sampling sites in the lowest practicable radiation zone; Ov 12.1-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.1.2.3.2 Minimizing Radiation Levels in Plant Access Areas and Vicinity of Equipment (Continued) (4) providing central control panels to permit remote operation of all essential instrumentation and controls from the lowest radiation zone practicable; (5) where practicable for package units, separating highly radioactive equipment from less radioactive equipment, instruments, and controls; (6) providing means and adequate space for utilizing moveable shielding for sources within the service area when required; (7) providing means to control contamination and to facili-tate decontamination of potentially contaminated areas where practicable; (8) providing means for decontamination of service areas; (9) providing space for pumps and valves outside of highly radioactive areas: (10) providing remotely operated centrifugal discharge and/or backflushable filter systems for highly radioactive radwaste and cleanup systems; (11) providing labyrinth entrances to radioactive pump, equipment, and valve rooms; (12) providing adequate space in labyrinth entrances for easy access; and (13) maintaining ventilation air flow patterns from areas of lower radioactivity to areas of higher radioactivity. 12.1-8

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3 5 l i F I 12.1.3 Operational Considerations t i (Description of operational considerations to maintain doses to operators ALARA is the responsibility of the Applicant.) i i I i i I l [ l I 1 I i' I, l l \\O e i 6 l l I i l I r l i i l i i l O I 12.1-9/12.1-10 r

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 12.2 CONTENTS I i Section Title Page 12.2 RADIATION SOURCES 12.2-1 12.2.1 Contained Sources 12.2-1 ] 12.2.1.1 Source Terms 12.2-1 12.2.1.1.1 General 12.2-1 12.2.1.2 Reactor Building 12.2-2 2.2 1. .l.1 da o f om the Re ctor Core 1 22 l 12.2.1.2.1.1.1 General 12.2-2 12.2.1.2.1.1.2 Physical Data 12.2-2 12.2.1.2.1.1.3 Core Boundary Neutron Fluxes 12.2-3 12.2.1.2.1.1.4 Gamma Ray Source Energy Spectra 12.2-3 12.2.1.2.1.1.5 Gamma Ray and Neutron Fluxes l Outside the Vessel 12.2-4 12.2.1.2.1.1.6 Gamma Ray Doese Rates and Fast Neutron Fluxes at the Vessel 12.2-5 i 12.2.1.2.2 Radioactive Sources in the Reactor Water, Steam and Offgas 12.2-6 12.2.1.2.3 Radioactive Sources in the Engi-neered Safeguard Systems 12.2-6 12.2.1.2.4 Radioactive Sources in the Residual Heat Removal System 12.2-6 i 12.2.1.2.5 Radioactive Sources in Reactor Core Isolation Cooling System 12.2-7 12.2.1.2.6 Radioactive Sources in Radwaste Systems 12.2-7 12,2.1.2.6.1 Radioactive Sources in the Reactor Water Cleanup System 12.2-7 l 12.2.1.2.6.2 Radioactive Sources in Liquid Radwaste System 12.2-8 1 12.2.1.2.6.3 Radioactive Sources in the Gaseous Radwaste System 12.2-8 l 12.2.1.2.6.4 Radioactive Sources in the Solid Radwaste System 12.2-9 l () 12.2.1.2.6.5 Radioactive Sources in the Fuel Pool Cleanup System 12.2-9 1 12.2-i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) Section Title Page 12.2.1.2.6.6 Radioactive Sources in the Suppression Pool Cleanup System 12.2-9 12.2.1.2.7 Radioactive Sources in Piping and Main Steam Systems 12.2-9 12.2.1.2.7.1 Radioactive Sources in Main Steam System 12.2-9 12.2.1.2.7.2 Radioactive Crud in Piping and Steam Systems 12.2-10 12.2.1.2.8 Radioactive Sources in the Spent Fuel 12.2-10 12.2.1.2.9 Other Radioactive Sources 12.2-10 12.2.1.2.9.1 Reactor Startup Source 12.2-10 12.2.1.2.9.2 Radioactive Sources in the Control Rod Drive System 12.2-11 12.2.2 Airborne Radioactive Material Sources 12.2-11 12.2.2.1 Production of Airborne Sources 12.2-11 12.2.2.2 Airborne Sources for Relief Valve Venting 12.2-13 12.2.2.3 Airborne Sources During Refueling 12.2-13 12.2.3 References 12.2-14 l l l l l l l 12.2-ii i

22A7007 GESSAR II Rev. 0 238 NUCLEAR ISLAND ~ SECTION 12.2 TABLES Table Title Page 12.2-1 Basic Reactor Data 12.2-15 12.2-2 Core Boundary Neutron Fluxes 12.2-17 12.2-3 Gamma Ray Source Energy Spectra 12.2-18 12.2-4 Gamma Ray and Neutron Fluxes Outside the Vessel Wall 12.2-23, 12.2-5 Past Neutron Fluxes and Gamma Ray Dose Around the Reactor Vessel 12;.2-25 12.2-6 Fission Product Gamma Source Strength in h RilR IIcat Exchanger Shell (MeV/sec) }2.2-26 12.2-7 Fission Product Invetory in RIIR lleat Exchanger Shell Four liours After Shutdown 12.2-27 12.2-8 Reactor Coolant Concentration Equilibrium Values Entering RCIC Turbine (pCi/g) 12.2-28 12.2-9 Reactor Water Cleanup Backwash Receiving Tank Sources (Ci) 12.2-29 ) 12.2-10 Reactor-Water Cleanup Representative Ileat Exchanger Shell Side Source Terms (Ci) 12.2-30 12.2-11 Reactor Water Cleanup Nonregenerative IIeat Exchanger Radiation Sources (Ci) 12.2-31 12.2-12 Reactor Water Cleanup Regenerative Ileat Exchanger Radiation Radiatica Sources (Ci) 12.2-32 12.2-33 12.2-13 Liquid Radwaste Component Inventories 12.2-14 Offgas System Component Inventory Activities 12.2-43 12.2-15 Expected Solid Waste Average Radioactivity 12.2-63 Content e 12.2-16 Radioactive Sources in the Fuel Ecol Filter Demineralized System 12.2-65 12.2-17 Radioactive Sources in the Suppression Pool 12.2-66 Cleanup System 12.2-18 Radioactive Crud in Piping and Steam Systens 12.2-67 12.2-19 Radioactive Sources in Control Rod Drive 12.2-68 System 12.2-20 Annual Airborne Releases of Elemental Iodine-131 According to Plant Operating Mode for Environmental Impact Evaluations Elemental I-131 Release (mci /yr) 12.2-69 3. !2.2-iii j

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 f TABLES (Continued) Table Title Page 12.2-21 Annual Airborne Releases of Nonelemental Iodine-131 Species According to Plant Operating Mode for Environmental Impact Evaluations 12.2-70 12.2-22 Annual Airborne Release of Noble Gas and Iodine for Environmental Impact Evaluations (Ci/yr) 12.2-71 12.2-23 Annual Airborne Releases for Environmental Impact Evaluations (Ci/yr) 12.2-72 ILLUSTRATIONS Title Page O Figure 12.2-73 12.2-1 Radiation Source Model 1 O I 12.2-iv E

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( 12.2 RADIATION SOURCES %J 12.2.1 Contained Sources 12.2.1.1 Source Terms 12.2.1.1.1 General With the exception of the vessel and drywell shields, shielding designs are based on fission product and activation product sources consistent with Section 11.1. For shielding, it is con-servative to design for fission product sources at peak values rather than an annual average, even though experience supports a lower annual average than the design average (Reference 1). It should be noted that activation products, principally Nitrogen-16, control shielding calculations in most of the primary system. In areas where fission products are significant, conservative allow-rN ance is made for transient decay while at the same time providing iU) for transient increase of the noble gas source, daughter product formation and energy level of emission. Areas where fission products are significant relative to Nitrogen-16 include: (1) the condenser off-gas system downstream of the jet air ejector; (2) liquid and solid radwaste equipment; (3) portions of the RWCS; and (4) portions of the feedwater system downstream of the hotwell including condensate treatment equipment. For application, the design sources are grouped first by location and then by equipment type (e.g., reactor building, core sources). The following paragraphs represent the source data in various pieces of equipment throughout the plant. General locations of equipment are shovn in the general plant arrangement drawings of section 1.2. v 12.2-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.1.2 Reactor Building 12.2.1.2.1 Peactor Vessel Sources 12.2.1.2.1.1 Radiation from the Reactor Core 12.2.1.2.1.1.1 General The information in this section defines a reactor vessel model and the associated gamma and neutron radiation sources. This section is designed to provide the data required for calculations beyond the vessel. The data selected were not chosen for any given pro-gram, but were chosen to provide information for any of several shield program types. In addition to the source data, ca]culated radiation dose levels are provided at locations surrounding the vessel. These data are given as a potential check point for cal-culations by shield designers. 12.2.1.2.1.1.2 Physical Data Table 12.2-1 presents the physical data required to form the model in Figure 12.2-1. This model was selected to contain as few separate regions as possible to adequately portray the reactor. Table 12.2-1 provides nominal dimensions and material volume frac-tions for each boundary and region in the reactor model. To describe the reactor core, Table 12.2-1 provides thermal power, power density, core dimensions, core average material volume fractions and reactor power distributions. The reactor power dis-tributions are given for both radial and axial distributions. These data contain uncertainties in the volume regions near the edge of the core. The level of uncertainties for these regions is estimated at 20%. O 12.2-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 f-~) 12.2.1.2.1.1.3 Core Boundary Neutron Fluxes V Table 12.2-2 presents peak axial neutron multigroup fluxes at the core equivalent radius. The core-equivalent radius is a hypo-thetical boundary enclosing an area equal to the area of the fuel bundles and the coolant space between them. The peak axial flux occurs adjacent to the portion of the core with the greatest power. As shown by the data in Table 12.2-1, this point is below the core midplane. Since these data are calculated with a core-equivalent radius, the flux represents a mean flux in the azimuthal angle around the core. While the flux within any given energy group is not known within a factor of 2, the total cal-culated core boundary flux is estimated to be within 50%. 12.2.1.2.1.1.4 Gamma Ray Source Energy Spectra l Table 12.2-3 presents average gamma ray energy spectra per watt of /~N j (a) reactor power in both core and noncore regions. In Table 12.2-3, part A, the energy spectra in the core are presented. The energy spectra in the core represent the average gamma ray energy released l by energy group per watt of core thermal power. The energy spectra in MeV per sec per watt (MeV/sec/W) can be used with the total core power and power distributions to obtain the source in any part of the core. The gamma ray energy spectra include the fission gamma rays, the fission product gamma ray and the gamma rays resulting from inelastic neutron scattering and thermal neutron capture. The total gamma ray energy released in the core is estimated to be accurate to within 10%. The energy release rate above 6 MeV may be in error by as much as a factor of 12. Table 12.2-5, part B, gives a gamma ray energy spectrum in MeV/ sec/W in spent fuel as a function of time after operation. The () data were prepared from tables of fission product decay gamma 12.2-3 l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.1.2.1.1.4 Gamma Ray Source Energy Spectra (Continued) 0 fitted to integral measurements for operation times of 10

sec, or approximately 3.2 years.

To obtain shutdown sources in the core, the gamma ray energy spcetra are combined with the core thermal power and power distributions. Shutdown sources in a single fuel element can be obtained by using the gamma ray energy spectra and the thermal power the element contained during operation. Table 12.2-3, part C, gives the gamma ray energy spectra in the cylindrical regions of the reactor from the core through the 2 vessel. The energy spectra are given in terms of MeV/cm /sec/W at the inside surface and outside surfaces of the region. This energy spectrum, multiplied by the core thermal power, is the gamma ray source. The point on the inside surface of the region is the maximum point within the region. In the radial direction, the variation in source intensity may be approximated by an exponential fit to the data on the inside and outside surfaces of the region. The axial variation in a region can be estimated by using the core axial variation. The uncertainty in the gamma ray energy spectra is due primarily to the uncertainty in the neutron flux in these regions. The uncertainty in the neutron flux is estimated to vary from approximately tS0% at the core boundary to a factor of t3 at the outside of the vessel. The calculations were carried out with voids beyond the vessel. The presence of shield mate-rials beyond the vessel will cause in increase in the gamma source on the outside of the vessel. 12.2.1.2.1.1.5 Gamma Ray and Neutron Fluxes Outside the vessel Table 12.2-4 presents the maximum axial neutron and gamma ray fluxes outside the vessel. The maximum axial flux occurs on the vessel opposite the elevation of the core with the maximum power level. This elevation can be located using the data from Table 12.2-1. The fluxes at this elevation are based on a mean 12.2-4

GESSAR II 2'2A7007 238 NUCLEAR ISLAND Rev. 0 ,I 12.2.1.2.1.1.5 Gamma Ray and Neutron Fluxes Outside the Vessel (Continued) radius core and do not show azimuths angle variations. The calcu-lational model for these fluxes assumed no shield materials beyond the vessel wall. The presence of shield materials will signi-ficantly alter the neutron fluxes in the lower end of the neutron energy spectrum. The gamma ray calculations include gamma ray sources from all of the cylindrical regions between the center of the core and the edge of the vessel. While the uncertainties in a given energy group flux may be a factor of 13, the uncertainties in the total integral flux are estimated to be within a factor of two. 12.2.1.2.1.1.6 Gamma Ray Dose Rates and Fast Neutron Fluxes at the Vessel Table 12.2-5 presents calculated fast neutron fluxes greater than g,/ 1 MeV and gamma ray dose rates in rad per hour at three points outside of the vessel. The points include a point on the radius of the vessel at the elevation of the peak power in the core, a i point on the vessel centerline at the top head and a point on the vessel centerline at the bottom head. At the vessel radius, the calculated fast flux and gamma ray dose rates contain the same type of uncertainties as the total integral fluxes. This uncertainty is estimated to be within a factor of 2. The top head data in Table 12.2-5 have been treated in a very conserva-i tive manner. As a result, the data at the top head represent maxi-mum conditions. It is estimated that the flux at this location could be more that a factor of 3 lower. At the centerline posi-i tion on the bottom head, the neutron calculation represents an upper limit. At this point, the calculated neutron flux from the core was orders of magnitudes less than the data given in Table 12.2-5. The calculated gamma dose rate from the core is ) not known to better than a factor of 2. At the bottom head, there 12.2-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.1.2.1.1.6 Gamma Ray Dose Rates and Fast Neutron Fluxes at the Vessel (Continued) there will be a contribution to the gamma dose rate from accumula-tions of radioisotopes in the bottom of the vessel. This contri-bution will be dependent on the history of the plant. The radio-isotope contribution is not included in Table 12.2-5. 12.2.1.2.2 Radioactive Sources in the Reactor Water, Steam and Offgas T: ' radioactive sources in the reactor water, steam and of fgas are cos' red and discussed in Chapter 11 (Subsections 11.1.1 through 11.1.4). This material provides the concentrations during normal operation of the radioisotopes in the reactor vessel or leaving the reactor vessel. 12.2.1.2.3 Radioactive Sources in the Engineered Safeguard Systems The cngineered safeguard systems include the HPCS and the LPCS Systems, which are described in Section 6.3. During operation, the systems take suction from either the condensate storage tank or from the suppression pool. The radiation source in the equip-ment is the activity of the water transported through the system. During routine testing of the systems, the activity of the suction water is the average expected condition of the water in the sup-pression pool or the condensate storage tank. Following a loss-of-coolant accident, (LOCA), the activity in the suppression pool water is defined by Regulatory Guide 1.7. 12.2.1.2.4 Radioactive Sources in the Residual Heat Removal System The radioactive sources in the Residual Heat Removal ( RH R) System were calculated for the system operating in the reactor shutdown mode. In this mode, the system recirculates reactor coolant to 12.2-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.2.1.2.4 Radioactive Sources in the Residual lleat Removal \\s / System (Continued) remove reactor decay heat (Subsection 5.4.7). The R!lR System is operated from approximately 2-4 hours after shutdown until the end of the refueling period. The source in the RIIR System is the activity in the volume of reactor water contained in the system. This should include the increase of activity as a result of depressurization. The sources are provided in Tables 12.2-6 and 12.2-7. 12.2.1.2.5 Radioactive Sources in Reactor Core Isolation Cooling System The radioactive sources in the Reactor Core Isolation Cooling (RCIC) System were evaluated for the systems operating in the reactor shutdown mode. These systems may be utilized during reactor shutdown if the main condenser is unavailable. The sys-O (,/ tems are operated from the time of reactor shutdown for approxi-mately 2 hr until a reactor pressure of 50 psig for the RCIC is achieved. Below 150 psig, the RCIC flow decreases. The source in the system is the activity in the volume of reactor water and steam contained in the system. During routine testing of the system, the source in the equipment is the activity of the steam driving the system turbine. This activity is controlled by the Nitrogen-16. The radiation source data used in the shield design for this system is shown in Table 12.2-8. 12.2.1.2.6 Radioactive Sources in Radwaste Systems 12.2.1.2.6.1 Radioactive Sources in the Reactor Water Cleanup System {) The radioactive sources are the result of the activity in the reactor water in transit through the system or accumulation of 12.2-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.1.2.6.1 Radioactive Sources in the Reactor Water Cleanup System (Continued) radioisotopes removed from the water. Components for this system include regenerative and nonregenerative heat exchangers, pumps, valves, filter demineralizers and the backwash receiving tank (Subsection 5.4.8). The accumulated sources in the filter demin-eralizers, backwash receiving tanks and heat exchangers are given in Tables 12.2-9 through 12.2-12. The source is present in the filters and receiving tank during all modes of operation. Therefore, backwashing capability is provided to remove the residual activity for effective radwaste handling. 12.2.1.2.6.2 Radioactive Sources in Liquid Radwaste System The Liquid Rai.;aste System is composed of three subsystems designed to collect, treat and recycle or discharge different categories of waste water (Subsection 11.2.2). The radioactive sources for the components in the systems are provided in Table 12.2-13. The isotopic inventories in the liquid radwaste components were calculated assuming a fission product release rate from the fuel equivalent to that required to produce 100,000 pCi/see of offgas following a 30-min holdup period. 12.2.1.2.6.3 Radioactive Sources in the Gaseous Radwaste System The gaseous effluent treatment systems are designed to limit the dose to offsite persons from routine station release. The off-gases are treated through the use of a ca.talytic recombiner and low-temperature charcoal adsorption (RECHAR) system (Sub-section 11.3.2). The system is designed to handle an annual average noble gas release equivalent to 100,000 pCi/sec after a 30-min delay. The accumulation of gaseous radioisotopes and the solid daughter products resulting from the decay of the noble gases is given in Table 12.2-14. The inventory in the components has 12.2-8

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ,f'% 12.2.1.2.6.3 Radioactive Sources in the Gaseous Radwaste System C) (Continued) been evaluated for several possible operating modes. In all cases, a 1-yr operating time has been used to accumulate the decay activities. This is sufficient time for most isotopes to reach equilibrium. 12.2.1.2.6.4 Radioactive Sources in the Solid Radwaste System The solid radwaste system provides the capability for solidifying and packaging waste from the other radwaste systems (Sub-section 11.4.2). The wastes are not solidified separately by type or source. The final waste is placed in a steel container. The expected average radioactivity content of the solid waste per container is given in Table 12.2-15. (~'} 12.2.1.2.6.5 Radioactive Sources in the Fuel Pool Cleanup System LJ The radiation source data used in the shield design of the Fuel Pool Cleanup (FPCCU) System filter demineralizer system is given in Table 12.2-16. 12.2.1.2.6.6 Radioactive Sources in the Suppression Pool Cleanup System b The radiation source data used in the shield of the Suppression Pool Cleanup (SPCU) System is given in Table 12.2-17. 12.2.1.2.7 Radioactive Sources in Piping and Main Steam Systems 12.2.1.2.7.1 Radioactive Sources in Main Steam System All radioactive materials in the Main Steam System result from radioactive sources carried over from the reactor during plant 7-k_,/ operation. In most of the components carrying live steam, the 12.2-9 l i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.1.2.7.1 Radioactive Sources in Main Steam System (Continued) source is dominated by Nitrogen-16. In components where N-16 has dccayed, the other activities carried by the steam become sig-nificant. During plant shutdown, there is a residual activity resulting from prior plant operations. These data will be pro-vided by the Applicant. 12.2.1.2.7.2 Radioactive Crud in Piping and Steam Systems The inside surfaces of the piping and all reactor and power sys-tems components become coated with activated corrosion products, commonly called crud. The quantity of crud on the components is dependent on a number of factors, including power history, water quality and fuel experience. The piping and components carrying reactor water are coated with higher levels of crud than piping and components carrying steam. Table 12.2-18 shows the crud accumulation data used in the design of this plant. 12.2.1.2.8 Radioactive Sources in the Spent Fuel The radiation source for spent fuel is given in Sub-section 12.2.1.2.1.1.4 (Table 12. 2-3) in terms of MeV/sec/W. The design calculation is carried out for a mean element for an appropriate decay time. 12.2.1.2.9 Other Radioactive Sources 12.2.1.2.9.1 Reactor Startup Source 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 O 12.2-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 [ } 12.2.1.2.9.1 Reactor Startup Source (Continued) \\_/ source remains within the reactor for its lifetime. Thus, no unique shielding requirements are required after reactor operation. 12.2.1.2.9.2 Radioactive Sources in the Control Rod Drive System The control rod drive (CRD) source term data are provided in Table 12.2-19. The system is described in Subsection 3.9.4. 12.2.2 Airborne Radioactive Material Sources This subsection deals with the source models and parameters required to evaluate airborne concentrations of radionuclides during plant operations in various plant radiation areas where personnel occupancy is expected. v\\ 12.2.2.1 Production of Airborne Sources Design efforts are directed towards keeping contained all the radioactive material, whether it is in a solid, liquid or gaseous form; however, the unavoidable leaks from process systems and some processes of refueling and decontamination lead to airborne radioactivity. Leakage of fluids from the process system will result in the release of radionuclides into plant buildings. In general, the noble radiogases will remain airborne and will be released to the atmosphere with little delay via the building ventilation exhaust ducts. The radionuclides will partition between air and water to approach equilibrium conditions. Airborne iodines will "plateout" on most surfaces, including pipe, concrete, and paint. A sig-nificant amount of radiciodine remains in air or is desorbed from O 12.2-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.2.1 Production of Airborne Sources (Continued) surfaces. Radiciodinec are fottd in ventilation air as methyl iodide and as inorganic iodine which is here defined as particu-late, elemental and hypoiodous acid forms of iodine. Particulates will also be present in the ventilation exhaust air. The average annual release of I-131 is 1356 millicuries (mci) in elemental form. Other forms of I-131 amount of 2782 mci /yr. These forms of I-131 include hypoiodous acid, particulates and methyl iodide. The basis for these releases is as follows: (1) a calendar year consisting of 300 days of power opera-tions and one refueling / maintenance shutdown period; (2) a concentration of I-131 in reactor water of 31.2 pCi/kg; (3) a carryover of I-131 from reactor water to steam of 1.5%; (4) forward-pumped heater drains; (5) use of " clean" steam for the turbine gland seals; (6) a noble gas release rate of 100,000 pCi at t = 30 min and an I-131 release rate of 700 pCi/sec at t = 0, and (7) 24 drywell purges per year, 365 hours between each purge. The airborne radiological releases from building heating, ven-tilating, and air conditioning and the main condenser mechanical vacuum pump have been compiled and evaluated in References 3 and 5. O 12.2-12

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (s) 12.2.2.1 Production of Airborne Sources (Continued) Based upon the above conditions and values in References 2 and 4, airborne releases to the environment are summarized in Tables 12.2-20 through 12.2-23. Approximately 15,800 Ci/ plant /yr of noble radiogases are released; one-half of this total is released from the turbine building. Nineteen particulates have been identified. The total particulate release rate per plant is approximately 0.4 Ci/yr; the annual release of Co-60 is less than 0.03 Ci. 12.2.2.2 Airborne Sources for Relief Valve Venting A special consideration for ventilation source terms is.he activity release to the containment via relief valve discharge to the suppression pool. The dif ferent modes of this type of dis-() charge and the frequency of occurrence are discussed in Chapter 15. From measurements, it has been established that a correlttion can be made between activity release and reactor pressure change. A mathematical model has been developed which characterizes shutdown release transients to the suppression pool. 12.2.2.3 Airborne Sources During Refueling I The airborne radioactivity during refueling in the containment is expected to be similar to that observed in operating stations. Experience at operating BWR has shown that airborne radioactivity can result from the water in the reactor cavity exceeding 1000F and flaking of cobalt dioxide (coo ) fr m the dryer and separator 2 l if their surfaces are allowed to dry. Other potential airborne sources could occur during vessel head venting and fuel movement. The airborne radioactive material sources resulting from reactor vessel head and internals removal have been determined from opera-(O ting plant experience. The major radioisotopes found were I-131, j 12.2-13

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.2.2.3 Airborne Sources During Refueling (Continued) Co-60, and Mn-54, with Nb-95, Zr-95, Ru-103, and Ce-144 at moder-ate concentrations, and with Ce-141, Cs-137, Co-56, and Cr-51 at low concentrations. The radioactive particulates ranged as high ~0 -8 as 2 x 10 pCi/cc and the I-131 as high as 4 x 10 pCi/cc. To minimize the containment airborne radioactivity contribution due to removal of the reactor pressure vessel head: (1) the steam dryer and separator surfaces will be kept wet or covered; (2) the fuel pools are cooled through heat exchangers of large capacity; and the (3) ventilation system on the refueling pool is designed to sweep air from the pool surface and remove a large ll portion of potential airborne contamination. 12.2.3 References 1. J.E. Smith, " Noble Gas Experience in Boiling Water Reactors", Paper No. A-54, presented at Noble Gases Symposium, Las Vegas, Nevada, September 24, 1974. 2. " Airborne Releases from BWRs for Environmental Impact Evaluations", NEDO-21159-2 (1977). 3. American Nuclear Society, ANS-18.1, ANSI N237-197 (approved May ll, 1976), Table 5. 4. " Airborne Releases from BWRs for Environmental Impact Evaluations", NEDO-21159, March 1976. 5. " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors" (BWR-GALE Code) U.S. NRC NUREG-0016 Rev. 1, January 1979. O 12.2-14

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 fg Table 12.2-1 %/ BASIC REACTOR DATA A. REACTOR THERMAL POWER: 3579 MW B. AVERAGE POWER DENSITY: 54.07 W/cm3 C. PHYSICAL DIMENSIONS

• Radii in.

1. Core Equivalent Radius 92.58 2. Inside Shroud Radius 99.90 3. Outside Shroud Radius 101.90 4. Inside Vessel Radius - Nominal 119.0 5. Outside Vessel Radius - Nominal 125.0 6. Outside Vessel Radius - Reinforced - Nominal 125.75 7. Shroud Head Inside Radius 192.0 8. Vessel Top Head Inside Radius 119.0 9. Vessel Bottom Head Inside Radius 130.19 Elevation in. 7 1 10. Outside of Vessel Bottom Head -7.75 / 11. Inside of Vessel Bottom Head -0.25 12. Vessel Bottom Head Tangent 129.94 13. Bottom of Core Support Plate 202.56 14. Top of Core Support Plate '.04.56 15. Bottom of Active Fuel 213.50 16. Top of Reinforced Vessel Wall 210.00 17. Top of Active Fuel 363.5 18. Bottom of Top Guide 371.31 19. Top of Fuel Channel 377.87 20. Shroud Head Tangent 424.23 21. Inside of Shroud Head 452.27 22. Outside of Shroud Head 454.27 23. Normal Vessel Water Level 566.6 24. Top of Steam Dryer 720.63 25. Vessel Top Head Tangent 727.0 26. Inside of Vessel Top Head 846.0 27. Outside of Vessel Top Head 849.0 ()

• Figure 12.2-1 shows the relative locations of the dimensions.

12.2-15

GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. O Table 12.2-1 (Continued) D. MATERIAL DENSITIES * - gra-s/cm of REGION VOLUME Region Coolant "2 Zircaloy 304L Stainless A 0.740 0.0 0.0 0.178 B 0.338 0.0 0.0 4.349 C 0.318 2.334 0.978 0.056 C-1 0.597 0.0 0.166 1.697 C-2 0.234 0.0 1.099 0.255 D 0.240 0.0 1.004 1.209 E 0.390 0.0 0.0 0.0 F 0.669 0.0 0.0 0.200 G 0.036 0.0 0.0 0.0 11 0.740 0.0 0.0 0.0 I 0.740 0.0 0.0 0.26 E. TYPICAL CORE POWER DISTRIBUTIONS Axial Power Distribution Radial Power Distribution (Typical End-of--Life) Percent of Equivalent Relative Elevation ** Relative Radius Power (in.) Power 0 1.2 -75 0.343 20 1.2 -68 0.755 35 1.19 -60 1.055 50 1.17 -48 1.190 60 1.15 -36 1.200 70 1.12 -24 1.190 80 1.05 -12 1.170 85 0.995 0 1.155 90 0.778 12 1.140 92.5 0.590 24 1.105 95.0 0.430 36 1.055 97.0 0.375 48 0.945 98.0 0.395 60 0.715 99.0 0.432 68 0.462 100.0 0.518 75 0.212

• Figure 12.2-1 shows the region locations.
  • Elevations are measured from the mid plate of the core.

12.2-16

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 12. 2'-2 CORE BOUNDARY NEUTRON FLUXES Neutron Flux Energy Bounds (neutrons /cm2-sec) 16.5 MeV 3.9 E+10 10.00 MeV 5.5 E+11 l 6.065 MeV 2.0 E+12 3.679 MeV 3.8 E+12 2.231 MeV 4.4 E+12 1.353 MeV 3.9 E+12 C20.8 kev 3.8 E+12 497.9 kev 2.6 E+12 302.0 kev 2.3 E+12 183.2 kev 3.2 E+12 67.38 kev 2.2 E+12 24.79 kev 2.2 E+12 () 9.119 kev 2.0 E+12 3.355 kev 2.0 E+12 1.234 kev 1.9 E+12 454.0 eV 2.0 E+12 167.0 eV 1.9 E+12 61.44 eV 1.8 E+12 22.60 eV 8.8 E+11 13.71 eV 8.8 E+11 8.315 eV 8.2 E+11 5.043 eV 8.4 E+11 3.059 eV 8.3 E+11 1.855 eV 8.2 E+11 1.125 eV 8.8 E+11 0.616 eV 3.2 E+13 0.00 eV O 12.2-17

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 12.2-3 GAMMA RAY SOURCE ENERGY SPECTRA A. GAMMA RAY SOURCES IN THE CORE DURING OPERATION Energy Bounds Gamma Ray Source (MeV) (MeV/sec-W) 16.5 7.8 E+8 8.0 7.3 E+9 6.0 5.9 E+10 4.0 5.8 E+10 3.0 5.2 E+10 2.6 6.7 E+10 2.2 7.2 E+10 1.8 8.3 E+10 1.4 9.1 E+10 1.0 7.5 C+10 0.75 6.8 E+10 0.5 6.1 E+10 0.25 9.8 E+10 0.003 0 12.2-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 T Table 12.2-3 (Continued) (V Regi n I (Jet Pumps) Vessel Energy Bounds (MeV) Inside Outside Inside Outside 16.5 1.4 2.2 E-2 2.2 E-1 2.1 E-4 8.0 4.6 6.9 E-2 2.1 1.6 E-3 6.0 1.3 2.0 E-2 5.6 E-1 1.6 E-3 4.0 6.3 E-1 9.3 E-3 2.8 E-1 2.0 E-3 3.0 2.6 E-1 3.7 E-3 1.0 E-1 1.2 E-3 2.6 6.1 4.7 E-2 4.7 E-2 1.1 E-3 ( l.8 E-1 2.4 E-3 5.3 E-2 1.4 F-3 1.8 4.2 E-1 6.0 E-3 1.8 E-1 1.1 E-3 1.4 2.1 E-1 2.9 E-3 7.5 E-2 9.3 E-4 1.0 3.2 E-1 3.6 E-3 9.1 E-2 5.5 E-3 0.75 1.6 E-1 2.4 E-3 6.4 E-2 4.2 E-5 0.50 6.6 E-1 9.9 E-3 2.6 E-1 2.0 E-4 0.25 l 5.1 E-1 7.6 E-3 1.9 E-2 1.8 E-4 0.003 O 12.2-21/12.2-22

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 /~'s Table 12.2-4 GAMMA RAY AND NEUTRON FLUXES OUTSIDE THE VESSEL WALL A. NEUTRON FLUXES Energy Bounds Neutrons /cm -sec 16.5 MeV 5.8 E+6 10.00 MeV 2.9 E+7 6.065 MeV 2.2 E+7 3.679 MeV 4.5 E+7 2.231 MeV 7.5 E+7 1.353 MeV 1.1 E+8 820.8 l.6 E+8 497.9

V 1.5 E+8 302.0 kev 9.1 E+7 183.2 kev 1.1 E+8 67.38 kev 1.2 E+7 24.79 kev 6.7 E+7 O

\\,,/ 9.11 kev 1.4 E+7 3.355 kev 8.5 E+6 1.234 kev 6.4 E+6 454.0 kev 2.9 E+6 167.0 eV 4.2 E+6 61.44 eV 3.9 E+6 l 22.60 eV 1.9 E+6 13.71 eV 2.0 E+6 l 8.315 eV 1.8 E+6 5.043 eV l.6 E+6 3.059 eV 1.5 E+6 1.855 eV 1.4 E+6 l 1.125 eV 7.9 E+5 0.616 eV 6.0 E+5 0.000 eV l m N..)) I I 12.2-23 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-4 (Continued) B. GAMMA RAY ENERGY PLUXES Energy Bounds 2 MeV MeV/cm -sec 16.5 1.0 E+9 8.0 3.4 E+9 l i 6.0 l 3.3 E+9 4.0 t l 1.7 E+9 3.0 l 7.0 E+8 2.6 1.0 E+8 2.2 6.9 E+8 1.8 6.1 E+8 1.4 5.3 E+8 l 1.0 l 3.2 E+8 1 O.75 4.2 E+8 0.50 1 4.0 E+8 1 1 0.25 1 l.5 E+7 0.003 9 l l 12.2-24

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i Table 12.2-5 i FAST NEUTRON FLUXES AND GAMMA RAY DOSE j AROUND THE REACTOR VESSEL

• 1 1

Gamma Ray Fast Neutron Flux Dose Rate 2 (n/cm -sec) (rads /hr) Location (>l.0 MeV) Core - at Peak Axial Location 2.7 E+8 2.0 E+4 Above Top Head 1.0 E+2 2.2 E+1 Below Bottom Head <0.1 1.7 E-l t l

• The data presented in thie table are calculations of the primary and secondary radiation from the core vessel.

In power plant applications, the neutron fluxes above and below the core will include contributions from radiation exiting the vessel at mid- ] plane and being scattered to these locations. O 4 1 = i i i I i l l l ,_) l 12.2-25 1 ,r.. -,.--.m.,,--r . =. - r..~. .m.

i i Table 12.2-6 FISSION PRODUCT GAMMA SOURCE STRENGTH IN RHR HEAT EXCHANGER SHELL (MeV/sec) l (LATER) I j u W CD l om MM g MM i g y a HH i MH =* l i l l l u %N O> <: 4 O O OM 9 e

.. -_ _.- - -. ~. - _. l GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. O f 4 Table 12.2-7 FISSION PRODUCT INVENTORY IN RHR HEAT EXCHANGER SHELL FOUR HOURS AFTER SHUTDOWN (UCi) (LATER) l 3 Y I l' a r I .i I f } 4 O i i l l t i 4 ' 9 1 12.2-27 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-8 REACTOR COOLANT CONCENTRATION EQUILIBRIUM VALUES ENTERING RCIC TURBINE (uCi/g) Isotope Concentration Isotope Concentration N-13 6.6E-03 Kr-88 1.lE-02 N-16 4.8E-01 Kr-69 1.9E-04 N-17 1.5E-02 Xe-131m 3.7E-05 F-18 4.0E-03 Xe-133m 1.9E-04 0-19 6.9E-01 Xe-133 4.7E-03 Na-24 2.3E-06 Xe-135m 6.0E-03 Mn-56 5.2E-04 Xe-135

1. 4 E -0 2 Co-58 5.9E-06 Xe-137 7.3E-04 Br-83 5.lE-04 Xe-138 1.7E-02 Br-84 6.7E-04 Sr-91 3.4E-04 I-131 4.3E-04 Sr-92 3.4E-04 I-132 4.7E-03 Tc-99 1.6E-04 I-133 3.lE-03 Tc-101 5.3E-05 I-134 6.8E-03 Cs-138 1.4E-02 I-135 4.8E-03 Ba-139 2.8E-03 Kr-83m 2.lE-03 Ba-140 1.9E-05 Kr-85m 3.4E-03 Ba-141 5.8E-04 Kr-85 1.2E-05 Ba-142 1.3E-04 Kr-87 9.7E-03 Np-239 2.5E-04 O

12.2-28

- - - -.- - - -.. -. ~. - ~ - - -. - O O O t Table 12.2-9 i REACTOR WATER CLEANUP BACKWASH RECEIVING TANK SOURCES (Ci) 4 1 Source Volume = 2600 gal. Total Curies = 2400 Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies i l BR-83 5.0E 00 SR-89 2.8E 01 ZR-95 3.7E-01 NA-24 3.0E 00 BR-84 1.8E 00 SR-90 2.2E 00 ZR-97 4.9E-02 P-32 1.9E-01 BR-85 6.0E 02 SR-91 6.4E 01 'NB-95 1.2E 00 CR-51 5.lE 00 l I-131 1.7E 02 SR-92 3.lE 01 RU-103 1.8E-01 MN-54 4.4E 01 I-132 1.2E 02 Y-90 2.2E 00 RU-106 -2.5E-02 MN-56 1.2E 01 t I-133 2.8E 0? Y-91M 4.4E 01 RH-103M-1.8E-01 CO-58 5.3E 01 m I-134 2.7E 01 MO-99 1.lE 02 RH-106 2.5E-02 CO-60 5.5E 00 z i j i [ I-135 1.3E 02 TC-99M 4.7E 01 LA-140 8.2E 01 FE-59 8.6E-01 8@ TC-101 3.8E 00 CE-141 3.3E-01 NI-65 7.5E-02 gm t 7 Total 7.3E 02 .TE-129M 3.0E 00 CE-143 9.9E-02 ZN-65 2.2E-02 l TE-132 7.4E 01 CE-144 3.3E ZN-69M 4.lE-02 55 l CS-134 1.5E 00 PR-143 '3.lE-01 AG-110M 6.7E-01 HH 1 (" CS-136 8.5E 01 ND-147 1.lE-01 W-187 7.lE 00 CS-137 2.3E 00 z CS-138 1.lE 01 Total 8.5E 01 Total 8.8E 01 o i l BA-137M 2.3E 00 i. BA-139 2.4E 01 BA-140 7.lE 01 l BA-141 5.9E 00 BA-142 3.lE 00 NP-239 1.0E 03 l Total 1.5E 03 i l The source term in each of.the filter demineralizers is one half the source term in the ,M backwash receiving tank. g> 8 04 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-10 REACTOR-WATER CLEANUP REPRESENTATIVE HEAT EXCilANCER S!! ELL SIDE SOURCE TERMS (Ci) Total 1.263E-00 3T 1.3E-03 97NBM 1.5E-03 135XEM 9.5E-01 190 4.lE-03 99TCM 1.7E-03 135XE 1.6E-01 83KRM 4.lE-02 101TC 2.0E-03 137XE 1.6E-02 85KRM 1.0E-03 102TCM 1.5E-03 137BAM 2.2E-03 87KR 2.lE-03 132I 3.0E-03 138XE 3.lE-03 89KR 4.9E-03 133I 1.7E-03 130CS 1.0E-03 91YM 4.2E-03 f 133XEM 1.5E-03 139BA 1.2E-03 92SR 1.5E-03 133XS 2.3E-02 141BA 2.lE-03 92Y 1.lE-03 134I 5.4E-03 142BA 1.9E-03 93SR 1.9E-03 135I 2.8E-03 239NP 2.5E-03 O i i l I l l l O l r l l 12.2-30

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-11 REACTOR WATER CLEANUP NONREGENERATIVE HEAT EXCHANGER RADIATION SOURCES (Ci) 1 Total 3.582E-00 3T 1.0E-03 93SR 1.7E-01 134TE 6.7E-03 13NO* 4.3E-02 94SR 7.5E-02 134IM 1.4E-02 16NO 3.lE-03 94Y 2.lE-02 134I 4.3E-01 18P 3.7E-03 95SR 1.0E-02 135I 2.2E-01 190 3.0E-02 95Y 2.7E-02 135XEM 6.0E-03 24NA 1.9E-03 97NBM 2.4E-02 135XE 1.5E-03 56MN 4.6E-02 97NB 4.lE-03 136TE 1.2E-03 58CO 4.6E-03 99NBM 1.3E-02 136IM 1.6E-02 83SEM 2.4E-03 99TCM 1.4E-01 136I 5.5E-02 83BR 2.6E-02 101MO 1.2E-02 137I 5.8E-03 84SE 5.8E-03 10lTC 1.6P-01 137XE 1.6E-02 848R 5.5E-02 102MO 1.3E-02 138XE 2.6E-03 ( 85SE 1.9E-03 102TCM 1.lE-01 138CSM 5.3E-03 85BR 4.0E-02 103TC 2.3E-02 138CS 2.7E-02 87BR 2.3E-02 103RHM 1.9E-03 139XE 1.2E-03 87KR 1.5E-03 104MO 1.lE-02 139CS 8.9E-02 88RB 3.6E-03 104TC 5.5E-02 139BA 8.3E-02 89KR 5.4E-03 105MO 3.8E-03 140CS 3.9E-02 1 89RB 3.0E-02 105TC 2.9E-02 140BA 6.7E-03 l 89SR 2.1E-03 106TC 1.7E-03 141CS 6.6E-03 ( 90KR 1.4E-03 131SB 3.7E-03 141HA 1.8E-01 90RBM 1.7E-02 131TE 4.2E-03 141LA 2.2E-03 90RB 5.lE-02 131I 1.9E-02 142HA 1.7E-01 91RB 3.7E-02 132I 2.4E-01 142LA 03 91SR 5.3E-02 133SB 1.4E-02 143LA

1. ; '..- 0 2 l

91YM 2.4E-03 133 TEM 3.5E-03 144LA 2.4E-02 92SR 1.2E-01 133TE 9.5E-03 187W 2.8E-03 92Y 1.6E-03 133I 1.4E-01 239NP 2.0E-01 12.2-31

GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. O Table 12.2-12 REACTOR WATER CLEANU? REGENERATIVE IIEAT EXCllANGER RADIATION RADIATION SOURCES (Ci) Total 2.417E-00 13NO* 2.50-02 94Y 1.lE-02 134I 2.5E-01 16AM** 6.3E-03 95SR 1.8E-02 135I 1.3E-01 16NO 8.0E-02 95Y 1.6E-02 135XEM 2.8E-03 18F 2.lE-03 97NBM 2.4E-02 136TE 1.7E-03 190 5.2E-02 97NB 2.3E-03 136IM 1.8E-02 24NA 1.lE-03 99NBM 9.4E-03 136I 4.5E-02 56MN 2.7E-02 99NB 3.0E-03 1371 1.lE-02 59CO 2.7E-03 99TCM 7.9E-02 137XE 1.0E-02 83SEM 2.lE-03 10lMO 7.2E-03 138XE 1.6E-03 83BR 1.5E-02 10lTC 9.3E-02 138CSM 3.7E-03 84SE 3.9E-03 102MO 7.5E-03 138CS 1.6E-02 84BR 3.2E-02 102TCM 7.2E-02 139XE 1.5E-03 85SE 2.4E-03 103TC 2.4E-02 139CS 5.5E-02 85BR 2.8E-02 103RilM 1.lE-03 139BA 4.8E-02 87BR 2.3E-02 104MO 8.5E-03 140CS 3.6E-02 88BR 3.0E-03 104TC 3.3E-02 140BA 3.9E-03 88RB 2.10-03 105MO 3.8E-03 141CS 1.2E-02 89KR 3.7E-03 105TC 1.8E-02 141BA 1.0E-01 89RB 1.7E-02 106TC 2.2E-03 141LA 1.0E-03 l 89SB 1.2E-03 131SB 2.2E-03 142CS 1.lE-01 90RB 2.lE-03 131TE 2.4E-03 142CS 1.0E-03 90RBM 1.2E-02 131I 1.lE-02 143BA 3.7E-03 j 90RB 3.5E-02 1321 1.4E-01 143LA 1.lE-02 l 91RB 3.6E-02 133SB 9.7E-03 144BA 1.3E-03 91SB 3.lE-02 133 TEM 1.9E-03 144LA 3.0E-02 91YM 1.lE-03 133TE 5.5E-03 187W l.6E-03 92SR 6.8E-02 133I 8.lE-02 239NP 1.lE-01 l 93SR 1.lE-01 134TE 3.9E-03 l 94SR 6.4E-02 134IM 9.3E-03 l

• NO indicates isotopes of nitrogen as oxides.
  • AM indicates isotopes of nitrogen ammonia, Nil3*

l 12.2-32 l

-...~...-- -.- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( I {g Table 12.2-13 LIQUID RADWASTE COMPONENT INVENTORIES i l The inventory in the liquid radwaste components l is provided in the following table for a deep j bed system. The data in Table 12.2-13 were j generated assuming a fission product release l l from the fuel equivalent to that required to l l l produce 100,000 pCi/sec of offgas following a 30-min holdup period. 1 I l O i i l l l { I l i 12.2-33 f l

l t 1 i Table 12.2-13 (Continued) i CONCENTPATED WASTE TANK A700 l I j Source volume = 3000 gal. normal, 25,000 gals. full 1 Total Curies = 29 j Soluble Fission Insoluble Fission Activation ( Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 2.5E-04 SR89 9.4E-02 ZR-95 1.2E-03 NA-24 4.0E-04 BR-84 1.9E-06 SR-90 8.9E-03 ZR-97 7.5E-06 P-32 4.4E-04 BR-85 4.7E-10 SR-91 4.8E-03 NB-95 3.7E-03 CR-51 1.4E-02 w I-131 2.6E-01 SR-92 2.6E-04 RU-103 5.2E-04 MN-54 1.5E-03 l I-132 6.2C-02 Y-90 8.9E-03 RU-106 8.7E-05 MM-56 9.5E-05 2 l s I-133 1.5E-00 Y-91M 3.3E-03 RH-103M 5.2E-04 CO-58 1.7E-01 CO l I-134 1.0E-04 MO-99 7.3E-02 RH-106 8.7E-05 CO-60 2.0E-02 05 l y I-135 7.3E-02 TC-99M 1.7E-03 LA-140 1.8E-01 FE-59 2.6E-03 f u TC-101 2.7E-07 CE-141 1.0E-03 NI-65 5.6E-07 W% i TOTAL 2.8E-01 TE-129M 7.9E-03 CE-143 3.lE-05 ZN-65 8.4E-05 TE-132 5.9E-02 CE-144 1.2E-03 ZN-69M 5.0E-06 CS-134 6.lE-03 PR-143 6.9E-04 AG-110M 2.3E-03 k l CS-136 1.9E-03 ND-147 2.2E-04 W-187 1.6E-03 o CS-137 9.4E-03 CS-138 3.9E-06 TOTAL 1.9E-01 TOTAL 2.lE-01 BA-137M 9.4E-03 i BA-139 5.4E-05 BA-140 1.6E-01 [ BA-141 6.8E-07 BA-142 1.3E-07 TOTAL 1.0E-00 w xw o> <W . O O 04 l 9 9 e

O O O Table 12.2-13 (Continued) LOW CONDUCTIVITY OIL SEPARATOR D035 Source Volume = 6300 gal. Total Curies = 10 Soluble Fission Insoluble Fission Activation Products Products Halogens Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 1.0E-01 SR-89 2.5E-02 ZR-95 3.3E-04 NA-24 1.4E-02 BR-84 6.0E-02 SR-90 1.9E-03 ZR-97 2.lE-04 P-32 1.8E-04 BR-85 3.2E-03 SR-91 4.4E-01 NB-95 1.0E-03 CR-51 4.6E-03 I-131 3.2E-01 SR-92 5.7E-01 RU-103 1.6E-04 MN-54 3.9E-04 m I-132 9.lE-01 Y-90 1.9E-03 RU-106 2.2E-05 MN-56 2.4E-01 ~ I-133 1.0E-00 Y-91M 3.0E-01 RH-103M 1.6E-04 CO-58 4.7E-02 g@ r I-134 8.4E-01 MO-99 1.5E-01 RH-106 2.2E-05 CO-60 4.8E-03 &m N I-135 1.2E-00 TC-99M 4.8E-01 LA-140 7.6E-02 FE-59 7.7E-04 TC-101 1.3E-01 CE-141 3.2E-04 NI-65 1.4E-03 55 w TOTAL 4.4E-00 TE-129M 2.7E-03 CE-143 2.3E-04 ZN-65 1.9E-05 -H TE-132 1.0E-01 CE-144 2.9E-04 ZN-69M 2.lE-04 h CS-134 1.3E-03 PR-143 3.0E-04 AG-110M 6.0E-04 CS-136 8.0E-04 ND-147 1.lE-04 W-187 2.2E-02 o CS-137 2.0E-03 CS-138 3.7E-01 TOTAL 7.9E-02 TOTAL 3.3E-01 BA-137M 2.0E-03 BA-139 6.2E-01 BA-140 6.6E-02 BA-141 2.0E-01 BA-142 1.1E-01 NP-239 1.6E-00 TOTAL 5.2E-00 $?3 oS

i t ] Table 12.2-13 (Continued) l 4 HIGH CONDUCTIVITY OIL SEPARATOR D015 l Source Volume = 6300 gal. Total Curies 1.5 = j Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies t BR-83 7.lE-03 SR-89 1.2E-02 ZR-95 1.5E-04 NA-24 1.6E-03 BR-84 2.7E-03 SR-90 9.8E-04 ZR-97 2.5E-05 P-32 7.0E-05 BR-85 1.5E-04 SR-91 4.3E-02 NB-95 4.7E-04 CR-51 2.0E-03 w l I-131 3.5E-01 SR-92 3.0E-02 RU-103 7.lE-05 MN-54 1.9E-04 m l I-132 8.0E-02 Y-90 9.8E-04 RU-106 1.lE-05 MN-56 1.2E-02 [ I-133 2.3E-01 Y-91M 2.9E-02 RH-103M 7.lE-05 CO-58 2.2E-02 1-134 4.0E-02 MO-99 3.3E-02 RH-106 1.lE-05 CO-60 2.4E-03 t* m I-135 1.5E-01 TC-99M 3.8E-02 LA-140 3.0E-02 FE-59 3.5E-04 l w TC-101 4.0E-03 CE-141 1.4E-04 NI-65 7.2E-05 TOTAL 8.6E-01 TE-129M 1.2E-03 CE-143 3.5E-05 ZN-65 9.6E-06 -H TE-132 2.3E-02 CE-144 1.4E-04 ZN-69M 2.3E-05 3" CS-134 6.8E-04 PR-143 1.10-04 TS-110M 2.9E-04 h CS-136 3.lE-04 ND-147 4.0E-05 E-187 2.9E-03 O l CS-137 1.0E-03 CS-138 1.2E-02 TOTAL 3.lE-02 "OTAL 4.4E-07 l BA-13;." 1.0E-03 BA-139 2.3E-02 BA-140 2.6E-02 BA-141 6.2E-03 l BA-142 3.5E-03 NP-239 3.2E-01 ( TOTAL 6.lE-01 l xw I o> i ( .O O O4 l i i G G G

i O O O Table 12.2-13 (Continued) i LOW CONDUCTIVITY COLLECTOR TANK A186 Source Volume = 57,000 gal. Tobal Curies = 30 Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies 4 l BR-83 2.6E-02 SR-89 7.6E-01 ZR-95 1.0E-02 NA-24 2.0E-02 BR-84 3.4E-03 SR-90 6.6E-02 ZR-97 3.2E-04 P-32 4.4E-03 I BR-85 1.7E-05 SR-91 4.3E-01 NB-95 3.lE-02 CR-51 1.3E-01 N I-131 4.6E-00 SR-92 1.6E-01 RU-103 4.6E-03 MN-54 1.3E-02 I-132 1.2E-00 Y-90 6.6E-02 RU-106 7.2E-04 MN-56 6.6E-02 z I-133 2.2E-00 Y-91M 2.9E-01 RH-103M 4.6E-03 CO-58 1.5E-00 l g N I-134 7.8E-02 MO-99 1.3E-00 RH-106 7.2E-04 CO-60 1.6E-01 em i w I-135 9.2E-01 TC-99M 3.0E-01 LA-140 1.8E-00 FE-59 2.3E-02 ll 0 TC-101 3.4E-03 CE-141 9.lE-03 NI-65 3.9E-04 55 l 4 TOTAL 9.0E-00 TE-129M 7.4E-02 CE-143 7.8E-04 ZN-65 6.4E-04 wH t TE-132 1.0E-00 CE-144 9.6E-03 ZN-69M 2.7E-04 h i CS-134 4.5E-02 PR-143 7.lE-03 AG-110M 2.0E-02 l CS-136 1.9E-02 ND-147 2.4E-03 W-187 4.9E-02 o i CS-137 7.0E-02 t CS-138 2.2E-02 TOTAL 1.9E-00 TOTAL 2.0E-00 j BA-137M 7.0E-02 j BA-139 9.3E-02 l I BA-140 1.6E-00 BA-141 6.6E-03 BA-142 2.2E-03 l NP-239 1.lE-01 TOTAL 1.7E-01 w ,i 8 ou I i i i

Table 12.2-13 (Continued) HIGH CONDUCTIVITY COLLECTOR TANKS A441 Source Volume = 18,000 gal. Total Curies 15 = Soluble Fission Insoluble Fission Activation Halogens Products Products Products I_s_o to pe Curies Isotope Curies Isotope Curies Isotope Curies BR-83 4.8E-03 SR-89 2.6E-02 ZR-95 3.3F-04 NA-24 1.2E-03 BR-84 1.6E-04 SR-90 2.3E-03 ZR-97 2.0E-05 P-32 1.5E-04 BR-85 4.3E-07 SR-91 2.2E-02 NB-95 1.0E-03 CR-51 4.2E-03 N I-131 1.0E-01 S R-9 2 4.3E-03 RU-103 1.5E-04 MN-54 4.lE-04 I-132 7.4E-02 Y-90 2.3E-03 RU-106 2.3E-05 MN-56 1.7E-03 n H I-133 3.3E-00 Y-91M 1.5E-02 RH-103M 1.5E-04 CO-58 4.7E-02 co I-134 5.3E-03 MO-99 5.4E-02 RH-106 2.3E-05 CO-60 5.2E-03 0$ 7 I-135 4.9E-01 TC-99M 1.2E-02 LA-140 6.3E-02 FE-59 7.4E-04 s$ g TC-101 5.2E-05 CE-141 2.9E-04 NI-65 9.9E-06 %x TOTAL 1.4E-01 TE-129M 2.4E-03 CE-143 4.4E-05 ZN-65 2.2E-05 TE-132 3.8E-02 CE-144 3.lE-04 ZN-69M 1.6E-05 CS-134 1.6E-03 PR-143 2.3E-04 AG-110M 6.2E-04 y CS-136 6.6E-04 ND-147 8.0E-05 W-187 3.0E-03 o CS-137 2.5E-03 CS-138 3.3E-04 TOTAL 6.6E-02 TOTAL 6.4E-02 BA-137M 2.5E-03 BA-139 1.8E-03 BA-140 5.5E-02 BA-141 1.0E-04 BA-142 3.4E-05 NP-239 5.0E-01 TOTAL 7.4E-01 xU S5

• O O

OM O O O

i 1 GESSAR II 22A7007 1 238 NUCLEAR ISLAND Rev. O l i 4 Table 12.2-13 (Continued) I i. FILTRATE TANK A287 i ti } 1 l j Source Volume = 3000 gal. Ci Unit = l l 4 Tiet - 3.504E-05 3.504E-04 1.504E-0$ 3.5J4E-05 TOTAL 0. TOTAL 0. TOTAL 0. TOTAL 1.842L-c9 IT 6.498E-01 90S8 1.423E-05 12958 6.059E-06 143ra 7.e24E-06 l iNJ 9.629E-12 90YM 1.663E-10 1297 EM S.963E-07 1,44E 0. l' 13 AM

• 4.564L-12 90V

1. 2 9 3E-05 129TE 1.282L-05 144C5 0.

j 159)*

• 1.779E-10 916n O.

1315M 2.685t*07 144hA 0. 16NJ 0. 91s# 0. 131 TEM 2.9saE*06 14 4 LA 0. i 16AM 0. 91RB 0. 131TE 1.524L-06 144CL J.395L-06 16 w) O. 9159

4. 0 2 sE-0 )

131I 2.452E-03 147hc 2.0 76 L-u6 !?%2 0. 91YM 1.910E-03 1}1xEM %.565E-04 14 7 t'M 5.14at-07 i 17A4 0. 91Y 7.S47F-05 132TE 1.es6E-0) 149ND 9.3191-06 1740 0. 9289 0. 1321 E.549E-0) 149PM 4.848E-06 I 16r 9.404k-0% 92EM 0. 13358 2.2e1E-34 187W

3. 0 41 L - 0 4 O.

92RD 0, 11)TFM 2.076E-04 239%F J. 2 5 ",L - U J f 19'.4 24 1.462E-04 9J5n 4.960E-0) 13 3TE S.769E-07 4 i 12F

2. 217L-06 92Y 8.004E-Gl 1331M 0.

51ca 5.6 76E-05 timp 0. 1311 1.573E-02 i 544N 4.463E-06 91R B 0. 133 TEM 1.505f-02 56MN 1.077E-0) 915p 2.665[-12 l i lX E 2.279E-01 k s eco 5.6666-04 9 3Y 4.625E-04 134TE 1.395E-05 59# E 9.050E-06 44FB 0. 1341M 4.271E-J4 4 0' ') 5.414L-05 94MB 0. 1341 2.096E-01 t 6%%1 1.09tt-Oi 94S8 0. 134RLM S.534E-J6 I i 6134 2.J31L-07 94Y 4.43CE-07 1353 1.809E-02 l 1 692' M 2.7516-06 95>R 0. 135xLM 6.1670+04 k I S l A5 D. 958n O. 135XE 1.383E-00 l l 41r14 U. 9 5SH 0. 13 5C' M 8.974E-08 41s t. 2.130 L-o s 95v 2.789E-10 136TE D. j $3pm 9.748[-04 953R 4.7375-06 1361M 0. E lp KM

5. 792E-02 9 5 N hM 1.120E-07 1368 0.

44A5 0. 95h8 4.221E-06 1371 0. g 84SE 2.410L-26 9728 2.632E+05 13 7x E 8.421E-21 84kkM 4.022E-17 97hbM 1.23aE-0% 1 ) ?C S 1.439E-05 l 2 34MN 3.122E-04 97Ns 1.292E-Ci 137 LAM 6.402E-v4 p 95AS 0 992R 0 138tE B.071E-00 eiblM 0. 9 9 m kM ).809E-33 138C5M 1.789E-29 8551 9. 9949 0. 138CS 4.496E-05 I N5Dn 7.579E-J9 9990 3.276E-0% 134xE 0. 9%NPM 4.SJ9L-03 99?CM 1.035L-02 139C5 1.0u4E-10 0 5F.m 1.094E-06 997C 0. 139EA 1.546E-03 i i 87AS 0. 10140 1.1951-08 140xE 0. 8 73 [ 0. 101Tt' 1.312L-07 140C8 0. I' 8/PH 0. 102Mo 3.0420-10 140aA S.J05E-04 6?FN

1. 7 06L-C A 102TCM 1.3J6E-10 14 G LA 9,8 5 3E.-0 4 Stit o.

102TC 0. 141AL 0. 4en9 0. 10)TC 0. 141CS 0. BHM 1.t'UE-03 10 3PU 2.173E-06 141RA 1.85%E-06 98RH 2.686E-04 10 3R nM 1.757E-05 141LA 1.7140-05 a9sh D. 104%o 0. 141CE 2.216E-05 8984 0. 104TC 5.137E-07 142KE 0 39F#

1. 5 3 5 E - 2 %

105%o 0. 14JCS 0. eius

1. 5 35E- 0 7 105TC 1.861E-12 142hA 2.317E-09 s

095u 2.54IL-04 105PU

1. 2 9",E-0 4 142 LA 5.317L-04 j

811M 0. 10 5 R t1H 1.756E*05 143xE 0 90PF 0. 105MH 3.605E-05 141CS 0. 90pd G. 106TC 0. 143EA 0. 70MbH 9.8%2E-J1 106RU 1.%64E-07 143LA 3.0550-08 90#8 2,8 32E-2 2 110AGM 6.716E-06 143CE 4.502E-05 .AM iu r.t.. i.et y. a a e ry.o...-.. w 3 .. o imlic... i.e.er. or r.iiro9.... 10... V f 1 i t I tilllli I l 12!. ;!-- 3 9 [ -. + -

Table 12.2-13 (Continued) DISTILLATE TANK A776 Source volume = 3000 gal. Total Curies = 0.00024 Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 6.9-E-08 SR-89 4.4E-07 ZR-95 5.5E-09 NA-24 2.0E-08 BR-84 1.4E-09 SR-90 3.9E-08 ZR-97 3.3E-10 P-32 2.5E-09 BR-85 0. SR-91 3.6E-07 NB-95 1.7E-08 CR-51 7.0E-08 I-131 1.7E-04 SR-92 6.3E-08 RU-103 2.5E-09 MN-54 6.9E-09 I-132 1.2E-06 Y-90 3.9E-08 RU-106 3.9E-10 MN-56 2.4E-08 = e I-133 5.4E-05 Y-91M 2.4E-07 RH-103M 2.5E-09 CO-58 7.9E-07 EO I-134 5.9E-08 MO-99 9.0E-07 RH-106 3.9E-10 CO-60 8.7E-08 Q@ N I-135 7.8E-06 TC-99M 2.0E-07 LA-140 1.lE-06 FE-59 1.2E-08 1 TC-101

2. 6E-10 CE-141 4.9E-09 NI-65 1.4E-10 TOTAL 2.3E-04 TE-129M 4.0E-J8 CE-143 7.2E-10 ZN-65 3.8E-10

(( TE-132 6.4E-07 CE-144 5.2E-09 ZN-69M 2.7E-10 g CS-134 2.7E-08 ND-147 1.3E-09 W-187 5.0E-08 z CS-136 1.1E-08 0 CS-137 4.lE-08 TOTAL 1.lE-06 TOTAL 1.lE-06 CS-138 3.0E-09 BA-137M 4.lE-08 BA-139 2.3E-08 BA-140 9.2E-07 BA-141 6/3E-10 BA-142 1.3E-10 NP-239 8.3E-06 TOTAL 1.2E-05 g NY <.a 05 O O O

O O b/ i V V Table 12.2-13 (Continued) PRECOAT TANK A275 Source Volume = 2000 gal. Total Curies = 0.011 Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 7.6E-06 SR-89 3.0E-04 ZR-95 3.9E-06 NA-24 7.3E-06 BR-84 3.7E-07 SR-90 2.6E-05 ZR-97 1.2E-07 P-32 1.7E-06 BR-85 0. SR-91 1.6E-04 NB-95 1.2E-06 CR-51 5.lE-05 S$ I-131 1.8E-03 SR-92 5.0E-05 RU-103 1.8E-06 MN-54 5.0E-06 I-132 4.lE-04 Y-90 2.6E-05 RU-106 2.8E-07 MN-56 2.0E-05 = H I-133 8.3E-04 Y-91M 1.lE-04 RH-103M 1.8E-06 CO-58 5.8E-04 I-134 1.4E-05 MO-99 5.1E-06 RH-106 2.8E-07 CO-60 6.4E-05 yy I-135 3.2E-04 .TC-99M 1.lE-04 LA-140 7.22-04 FE-59 9.lE-06 ^ TC-101 7.6E-08 CE-141 3.0E-07 NI-65 1.2E-ll TOTAL 3.4E-03 TE-129M 2.9E-05 CE-143 3.0E-07 ZN-65 2.5E-07 (( TE-132 3.9E-04 CE-144 3.8E-06 ZN-69M 9.8E-08 g CS-134 1.7E-05 PR-143 2.8E-06 AG-110M 7.6E-06 z CS-136 7.6E-06 ND-147 9.lE-07 W-187 1.9E-05 D CS-137 2.7E-05 CS-138 2.4E-06 TOTAL 7.4E-04 TOTAL 7.6E-04 BA-137M 2.7E-05 BA-139 2.3E-05 BA-140 6.3E-04 BA-141 2.9E-07 BA-142 2.0E-08 NP-239 4.5E-03 TOTAL 6.4E-03 g <a O OM I l

Table 12.2-13 (Continued) DETERGENT WASTE TANKS A838 Source Volume = 1500 gal. Total Curies 0.000013 = Soluble Fission Insoluble Fission Activation Halogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 3.8E-08 SR-89 4.4E-08 ZR-95 6.0E-10 NA-24 1.7E-08 BR-84 1.7E-09 SR-90 3.4E-09 ZR-97 2.7E-10 P-32 3.2E-10 BR-85 0. SR-91 4.0E-07 NB-95 1.9E-09 CR-51 8.6E-09 N I-131 1.lE-06 SR-92 1.2E-07 RU-103 2.9E-10 MN-54 7.0E-10 I-132 4.3E-07 Y-90 3.4E-09 RU-106 0. MN-56 5.0E-08 = I-133 4.2E-06 Y-91M 2.7E-07 RH-103M 2.9E-10 CO-58 8.6E-08 gQ g N I-134 4.2E-08 MO-99 2.8E-07 RH-106 0. CO-60 8.7E-09 em w I-135 2.lE-06 TC-99M 2.8E-07 LA-140 1.4E-07 FE-59 1.4E-09 ss i TC-101 1.7E-09 CE-141 5.9E-10 NI-65 3.0E-10 mW N TOTAL 7.9E-06 TE-129M 5.0E-09 CE-343 3.9E-10 ZN-65 0. ss TE-132 1.9E-07 CE-144 5.3E-10 ZN-69M 2.4E-10 k CS-134 2.3E-09 PR-143 5.6E-10 AG-110M 1.lE-09 CS-136 1.5E-09 ND-147 2.0E-10 W-187 3.3E-08 O CS-137 3.6E-09 CS-138 1.lE-09 TOTAL 1.5E-07 TOTAL 2.lE-07 BA-137M 3.6E-09 BA-139 5.6E-08 BA-140 1.2E-07 BA-141 3.3E-09 BA-142 1.lE-09 NP-239 2.9E-06 TOTAL 4.7E-06 xU e5 8 04 O O O

GESSAR II 22A7007 7'd NUCLEAR ISLAND Rev. 0 s Table 12.2-14 OFFGAS SYSTEM COMPONENT INVENTORY ACTIVITIES Calculated Offgas System component inventory activities are provided for four different sets of 'Cfgas system parameters. The inventory activities are provided as a basis of evaluating the site boundary dose rate for postulated accidents as analyzed in Section 15.7. The system parameters for the cases are shown below. I (3 V 12.2-43

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE IDENTIFICATION 30-Minute-Old Noble Gas Mix Case No. Flow (pCi/sec) Buildup Time 1 30 100,000 1 year 2 30 100,000 11 months 350,000 30 days 3 2 100,000 1 year 4 2 100,000 11 months 350,000 30 days Case 1 (30 scfm flow, 100,000 pCi/sec mix, 1-year buildup time) - This case represents the normal design offgas system operating conditions. Case 2 (30 scfm flow, 100,000 pCi/sec mix for 11 months plus 350,000 pCi/sec mix for 30 days; a total buildup time of 1 year). This case represents normal conditions for 11 months followed by a 30-day abnormal condition where a 250,000 pCi/sec spike is added to the normal 100,000 pCi/sec release. Case 3 (2 scfm flow, 100,000 pCi/sce mix, 1-year buildup time). This case represents a lower limit operating condition where the main condenser air inleakage is reduced. This causes a buildup of radioactive inventory in the " front end" of the system. (Note: system design mini-mum air flow can be maintained 6 scfm by an auxiliary air bleed.) Case 4 (2 scfm flow, 100,000 pCi/sec mix for 11 months plus 350,000 pCi/sec mix for 30 days; total buildup time 1 year). This case represents abnormal operating condi-tions of low flow and an additional activity release l spike of 250,000 pCi/sec for 30 days. I 12.2-44

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O '~'N Table 12.2-14 (Continued) (d CASE ONE OPPGAS PREHEAT (Ci) 13N2 1.lllE-03 90KR 1.774E-01 138XE 4.008E-02 16N2 2.452E-00 91KR 1.535E-01 139XE 1.824E-01 16AM* 1.074E-02 92KR 6.591E-03 140XE 1.709E-01 83KRM 1.339E-03 133XE 2.516E-03 141XE 4.102E-03 85KRM 1.994E-03 135XEM 1.193E-02 87KR 6.883E-03 135XE 7.480E-03 88KR 6.713E-03 137XE 8.288E-02 TOTAL 3.395E-00 = 89KR 7.087E-02 RECOMBINER (Ci) 13N2 3.314E-03 90RB 2.415E-03 138XE 1.197E-01 16N2 6.640E-00 91KR 4.238E-01 139XE 5.351E-01 16AM 2.909E-02 91RB 6.220E-03 140XE 4.856E-01 171N2 1.021E-03 92KR 1.389E-02 140CS 6.463E-03 83KRM 4.001E-03 92RB 2.620E-03 141XE 8.462E-03 85KRM 5.958E-03 133XE 7.518E-03 (N, 87KR 2.056E-02 135XEM 3.562E-02 (j 88KR 2.006E-02 135XE 2.235E-02 85KR 2.110E-01 137XE 2.469E-01 TOTAL 9.373E-00 = 90KR 5.183E-01 OFFGAS CONDENSER (Ci) 13N2 9.194E-02 91KR 5.248E-00 139XE 1.217E+01 16N2 5.931E-01 91RB 1.511E-00 139CS 3.994E-01 16AM 3.036E-01 92KR 4.101E-02 139BA 1.406E-03 17N2 6.715E-03 92RB 5.880E-02 140XE 7.663E-00 190 1.384E-02 93KR 1.887E-03 140CS 2.001E-00 83KRM 1.125E-01 93RB 3.351E-03 141XE 2.335E-02 85KRM 1.676E-01 133XEM 8.682E-03 141CS 1.900E-02 87KR 5.778E-01 133XE 2.ll6E-01 142XE 1.066E-03 88KR 5.641E-01 134I 1.108E-03 142CS 1.707E-03 88RB 9.690E-03 135XEM 9.932E-01 89KR 5.672E-00 135XE 6.296E-01 89RB 1.135E-01 137XE 6.692E-00 90KR 1.126E-01 138XE 3.334E-00 90RBM 8.690E-02 138CS 3.161E-02 TOTAL 1.304E-02 = 90RB 1.093E-00 /'3

• AM indicates isotopes of nitrogen as ammonia, NH3*

12.2-45

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE ONE WATER SEP (Ci) 13N2 1.276E-02 90RBM 1.215E-03 138XE 4.622E-01 16N2 1.887E-00 90RB 1.555E-02 139XE 1.317E-00 83KRM 1.580E-02 91KR 2.087E-01 139CS 5.849E-03 85KRM 2.357E-02 91RB 8.681E-03 140XE 4.813E-01 87KR 8.110E-02 133XEM 1.218E-03 140CS 1.838E-02 88KR 7.929E-02 133XE 2.968E-02 89KR 7.566E-01 135XEM 1.378E-01 89RB 2.059E-03 135XE 8.830E-02 90KR 1.147E-00 137XE 8.986E-01 TOTAL = 7.681E-00 OFFGAS DELAY LINE (Ci) 13N2 1.246E-00 90RB 4.715E-00 138CS 2.921E-00 16N2 4.510E-00 91KR 6.563E-01 139XE 2.037E+01 83KRM 1.937E-00 91RB 3.409E-01 139CS 3.714E-00 85KRM 2.930E-00 91SR 1.272E-03 139BA 9.004E-02 85KR 7.510E-03 131XEM 1.551E-02 140XE 2.401E-00 87KR 9.838E-00 133XEM 1.528E-01 140CS 1.137E-00 88KR 9.799E-00 133XE 3.725E-00 88RB 1.006E-00 135XEM 1.465E-01 89KR 4.628E+01 135XE 1.ll7E-01 89RB 5.450E-00 137XE 6.147E-01 90KR 1.435E+01 138XE 4.851+-l TOTAL = 2.739E-02 90RBM 4.791E-01 OFFGAS COOLER (Ci) 13N2 1.726E-02 90RBM 4.439E-03 138CS 5.127E-02 83KRM 3.457E-02 90RB 2.783E-02 139XE 1.054E-03 85KRM 5.301E-02 133XEM 2.789E-03 139CS 5.041E-02 87KR 1.736E-01 133XE 6.806E-02 139BA 1.651E-03 88KR 1.763E-01 135XEM 2.24.E-01 88RB 1.671E-02 135XE 2.032E-01 89KR 3.278E-01 137XE 5.231E-01 89RB 8.532E-02 138XE 7.317E-01 TOTAL = 2.776E-00 O 12.2-46

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 [~) Table 12.2-14 (Continued) G' CASE ONE MOISTURE SEP (Ci) 13N2 1.026E-02 89KR 1.917E-01 137XE 3.072E-01 83KRM 2.070E-02 133XEM 1.671E-03 138XE 4.360E-01 85KRM 3.175E-02 133XE 4.077E-02 87KR 1.039E-01 135XEM 1.33bE-01 88KR 1.056E-01 135XE 1.217E-01 Total = 1.508E-00 OFFGAS PREFILTER (Ci) 13N2 5.566E-02 133XEM 9.234E-03 138XE 2.378E-00 83KRM 1.142E-01 133XE 2.253E-01 139XE 2.370E-03 85KRM 1.754E-01 135XEM 7.293E-01 87KR 5.728E-01 135XE 6.729E-01 88KR 5.827E-01 137XE 1.618E-00 Total = 8.138E-00 89KR 9.933E-01 FILTER 1 YR (Ci) (-~) 88RB 6.866E-01 133I 5.532E-03 139CS 2.924E-02 89RB 1.188E-00 135I 2.777E-03 139BA 2.922E-03 89SR 1.178E-00 137CS

4. 374E-32 131I 7.024E-03 137BAM 2.060E-02 132I 1.019E-03 138CS 2.808E-00 Total =

5.949E-00 N 12.2-47 3

GESSAR II 22A7007 238 NUCLEAR ISLRND Rev. O Table 12.2-14 (Continued) O CASE ONE OFFGAS DRYER (Ci) 13N2 2.440E-01 89KR 3.626E-00 137XE 6.162E-00 83KRM 5.436E-01 89RB 3.431E-01 138XE 1.071E-01 85KRM 8.387E-01 131XEM 4.504E-03 138CS 4.908E-01 85KR 2.181E-03 133XEM 4.430E-02 139XE 3.496E-03 87KR 2.716E-00 133XE 1.081E-00 88KR 2.781E-00 135XEM 3.299E-00 88RB 2.275E-01 135XE 3.233E-00 Total = 3.635E-01 CHARCOAL DELAY (Ci) 13N2 1.492E-00 89SR 9.550E-00 137BAM 1.829E-00 83KRM 3.992E+01 131XEM 4.698E+01 138XE 9.545E+01 85KRM 1.488E+02 133XEM 9.419E+01 1380S 1.057E+02 85KR 2.778E-00 133XE 5.525E+03 139CS 3.374E-03 87KR 1.354E+02 135XEM 3.180E+01 139BA 3.895E-03 88KR 3.081E+02 135XE 1.208E+03 88RB 3.107E+02 135CS 1.823E-03 89KR 5.918E-00 137XE 1.279E+01 89RB 9.204E-00 137CS 2.059E-00 Total = 8.095E+03 AFTER-FILTER (Ci) 133XE 1.022E-03 134TE 0. FILTER 1 YR (Ci) 137CS 3.360E-01 137BAM 1.582E-01 Total = 4.946E-01 O 12.2-48

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4 () Table 12.2-14 (Continued) 4 CASE TWO I OFFGAS PREHEAT (Ci) j i 16N2 2.452E-00 91RB 1.780E-03 138XE 1.202E-01 i 16AM* 1.074E-02 92KR 1.977E-02 139XE 5.471E-01 i 83KRM 4.014E-03 92RB 1.098E-03 140XE 5.127E-01 85KRM 5.978E-03 93KR 1.506E-03 141XE 1.230E-02 87KR 2.064E-02 133XE 7.429E-03 i 88KR 2.013E-02 135XEM 3.574E-02 ) 89KR 2.126E-01 135Xe 2.240E-02 90KR 5.320E-01 137XE 2.486E-01 Total = 5.256E-00 91KR 4.602E-01 l RECOMBINER (Ci) i 13N2 3.314E-03 91RB 1.886E-02 140XE 1.454E-00 16N2 6.639E-00 92KR 4.168E-02 140CS 1.939E-02 1 17N2

1. 0 ? ') E-0 3 92RB 7.860E-03 141XE 2.538E-02 83KRM 1.19, -02 93KR 2.778E-03 141CS 1.071E-03 i

85KRM 1.786E-02 133XE 2.220E-02 142XE 1,634E-03 87KR 6.168E-02 135XEM 1.067E-01 1 1 88KR 6.015E-02 135XE 6.693E-02 l C9KR 6.328E-01 137XE 7.407E-01 i 90KR 1.555E-00 138XE 3.589E-01 i 90RB 7.245E-03 139XE 1.605E-00 91KR 1.271E-00 139CS 2.444E-03 Total = 1.477E-01 l i i i i O l 12.2-49

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) g CASE TWO OFFGAS CONDENSER (Ci) 13N2 9.193E-02 91SR 2.347E-03 139XE 3.652E+01 16N2 6.931E+01 92KR 1.230E-01 139CS 1.198E-00 16AM* 3.036E-01 92RB 1.764E-01 139BA 4.217E-03 17N2 6.718E-03 93KR 5.660E-03 140XE 2.299E+01 190 1.379E-02 93RB 1.005E-02 140CS 6.003E-00 83KRM 3.372E-01 131XEM 2.492E-03 141XE 7.004E-02 85KRM 5.026E-01 132I 1.794E-03 141CS 5.699E-02 87KR 1.733E-00 133I 1.044E-03 142XE 3.197E-03 88KR 1.692E-00 133XEM 2.583E-02 142CS 5.122E-03 88RB 2.906E-02 133XE 6.249E-01 89KR 1.701E+01 134I 3.322E-03 89RB 3.406E-01 135I 1.674E-03 90KR 3.377E+01 135XEM 2.976E+00 90RBM 2.607E-01 135XE 1.885E+00 90RB 3.280E-00 137XE 2.008E+01 91KR 1.475E+01 138XE 1.000E+01 Total = 2.518E+02 91RB 4.534E-00 WATER SEP (Ci) 13N2 1.276E-02 90RB 4.664E-02 138CS 1.782E-03 16N2 1.887E-00 91KR 6.261E-01 139XE 3.950E-00 83KRM 4.737E-02 91RB 2.604E-02 139CS 1.755E-02 85KRM 7.066E-02 133XEM 3.623E-03 140XE 1.444E-00 87KR 2.433E-01 133XE 8.766E-02 140CS 5.514E-02 88KR 2.378E-01 135XEM 4.129E-01 89KR 2.270E-00 135XE 2.644E-01 89RD 6.176E-03 137XE 2.696E-00 l 90KR 3.442E-00 138XE 1.386E-00 Total = 5.514E-02 90RBM 3.646E-03 O 12.2-50

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 12.2-14 (Continued) CASE TWO OFFGAS DELAY LINE (Ci) 13N2 1.245E-00 90RB 1.414E+01 138CS 8.763E-00 16N2 4.510E-00 91KR 1.969E-00 139XE 6.112E+01 83KRM 5.808E-00 91RB 1.023E-00 139CS 1.ll4E+01 85KRM 8.784E-00 91SR 3.816E-03 139BA 2.701E-01 85KR 4.268E-03 131XEM 4.387E-02 140XE 7.201E-00 87KR 2.951E+01 133XEM 4.544E-01 140CS 3.401E-00 88KR 2.939E+01 133XE 1.100E+01 88RB 3.016E+00 135XEM 4.388E+01 89KR 1.388E-02 135XE 3.345E+01 89RB 1.635E+01 137XE 1.844E+02 90KR 4.304E+01 138XE 1.455E+02 Total = 8.097E+02 90RBM 1.437E-00 OFFGAS COOLER (Ci) 13N2 1.725E-02 90RBM 1.332E-02 138XE 2.195E-00 83KRM 1.037E-01 90RB 8.350E-02 138CS 1.538E-01 w 85KRM 1.590E-01 133XEM 8.298E-03 139XE 3.162E-03 1 87KR 5.207E-01 133XE 2.010E-01 139CS 1.512E-01 88KR 5.286E-01 135XEM 6.713E-01 139BA 4.953E-03 88RB 5.012E-02 135XE 6.084E-01 140CS 2.214E-03 89KR 9.832E-01 137XE 1.569E-00 89RB 2.559E-01 Total = 8.286E-00 MOISTURE SEP (Ci) 13N2 1.026E-02 89RB 2.150E-03 138XE 1.308E-00 83KRM 6.206E-02 133XEM 4.971E-03 138CS 2.311E-03 85KRM 9.521E-02 133XE 1.204E-01 139XE 1.688E-03 87KR 3.ll6E-01 135XEM 4.002E-01 88KR 3.166E-01 135XE 3.645E-01 88RB 1.017E-03 137XE 9~.217E-01 Total = 4.498E-00 89KR 5.750E-01 m %d 12.2-51

GESSAR II 22A7007 238 NUCLEAR ISLAWD Rev. O Table 12.2-14 (Continued) CASE TWO OFFGAS PREFILTER (Ci) 13N2 5.565E-02 131XEM 2.654E-03 137XE 4.854E-00 83KRM 3.424E-01 133XEM 2.747E-02 138XE 7.135E-00 85KRM 5.258E-01 133XE 6.654E-01 139XE 7.110E-03 87KR 1.718E-00 135XEM 2.185E-00 88KR 1.748E-00 135XE 2.015E-00 Total = 2.428E+01 89KR 2.998E-00 FILTER 1 YR (Ci) 88RB 2.059E-00 1331 1.656E-02 138CS 8.423E-00 89RB 3.563E-00 134I 2.0llE-03 139CS 8.771E-03 89SR 3.535E-00 135I 8.324E-03 139BA 8.764E-03 90RB 1.023E-03 137CS 1.312E-01 131I 2.074E-02 137BAM 4.669E-02 Total = 1.783E+01 132I 3.015E-03 13N2 2.439E-01 138CS 1.472E-00 133XEM 1.318E-01 83KRM 1.630E-00 89KR 1.088E-01 133XE 3.193E-00 85KRM 2.515E-00 89RB 1.029E-00 135XEM 9.883-00 85KR 1.240E-03 90KR 1.086E-03 138XE 3.212E-01 87KR 8.147E-00 131XEM 1.274E-02 139CS 1.574E-03 88KR 8.341E-00 137XE 1.848E-01 88RB 6.824E-01 139XE 1.049E-02 Total = 1.085E-02 135XE 9.681E-00 OFFGAS DRYER (Ci) 13N2 2.439E-01 131XEM 1.274E-02 Total = 1.085E+02 83KRM 1.630E+00 133XEM 1.318E-01 85KRM 2.515E+00 133XE 3.193+00 l 85KR 1.240E-03 135XEM 9.883E+00 87KR 8.147E+00 135XE 9.681E+00 88KR 8.341E+00 137XE 1.848E+01 88RB 6.824E-01 138XE 3.212E+01 89KR 1.088E+01 138CS 1.472E+00 89RB 1.029E+00 139XE 1.049E-02 90KB 1.086E-03 139CS 1.574E-03 O 1 12.2-52

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 12.2-14 (Continued) CASE TWO CilARCOAL DELAY (Ci) 13N2 1.492E+00 131XEM 1.328E+02 138CS 3.170E+02 83KRM 1.197E-02 133XEM 2.802E+02 139XE 1.192E-03 85KRM 4.461E+02 133XE 1.631E+04 139CS 1.012E-02 85KR 1.606E-00 135XEM 9.528E+01 139BA 1.168E-02 87KR 4.060E+02 135XE 3.619E+03 88KR 9.240E+02 135CS 1.091E-05 88RB 9.313E+02 137XE 3.838E+01 89KR 1.776E+01 137CS 1.295E-00 89RB 2.761E+01 137BAM 6.146E-01 89SR 2.846E+01 138XE 2.863E+02 Total = 2.399E+04 AFTER-FILTER (Ci) 133XE 2.679E-03 134TE O Total = 3.684E-03 o O 12.2-53

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE THREE OFFGAS PREHEAT (Ci) 13N2 1.178E-03 89KR 7.764E-02 135XE 8.588E-03 16N2 2.489E-00 90KR 1.877E-01 137XE 9.386E-02 16AM* 1.489E-02 91KR 1.566E-01 138XE 4.582E-02 83KRM 1.483E-03 92KR 6.670E-03 139XE 1.979E-01 85KRM 2.208E-03 133XE 2.88E-03 141XE 4.151E-03 87KR 7.620E-03 135XEM 1.364E-02 140XE 1.773E-01 88KR 7.434E-03 Total = 3.497E-00 RECOMBINER (Ci) 13N2 3.516E-03 89KR 2.311E-01 135XEM 4.074E-02 16N2 6.731E-00 90KR 5.484E-01 135XE 2.566E-02 16AM 2.942E-02 90RB 2.587E-03 137XE 2.796E-01 17N2 1.032E-03 91KR 4.321E-01 138XE 1.368E-01 83KRM 4.430E-03 91RB 6.423E-03 139XE 5.806E-01 85KRM 6.598E-03 92KR 1.401E-02 140XE 5.024E-01 87KR 2.277E-02 92RB 2.672E-03 140CS 6.783E-03 88KR 2.221E-02 133XE 8.630E-03 141XE 8.528E-03 Total = 9.653E-00 OFFGAS CONDENSER (Ci) 13N2 9.752E-02 90RB 1.167E-00 138XE 3.811E-00 16N2 6.958E+01 91XR 5.305E-00 138CS 3.660E-02 16AM 3.041E-01 91RB 1.546E-00 139XE 1.318E-01 17N2 6.706E-03 92KR 4.082E-02 139CS 4.378E-01 190 1.396E-02 92RB 5.880E-02 139BA 1.560E-03 83KRM 1.2450-01 93KR 1.875E-03 140XE 7.892E-00 85KRM 1.856E-01 93RB 3.354E-03 140CS 2.085E-00 87KR 6.396E-01 131XEM 1.0llE-03 141XE 2.323E-02 88KR 6.247E-01 133XEM 9.967E-03 141CS 1.916E-02 88RB 1.087E-02 133XE 2.429E-01 142XE 1.059E-03 89KR 6.210E-00 134I 1.122E-03 142CS 1.703E-03 89RB 1.259E-01 135XEM 1.136E-00 90KR 1.187E+01 135XE 7.228E-01 90RBM 9.282E-02 137XE 7.575E-00 Total = 1.352E-02 O 12.2-54

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (i) Table 12.2-14 (Continued) w CASE THREE WATER SEPARATOR (Ci) 13N2 1.852E-01 90KR 1.005E-01 135XE 1.324E-00 16N2 5.879E-00 90RBM 1.492E-01 137XE 1.231E-01 83KRM 2.359E-01 90RB 1.838E-00 138XE 6.752E-00 85KRM 3.524E-01 91KR 7.641E-01 138CS 1.288E-01 87KR 1.209E-00 91RB 3.597E-01 139XE 1.240E-01 88KR 1.185E-00 131XEM 1.851E-03 139CS 8.001E-01 88RB 4.080E-02 133XEM 1.823E-02 139BA 5.936E-03 89KR 1.023E-01 133XE 4.445E-01 140XE 2.394E-00 89RB 4.089E-01 135XEM 2.017E-00 14CCS 1.058E-00 Total = 7.255E-01 OFFGAS DELAY LINE (Ci) 13N2 2.884E-00 90RBM 5.871E-01 138XE 1.506E-02 16N2 3.153E-02 90RB 4.512E-00 138CS 5.729E-01 83KRM 2.126E-01 91KR 1.243E-02 139XE 7.987E-00 ~3 85KRM 3.836E-01 91RB 1.668E-01 139CS 7.835E-00 y,) 85KR 1.126E-01 91SR 2.691E-02 139BA 2.096E-00 87KR 9.458E-01 91YM 5.357E-03 140XE 1.659E-01 88KR 1.188E-02 131XEM 2.318E-01 140CS 6.008E-01 88RB 4.739E-01 133XEM 2.261E-00 140BA 2.798E-03 89KR 4.716E-01 133XE 5.553E-01 89RB 2.266E-01 135XEM 4.845E-01 89SR 9.874E-03 135XE 1.613E-02 90KR 4.645E-00 137XE 7.025E-01 Total = 9.678E-02 OFFGAS COOLER (Ci) 83KRM 2.669E-01 88RB 1.769E-01 133XE 1.010E-00 85KRM 6.029E-01 89RB 1.215E-02 135XEM 2.670E-02 85KR 2.057E-03 131XEM 4.226E-03 135XE 2.740E-00 87KR 9.826E-01 133XEM 4.081E-02 138XE 5.995E-02 88KR 1.700E-00 138CS 2.448E-01 139BA 2.296E-02 Total = 7.893E-00 12.2-55

GESSAR II 22A7007 238 RUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE Ti!REE MOISTURE SEPARATOR (Ci) 83KRM 1.583E-01 88RB 4.756E-02 135XEM 1.485E-02 85KRM 3.597E-01 131XEM 2.532E-03 135XE 1.638E-00 85KR 1.232E-03 133XEM 2.444E-02 138XE 3.314E-02 87KR 5.799E-01 133XE 6.047E-01 88KR 1.0llE-00 Total = 4.47E-00 OFFGAS PREFILTER (Ci) 83KRM 8.533E-01 88RB 3.020E-03 135XEM 6.867E-02 85KRM 1.967E-00 131XEM 1.399E-02 135XE 9.007E-00 85KR 6.811E-03 133XEM 1.350E-01 138XE 1.52E-01 87KR 3.090E-00 133XE 3.341E-00 88KR 5.498E-00 Total = 2.414E-01 FILTER 1 YR (Ci) 88RB 6.449E-00 138CS 1.832E-01 Total = 6.633E-00 0 OFFGAS DRYER (Ci) 83KRM 3.635E-00 88KR 2.438E+01 133XE 1.604E-01 85KRM 8.991E-00 88RB 1.761E+01 135XEM 1.471E-01 85KR 3.275E-02 131XEM 6.720E-02 135XE 4.225E+01 87KR 1.245E+01 133XEM 6.461E-01 138CS 1.549E-01 138XE 3.064E-01 Total = 1.267E+02 0 12.2-56

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 12.2-14 (Continued) CASE TilREE CilARCOAL DELAY (Ci) 14C 6.786E-03 88RB 1.758E+02 135XE 1.00E-03 83KRM 1.616E+01 131XEM 5.120E+01 135CS 1.263E-02 85KRM 1.023E+02 133XEM 9.128E+01 138XE 7.800E-02 85KR 4.157E+01 133XE 5.475E+03 138CS 2.298E-01 87KR 3.587E+01 135XEM 4.334E+02 88KR 1.690E+02 Total = 7.164E+03 AFTER FILTER (Ci) 85KR 7.550E-03 Total = 7.562E-03 FILTER 1 YR (Ci) Total = 1.731E-08 O i I ' O 12.2-57

GESSAR Il 22A7007 238 NUCLEAR 1SLAND Rev. O Table 12.2-14 (Centinued) CASE FOUR OFFGAS PREHEAT (Ci) 13N2 1.178E-03 91KR 4.698E-01 137XE 2.816E-01 16N2 2.489E-00 91RB 1.838S-03 138XE 1.375E-01 16AM* 1.088E-02 92KR 2.001En02 139XE 5.938E-01 83KRM 4.445E-03 92RB 1.123E-03 140XE 5.319E-01 85KRM 6.621E-03 93KR 1.524E-03 140CS 1.892E-03 87KR 2.286E-02 133XE 8.529E-03 141XE 1.245E-02 88KR 2.229E-02 135XEM 4.088E-02 89KR 2.329E-01 135XE 2.571E-02 Total = 5.486E-00 90KR 5.632E-01 RECOMBINER (Ci) 13N2 3.516E-03 90RB 7.761E-03 137XE 8.387E-01 16N2 6.731E-00 91KR 1.296E-00 138XE 4.104E-01 16AM 2.942E-02 91RB 1.927E-02 139XE 1.742E-00 17N2 1.033E-03 92KR 4.201E-02 139CS 2.685E-03 83KRM 1.328E-02 92RB 8.016E-03 139CS 2.685E-03 85KRM l.979E-02 93KR 2.796E-03 140XE 1.507E-00 87KR 6.829E-02 133XEM-1.053E-03 140CS 2.035E-02 88KR 6.662E-02 133XE 2.349E-02 141XE 2.558E-02 89KR 6.933E-01 135XEM 1.221E-01 141CS 1.093E-03 90KR 1.645E-00 135XE 7.684E-02 142XE 1.644E-03 Total = 1.543E+01 O J 12.2-58

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f) Table 12.2-14 (Continued) v CASE FOUR OFFGAS CONDENSER (Ci) / 13N2 9.751E-02 92KR 1.224E-01 139CS 1.313E-00 / 16N2 6.958E+01 92RB 1.764E-01 139BA 4.681E-03 16AM* 3.041E-01 93KR 5.625E-03 140XE 2.368E+01 17N2 6.710E-03 93RB 1.006E-02 140CS 6.254E-00 190 1.391E-02 131XEM 2.860E-03 141XE 6.970E-02 83FRM 3.734E-01 132I 1.817E-03 141CS 5.748E-02 05KRM 5.566E-01 133I 1.058E-03 142XE 3.176E-03 87KR 1.919E-00 133XEM 2.965E-02 142CS 5.108E-03 88KR 1.874E-00 133XE 7.174E-01 88RB 3.259E-02 134I 3.365E-03 Total = 2.655E+02 89KR 1.863E+01 1351 1.695E-03 89RB 3.776E-01 135XEM 3.403E-00 90KR 3.562E-01 135XE 2.164E-00 90RBM 2.784E-01 137XE 2.272E+01 90RB 3.502E-00 138XE 1.143E+01 91KR 1.592E+01 138CS 1.098E-01 91RB 4.638E-00 139XE 3.953E+01 91SR 2.431E-03 '(_j WATER SEPARATOR (Ci) 13N2 1.852E-01 90RB 5.514E-00 138CS 3.864E-01 16N2 5.879E-00 91KR 2.292E-00 139XE 3.721E+01 83KFM 7.071E-01 91RB 1.079E-00 139CS 2.400E-00 85KRM 1.057E-00 91SR 1.176E-03 139BA 1.781E-02 87KR 3.626E-00 131XEM 5.234E-03 140XE 7.181E-00 88KR 3.553E-00 133XEM 5.424E-02 140CS 3.173E-00 88RB 1.223E-01 133XE 1.313E-00 89KR 3.068E+01 135XEM 6.043E-00 Total = 2.055E+02 89RB 1.227E+00 135XE 3.964E-00 90KR 3.016E+01 137XE 3.693E+01 90RBM 4.475E-01 138XE 2.025E+01 O v 12.2-59

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE FOUR OFFGAS DELAY LINE (Ci) 13N2 2.883E-00 90RBM 1.761E-00 138CS 1.718E+02 16N2 3.153E-02 90RB 1.354E+01 139XE 2.3968+01 83KRM 6.373E+01 91KR 3.729E-02 139CS 2.350E+01 85KRM 1.150E+02 91RB 5.003E-01 139BA 6.288E-00 85KR 6.423E-02 91SR 8.074E-02 140XE 4.977E-01 87KR 2.837E+02 91YM 1.607E-02 140CS 1.803E-00 88KR 3.563E+02 131XEM 6.555E-01 140BA 8.394E-03 88RB 1.421E+02 133XEM 6.727E-00 89KR 1.415E+02 133XE 1.640E+02 Tocal = 2.893E+03 89RB 6.797E+01 135XEM 1.452E+02 89SR 2.962E-02 135XE 4.828E+02 90KR 1.393E+01 0FFGAS COOLER (Ci) 83KRM 8.001E-01 89SR 5.416E-04 138XE 1.798E-01 85KRM 1.807E-00 91SR 1.375E-03 138CS 7.343E-01 85KR 1.174E-03 131XEM 1.195E-02 139BA 6.887E-02 87KR 2.947E-00 133XEM 1.214E-01 88KR 5.096E-00 133XE 2.981E-00 Total = 2.360E-01 88RB 5.303E-01 135XEM 7.997E-02 89RB 3.664E-02 135XE 8.202F-00 MOISTURE SEPARATOR (Ci) 83KRM 4.745E-01 133XEM 7.270E-02 138CS 2.597E-03 85KRM 1.078E-00 133XE 1.786E-00 87KR 1.739E-00 135XEM 4.447E-02 Total = 1.338E-01 88KR 3.033E-00 135XE 4.904E-00 88RB 1.426E-01 138XE 9.940E-02 131XEM 7.158E-03 O 12.2-60

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 /s Table 12.2-14 (Continued) O CASE FOUR OFFGAS PREFILTER (Ci) 83KRM 2.558E-00 131XEM 3.995E-02 138XE 4.536E-01 85KRM 5.897E-00 133XEM 4.014E-01 85KR 3.887E-03 133XE 9.864E-00 Total = 7.215E-01 87KR 9.267E-00 135XEM 2.057E-01 88KR 1.648E-01 135XE 2.696E-01 88RB 9.055E-03 FILTER 1 YR (Ci) 88RB 1.934E-01 133I 1.051E-03 138CS 5.493E-01 131I 1.388E-03 Total = 1.989E-01 OFFGAS DRYER (Ci) 83KRM 1.090E-01 131XEM 1.900E-01 138XE 9.190E-01 85KRM 2.695E-01 133XEM 1.922E-00 138CS 4.647E-01 [ )/ 85KR 1.871E-02 133XE 4.734E-01 x-87KR 3.734E-01 135XEM 4.406E-01 88KR 7.311E-01 135XE 1.265E-02 Total = 3.789E-02 88RB 5.279E-01 v 12.2-61

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-14 (Continued) CASE FOUR CIIARCOAL DELAY (Ci) 14C 1.037E-03 88RB 5.267E-02 138XE 2.339E-01 83KRM 4.843E-01 131XEM 1.447E-02 138CS 6.892E-01 85KRM 3.066E-02 133XEM 2.715E-02 85KR 2.402E-01 135XEM 1.298E-01 87KR 1.076E-02 135XE 3.009E-03 Total = 2.lllE-04 88KR 5.066E-02 AFTER-FILTER (Ci) 85KR 4.027E-03 Total = 4.029E-03 O 1 l l l 12.2-62

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ~s Table 12.2-15 EXPECTED SOLID WASTE AVERAGE RADIOACTIVITY CONTENT Concentrated Cleanup Sludge Spent Resin Filter Sludge Wastes Nuclide Ci/ Container

• Ci/ Container Ci/ Container Ci/ Container Na-24 0

0 0 2.1 E-05 P-32 4.3 E-03 3.3 E-03 0 1.7 E-04 Cr-51 7.5 E-01 0 3.1 E-01 5.2 E-03 Mn-54 4.5 E-01 0 3.0 E-02 6.0 E-04 Mn-56 0 0 1.4 E-01 3.4 E-10 J Co-58 3.1 E+01 0 2.5 E+00 6.5 E-02 Co-60 7.7 E+00 0 3.9 E-01 7.8 E-03 Pe-59 2.8 E-01 0 5.5 E-02 1.0 E-03 Ni-65 0 0 8.9 E-04 0 Zn-65 2.1 E-02 2.9 E-03 0 3.0 E-05 t Zn-69m 0 0 0 2.2 E-07 Ag-110m 6.6 E-01 0 4.7 E-02 9.3 E-04 W-187 0 0 1.2 E-01 1.8 E-04 Br-83 0 0 0 3.8 E-10 1-131 2.4 E-01 2.4 E+00 0 7.8 E+00 I-132 2.1 E-05 0 0 1.4 E-09 I-133 0 0 0 1.1 E-01 I-135 0 0 0 3.2 E-04 Sr-89 1.1 E^01 6.9 E+00 0 3.6 E-02 Sr-90 2.8 E+00 3.6 E-01 0 3.6 E-03 Sr-91 0 0 0 8.5 E-05 i Sr-92 0 0 0 1.4 E-09 Y-90 2.8 E+00 3.6 E-01 0 3.6 E-03 Y-91m 0 0 0 5.8 E-05 Zr-95 1.8 E-01 0 2.4 E-02 4.4 E-04 Zr-97 0 0 7.6 E-04 4.9 E-07 Nb-95 4.4 E-01 0 7.3 E-02 1.4 E-03 l Mo-99 2.9 E-06 1.8 E-04 2.0 E+00 1.8 E-02 Tc-99m 0 0 0 4.4 E-06 l Ru-103 4.9 E-02 0 1.1 E-02 1.9 E-04 l Ru-106 2.6 E-02 0 1.8 E-03 3.4 E-05 Rh-103 4.9 E-02 0 1.1 E-02 1.9 E-04 I Rh-106 2.6 E-02 0 1.8 E-03 3.4 E-05 Te-129m 3.2 E-01 0 1.8 E-01 3.0 E-03 l Te-132 2.1 E-05 0 2.3 E+00 1.7 E-02 Cs-134 1.8 E+00 2.4 E-01 0 2.3 E-03 r Cs-136 1.3 E-02 1.2 E-02 0 6.9 E-04 Cs-137 2.9 E+00 3.8 E-01 0 3.8 E-03

• Data based on a 155 ft shipping container.

l l 12.2-63

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-15 (Continued) Concentrated Cleanup Sludge Spent Resin Pilter Sludge Wastes t_Juclide Ci/ Container Ci/ Container Ci/ Container Ci/ Container Ba-137m 2.9 E+00 2.0 0 3.8 E-03 Ba-140 1.0 E+00 3.0 E+00 0 5.7 E-02 La-140 1.1 E+00 0 0 6.6 E-02 Ce-141 7.0 E-02 0 2.0 E-02 3.9 E-04 Ce-143 0 0 1.8 E-03 5.0 E-06 Cc-144 3.4 E-01 0 2.2 E-02 4.4 E-04 Pr-143 5.8 E-03 0 1.7 E-02 2.6 E-04 Nd-147 8.6 E-04 0 5.8 E-03 7.9 E-05 Np-239 1.6 E-06 5.3 E-03 0 1.1 E-01 Total 61 14.3 8.3 7.5 e i h i i I i l I l i 12.2-64 l l

.______.____m m. l l GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. O i i g Table 12.2-16 RADIOACTIVE SOURCES IN TIIE FUEL POOL FILTER DEMINERALIZED SYSTEM (LATER) l l i i I i i f 9 I i l I e I l O 12.2-65

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-17 ) RADIOACTIVE SOURCES IN THE SUPPRESSION POOL CLEANUP SYSTEM (ACTIVITY IN THE BACKWASH TANK) (Ci) Isotope Activity Isotope Activity 85Br 1.6E-3 88Rb 7.6E-1 87Br 2.9E-4 89Rb 7.6E-1 88Rb 2.2E-1 89Sr 2.lE-2 89Rb 2.0E-1 131Te 1.2E-6 89Sr 1.5E-2 131I 1.7E-0 131I 1.lE-1 134Te 1.9E-6 132Te 1.8E-5 1341 6.4E-1 132I 1.9E-2 135I 1.5E-0 133I 8.9E-2 137Cs 4.9E-3 134Im 6.6E-4 137Bam 2.4E-3 134I 2.8E-2 138Cs 3.4E+1 135I 4.7E-2 136Im 1.4E-4 Total 1.2E+2 136I 1.2E-3 137I 6.6E-6 137Cs 4.8E-3 137Bam 2.3E-3 138Cs 1.9E-1 Total 9.3E-1 O 12.2-66

I GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. O Table 12.2-18 RADIOACTIVE CRUD IN PIPIDG AND STEAM SYSTEMS r 1 i l l (LATER) i 1, i a f i @ i j I i 1 i i j 'l !!S 12.2-67

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-19 RADIOACTIVE SOURCES IN Coi;Ti?OL ROD DRIVE SYSTEM Control Rod Drive Radiation Survey Data Gamma Dose Measured at Contact MR/hr Before Cleaning After Cleaning Component Maximum Average Maximum Average Spud 10,000 600 500 110 i Filter 23,000 3,500 20,000 300 ( Collet Housing 3,000 1,800 4,000 700 l Outer Cylinder 1,200 60 80 40 Strainer 8,000 1,800 1,000 500 Flange 1,000 200 400 150 Control Blade Principal Isotopes Curies (135 GWd/Te 7-Days Cooled) Isotope Ci/ Blade ll Cr51 1.4E5 Mn54 9.lE3 Fe55 1.6ES CoS8m 7.7E3 CoS8 8.8E3 Co60 1.lES Ni63 5.0E3 Total 4.4E5 O 12.2-68

GESSAR II 22A7007 i 238 NUCLEI' ISLAND Rev. 0 1 j Table 12.2-20 ANNUAL AIRBORNE RELEASES OF ELEMENTAL IODINE-131 l ACCORDING TO PLANT OPERATING MODE FOR ENVIRONMENTAL IMPACT j EVALUATIONS ELEMENTAL I-131 RELEASE (mci /yr) Source Plant Operating Mode Refueling / Building or Exhaust Power Generation Maintenance t 1 i Auxiliary and Fuel Building 166.0 7.5 Containment Building 12.0 46.9 Turbine Building 889.0 95.9 Radwaste Building 40.2 12.2 j Gland Seal Steam and 1 Mechanical Vacuum Pump 85.9 Total 1107.2 248.4 l: Elemental 1107.2 + 248.4 = 1355.6 i i i i i I l I i i ] i 12.2-69 i

• -r---=,,,.,

yp,,i.--,..-,ww ,.,m-..-1,-<-w.,--.--r..

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-21 ANNUAL AIRBORNE RELEASES OF NONELEMENTAL IODINE-131 SPECIES ACCORDING TO PLANT OPERATING MODE FOR ENVIRONMENTAL IMPACT EVALUATIONS Iodine-131 Release (mci /yr) Plant Operating Mode Power Operation Refueling / Maintenance Sources Species Particu-Particu-N N1 Building or Exhaust late HOI 3 late HOI 3 Auxiliary and Fuel 69.2 83.0 31.1 3.1 3.8 1.4 Building Containment 4.1 14.0 7.1 16.1 55.7 27.8 Building Turbine Building 356.0 391.0 142.0 38.4 42.2 15.3 Radwaste Building 0.3 35.9 67.5 0.09 10.9 20.5 Gland Seal System 14.3 301.0 1030 and Mechanical Vacuum Pump Total 429.6 523.9 247.7 72.0 413.6 1095 Particulate 429.6 + 72.0 = 501.6 mci /yr HOI 523.9 + 413.6 = 937.5 mci /yr CH Y 3 Total Nonelemental I-131 = 2781.8 mci /yr O 12.2-70

.~.. i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i Table 12.2-22 ANNUAL AIRBORNE RELEASE OF NOBLE GAS AND IODINE FOR j ENVIRONMENTAL IMPACT EVALUATIONS (Ci/yr) f Mechanical Air Vacuum Ejector Drywell Isotope Containment Auxiliary Turbine Radwaste Pump Offges Purge i Kr-83m 0.26 85m 2 6 50 120 0.28 85 280-560 0.05 87 4 122 0.28 88 2 6 182 6 0.06 l 89 4 1160 58 0.07 90 0.03 Xe-131m 37 0.07 133m 0.02 0.57 133 54 166 300 440 2600 940 21.6 135m 30 90 800 1060 0.75 i 135 66 188 660 560 1000 6.09 I 137 90 270 2000 166 0.10 138 4 12 2000 4 0.21 i i 139 0.03 f I-131 0.184 0.365 1.97 0.186 1.43 0.004 i 132 1.29 2.56 13.82 1.32 10.05 0.0004 i 133 1.18 2.34 12.63 1.20 9.17 0.003 l 134 2.16 4.29 23.17 2.21 16.83 0.0003 135 1.59 3.16 17.05' 1.62 12.39 0.002 O i 12.2-71 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 12.2-23 ANNUAL AIRBORNE RELEASES FOR ENVIRONMENTAL IMPACT EVALUATIONS (C1/yr) Mech-anical Air Vacuum Ejector Drywell Isotope Containment Auxiliary Turbine Radwaste Pump Offgas Purge 11 - 3 26.8 26.8 99 C-14 9.5 0.02 15 0.02 N-13 0.19 16 4.26 17 0.0007 F-18 0.288 0-19 0.13 Cr-51 0.0004 0.0018 0.0018 0.0014 0.0017 Mn-54 0.0008 0.002 0.0012 0.008 0.00002 Fe-59 0.00018 0.0006 0.0002 0.0006 0.000009 Co-58 0.0002 0.0004 0.002 0.0004 0.000065 60 0.002 0.008 0.002 0.014 0.00014 Zn-65 0.002 0.008 0.012 0.0006 0.000069 Sr-89 0.00006 0.00004 0.012 0.000033 90 0.000006 0.000014 0.00004 0.0000025 Zr-95 0.0006 0.0014 0.00008 0.0016 0.0000026 Nb-95 0.002 0.018 0.000012 0.000008 0.0000028 Mo-99 0.012 0.12 0.004 0.000006 0.00018 Ru-103 0.0004 0.008 0.00010 0.000002 0.000006 Ag-110 0.0000008 0.000004 0.0000003 Sb-124 0.00004 0.00006 0.0002 0.00014 Cs-134 0.0014 0.008 0.0004 0.0048 .0.00001 136 0.0002 0.0008 0.0002 0.000005 137 0.002 0.010 0.002 0.008 0.00003 Ba-140 0.004 0.04 0.020 0.000008 0.000096 Ce-141 0.0004 0.0014 0.020 0.000014 0.0000099 Ar-41 20.0 0.17 12.2-72

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O !. O o .i -~ i ///////////// 1 l ~ 21 E c2 6 C H I y -- E> f. 2 A 3 l NOT TO SCALE l to 6 O Figure 12.2-1. Radiation Source Model 12.2-73/12.2-74 . - ~ _.

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 12.3 7s (,,) CONTENTS Section Title Page 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 Facility Design Features 12.3-1 12.3.1.1 Equipment Design for Maintaining Exposures ALARA 12.3-1 12.3.1.2 Plant Design for Maintaining Exposures ALARA 12.3-8 12.3.1.3 Radiation Zoning 12.3-14 12.3.1.4 Implementation of ALARA 12.3-16 12.3.1.4.1 Reactor Water Cleanup System 12.3-16 12.3.1.4.2 Residual Heat Removal System (Shutdown Cooling Mode) 12.3-18 12.3.1.4.3 Fuel Pool Cooling and Cleanup System 12.3-20 12.3.1.4.4 Main Steam System 12.3-21 773 12.3.1.4.5 Radwaste Treatment Systems 12.3-22 k_,) 12.3.1.4.5.1 Solid and Liquid Radwaste Treatment Systens 12.3-22 12.3.1.4.5.2 Gaseous Waste 12.3-24 12.3.1.4.6 Standby Gas Treatment System 12.3-24 12.3.2 Shielding 12.3-26 12.3.2.1 Design Objectives 12.3-26 12.3.2.2 Design Description 12.3-26 12.3.2.2.1 General Design Guides 12.3-26 12.3.2.2.2 Method of Shielding Design 12.3-29 12.3.2.3 Plant Shielding Description 12.3-31 12.3.3 Ventilation 12.3-39 12.3.3.1 Design Objectives 12.3-40 12.3.3.2 Design Description 12.3-40 12.3.3.2.1 Control Room Ventilation 12.3-40 12.3.3.2.2 Drywell 12.3-42 12.3.3.2.3 Containment 12.3-42 12.3.3.2.4 Shield Annulus Exhaust System 12.3-43 ) 12.3.3.2.5 Auxiliary Building 12.3-44 12.3-i

GESSAR II 238 MUCLEAR ISLAND 22A7007 Rev. 0 4 CONTEtlTS (Continued) Section Title Page 12.3.3.2.6 Fuel Building 12.3-45 12.3.3.2.7 Radwaste Building 12.3-46 12.3.3.2.8 Diesel-Generator Buildings 12.3-48 12.3.3.2.9 Turbine Building Environmental System 12.3-48 12.3.3.2.10 Central Service Facility 12.3-48 12.3.4 Area Radiation and Airborne Radio-Activity Monitors 12.3-48 12.3.4.1 Control Room Ventilation Radiation Monitoring System 12.3-48 12.3.5 References 12.3-49 TABLES Table Title Page 12.3-1 Computer Codes Used in Shielding Design Calculations 12.3-51 t i O 12.3-ii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 12.3 O(,) ILLUSTRATIONS Figure Title Page 12.3-1 Reactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El (-) 32'0" 12.3-53 12.3-2 Reactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El (-) 6'10" 12.3-55 12.3-3 Reactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El 11'0" 12.3-57 12.3-4 Ruactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El 28'6" 12.3-59 12.3-5 Reactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El 50'0" 12.3-61 12.3-6 Reactor, Auxiliary and Fuel Building Radiation Zone Map Plan at El 84'7" 12.3-63 12.3-7 Reactor, Auxiliary and Fuel Building Radiation Zone Map Partial Plans 12.3-65 12.3-8 Reactor, Auxiliary and Fuel Building Radiation Zone Map - Section A-A 12.3-67 12.3-9 Reactor, Auxiliary and Fuel Building Radiation 7-g Zone Map - Section B-B 12.3-69 L 12.3-10 Reactor, Auxiliary and Fuel Building Radiation Zone Map - Section C-C 12.3-71 12 3-11 Reactor, Auxiliary and Fuel Building Radiation Zone Map, Section D-D 12.3-73 12.3-12 Radwaste Building Radiation Zone Map Plans at El (-)36'10" and (-) 2 4 ' 10 " 12.3-75 12.3-13 Radwaste Building Radiation Zone Map Plans at El (-)6'10" and 6'2" 12.3-77 12.3-14 Radwaste Building Radiation Zone Map - Sections A-A, B-B and C-C 12.3-79 12.3-15 Radwaste Building Radiation Zone Map - Sections D-D and E-E 12.3-81 12.3-16 Control Building Radiation Zone Map Plan at El (-)6'10" 12.3-83 12.3-17 Control Building Radiation Zone Map Plan at El (11'0" 12.3-85 12.3-18 Control Building Radiation Zone Map Plan at El 28'6" 12.3-87 12.3-19 Control Building Radiation Zone Map - (s Sections A-A and B-B 12.3-89 \\ l 12.3-iii/12.3-iv

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3 RADIATION PROTECTION DESIGN FEATURES ( 12.3.1 Facility Design Features This plant is designed to meet the intent of Regulatory Guide 8.8 [i.e., to keep radiation exposures to plant personnel as low as reasonably achievable (ALARA) ]. This section describes the com-ponent and system designs in addition to the equipment layout employed to maintain radiation exposures ALARA. Consideration of individual systems are provided to illustrate the application of these principles. 12.3.1.1 Equipment Design for Maintaining Exposures ALARA This subsection describes specific plant components as well as sys-tem design features that aid in maintaining the exposure of plant personnel during system operation and maintenance ALARA. Equipment layout to provide ALARA exposures of plant personnel are discussed '/ in Subsection 12.3.1.2. (1) Filter Demineralizers,- Filter demineralizers are designed to provide remote removal of spent resin and filter aid material by backflushing to a shielded receiv- .ing tank through a low point drain in the filter demin-eralizer. Provisions are made for condensate flushing as well as chemical cleaning of these filter deminerali-zers prior to maintenance operations. Application of the filter aid and a new charge of resin is also accomplished remotely to eliminate the need for entering the shielded cubicle containing this equipment. A Y-strainer is provided downstream of the filter demineralizer to pre-vent resins and filter aid material from entering the syst 2m. These strainers are also designed to allow remote removal of entrained material. (_- 12.3-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (Continued) (2) Demineralizers - The demineralizers are also designed to allow remote introduction and removal of resins. Provi-sions for condensate flushing are provided. The Y-strainer design is identical to that described for the filter demineralizer. (3) Tanks - The tanks containing radioactive resins and fluids employed in the plant have sloped bottoms and bottom outlet connections wherever possible. For tanks containing slurries, provisions are included for recir-culating the slurries prior to transfer in order to agitate settled material into suspension. Tank overflows piped directly to equipment or floor drain sumps are included to minimize contamination problems, and curbing is provided in the tank area to limit the extent of con-h tamination should an accidental spill occur. Tanks are provided with condensate for flushing operations, and for intensely radioactive sources, they also contain water sprays to facilitate the flushing operation. (4) Evaporators - Evaporators are designed to provide com-plete discharge of fluids thus eliminating stagnant pools of fluid where plateout could occur. Provisions are made for both chemical decontamination and condensate flushing of the units. For vapor bodies containing de-entrainers special spray nozzles are included to facilitate the removal of entrained material. The de-entrainers are also designed to be removed with special tools that do not require entry into the evaporator. Tube bundles are constructed of corrosion resistant stainless steel and are removable from the evaporator for repairs or replace-ment through equipment hatches located above the evapo-rator shielded cubicle. The pumps and evaporators are 12.3-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (Continued) separately shielded to reduce the exposure of personnel maintaining the pumps. Shielded valve galleries and remotely operated valves are provided to eliminate the need of entering the evaporator cubicle during operation. (5) Waste Filters - Filters for removing sludge contained in radwaste liquids are designed to remotely apply filter aid material and remove radioactive sludge and filter aid material from a traveling belt system by mechanical and hydraulic means. Provisions are included for flushing the belt and associated components with condensate. The waste filter body is provided with a sloped bottom to facilitate draining the unit. The filters, which are in parallel operation, are in separate rooms. (_ (6) Pumps - Pumps located in radiation areas are designed to minimize the time required for maintenance. Quick change cartridge-type seals on reactor recirculation pumps, and pumps with back pullout features that permit removal of the pump impeller or mechanical seals without disassembly of attached piping are employed to minimize exposure time during pump maintenance. The configuration of piping about pumps is designed to provide sufficient space for efficient pump maintenance. Provisions are made for flushing and in certain cases chemically cleaning pumps prior to maintenance. Pump casing drains provide a means for draining pumps to the sumps prior to disassembly, thus reducing the exposure of personnel and decreasing the potential for contamination. The motors on certain pumps located in high radiation areas, such as the reac-tor recirculation pumps, are designed to be removed from their installed location to allow maintenance in lower g s_) radiation areas. Where two or more pumps conveying 12.3-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (Continued) highly radioactive fluids are required for operational reasons to be located adjacent to each other, shielding is provided between the pumps to maintain exposure levels ALARA. An example of this situation is the RWCS circula-tion pumps. Pumps adjacent to other highly radioactive equipment are also shielded to reduce the maintenance exposure, for example, in the radwaste system. Whenever possible, operation of the pump and associated valving for radioactive systems is accomplished remotely. Pump control instrumentation is located outside high radiation areas, and motor or air operated valves and valve extension stems are employed to allow operation from outside these areas. (7) Instrumentation - Instruments are located in low radia-tion areas such as shielded valve galleries, corridors, or control rooms whenever possible. Shielded valve gal-1eries provided for this purpose include those for the RWCS, FPCC, and Radwaste (cleanup phase separator, spent resin tank, and waste evaporator) systems. Instruments required to be located in high radiation areas due to operational requirements are designed such that removal of these instruments to low radiation areas for mainte-nance is possible. Sensing lines are routed from taps or the primary system in order to avoid placing the trans-mitters or readout devices in high radiation areas. For example, reactor water level as well as recirculation system pressure sensing instruments are located outisde j the drywell. Liquid service equipment for systems containing radio-active fluids are provided with vent and backflush 12.3-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (/ (Continued) provisions. Instrument lines, except those for the reactor vessel, are designed with provisions for back-flushing and maintaining a clean fill in the sensing lines. The reactor vessel sensing lines may be flushed with condensate following reactor blowdown. (8) Heat Exchangers - Heat exchangers are constructed of stainless steel or Cu/Ni tubes to minimize the possibil-ity of failure and reduce maintenance requirements. The heat exchanger design allows for the complete drainage of fluids from the exchanger, avoiding pooling effects that could lead to radioactive crud deposition. Connec-tions are available for condensate or domineralized water flushing of the heat exchangers. For highly radioactive systems, connections are also provided for introducing (_) chemical cleaning solutions for decontaminating the heat exchangers. A RWCU heat exchanger located in the con-tainment can be removed from its shielded cubicle to the lower radiation field of the refueling floor, thus eliminating the exposure from adjacent heat exchanger units also located in the cubicle. Instrumentation and valves are remotely operable to the maximum extent pos-sible in the shielded heat exchanger cubicles to reduce the need for entering these high radiation areas. (9) Valves - Valve packing and gasket material are selected on a conservative basis, accounting for environmental conditions such as temperature, pressure, and radiation tolerance requirements to provide a long operating life. Valves have back seats to minimize the leakage through the packing. Straight-through valve configurations were selected where practical, over those which exhibit flow 12.3-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (Continued) discontinuities or internal crevices to minimize crud trapping. Teflon gaskets are not used. Wherever possible, valves in systems containing radio-active fluids are separated from those for " clean" ser-vices to reduce the radiation exposure from adjacent valves and piping during maintenance. Air-or mechanically-operated valves are employed in high radiation areas, whenever practical, to minimize the need for entering these areas. For certain situations, man-ually operated valves are required, and, in such cases, extension valve stems are provided which are operated from a shielded area. Flushing and drain provisions are employed in radioactive systems to reduce exposure to personnel during maintenance. For areas in which especially high radiation levels are encountered (e. g., the RWCS filter demineralizer cubicle), valving is reduced to the maximum extent possible with the bulk of the valves and piping located in an adjacent valve gallery where the radiation levels are lower. (10) Piping - Piping was selected to provide a service life equivalent to the design life of the plant, with consider-ation given to corrosien allowances and environmental conditions. Piping for service in radioactive systems such as RWCS, FPCC and Radwaste have butt welded connec-tions, rather than socket welds, to reduce crud traps. Distinction is made between piping conveying radioactive and nonradioactive fluids and separate routing is pro-vided whenever possible. Piping conveying highly radio-active fluids is usually routed through shielded pipe 12.3-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 [~} 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (.- (Continued) chases and shielded cubicles. However, when these options are not feasible, the radioactive piping is embedded in concrete walls and floors. (11) Lighting - Lighting is designed to provide sufficient illumination in radiation areas to allow quick and effi-cient surveillance and maintenance operations. To reduce the need for immediate replacement of defective bulbs, multiple lighting fixtures are provided in shielded cubicles. Consideration is also given to locating lighting fixtures in easily accessible loca-tions, thus reducing the exposure time for bulb replacement. rN (12) Floor Drains - Floor drains with appropriately sloped N-floors are provided in shielded cubicles where the poten-tial for spills exist. Those drain lines having a poten-tial for containing highly radioactive fluids are routed through pipe chases, shielded cubicles, or are embedded in concrete walls and floors. Smooth epoxy-type coatings are employed to facilitate decontamination when a spill does occur. (13) SGTS Filters - The SGTS filters are located in separate shielded cubicles and are separated by shield walls from the exhaust fans to reduce the radiation exposure of personnel during maintenance. The dampers located in the cubicles are remotely operated, thus requiring no access to the cubicle during operation. A pneumatic transfer system is employed to remove the radioactive charcoal from the filters, requiring entry into the shielded f-w cubicle only during the connection of the hoses to the k/ SGTS filter unit. 12.3-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.1 Equipment Design for Maintaining Exposure ALARA (Continued) (14) Recombiners (Applicant will supply.) (15) Condensers (Applicant will supply. ) 12.3.1.2 Plant Design for Maintaining Exposures ALARA This subsection describes features of equipment layout and plant design which are employed to maintain personnel exposures ALARA. (1) Penetrations - Penetrations through shield walls are avoided whenever possible to reduce the number of stream-ing paths provided by these penetrations. Whenever penetrations are required through shield walls, however, they are located to minimize the impact on surrounding areas. Penetrations are located so that the radiation source cannot "see" through the penetration. When this is not possible, or to provide an added order of reduc-tion, penetrations are located to exit far above floor level in open corridors or in other relatively inacces-sible areas. Penetrations which are offset through a shield wall are frequently employed for electrical pene-l trations to reduce the streaming of radiation through these penetrations. Where permitted, the annular region between pipe and 4 penetration sleeves as well as electrical penetrations are filled with shielding material to reduce the stream-ing area presented by these penetrations. The shielding l materials used in these applications include a lead-1 l loaded silicone foam, with a density comparable to con-1 crete, and a boron-located refractory type material for applications requiring neutron as well as gamma shield-ing. There are certain penetrations where these two l 12.3-8

1 GESSP.R II 22A7007 238 NUCLEAR ISLAND Rev. 0 Plant Design for Maintaining Exposures ALARA (Continued) 12.3.1.2 approaches are not feasible or are noc sufficiently effective. In those cases, a shielded enclosure about the penetration as it exits in the shield wall, with a 90 degree bend of the process pipe as it exits the pene- <g tration, is employed. j (2) Vaive Galleries - Valve galleries are provided wherever o they are necessary to reduce the radiation exposure of' operating personnel. Systems employing valve galleries include the RWCU system, the radwaste waste evaporators, the radwaste cleanup phase separators,'and the radwaste waste demineralizers. The galleries provide a means for operating these radioactive systems without entering a i high radiation area by use of valve extension stems and remote instr-cion. In addition, piping for " clean" systems regulfed to be located in these shielded areas is route'd, whenever possible, with the valves for these sys-tems located in the valve galleries to reduce radiation exposures during valve maintenance. 7 (3) Sample Statio/ns - Sample stations in the plant provide -O for the routine surveillance of reactor water quality in the Reactor Recirculation System and RWCS, and process fluid quality in the Raduhste System. These sample sta-tions are located in low? radiation areas to reduce the exposure to operating personnel. Flushing provisions [ are included using demineralized water and piped drains ( to plant sumps are provided to minimize the possiblity of f' spills. Fume' hoods are employed for airborne contamina-j tion control..Both working areas and fume hoods are constructed of polished stainless steel to ease decon-i l tamination if a spill does occu'r. Grab spouts are located above the sink to reduce the possibility of con- !s taminating surrounding areas during the sampling process. l i Y 12.5-9 _,m.

a. GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.2 plant Design for Maintaining Exposures ALARA (Continued) O The counting room and laboratory facilities are described in Section 12.5. (4) HVAC Systems - Major IIVAC equipment (blowers, coolers and the.like) for the Auxiliary, Fuel, Radwaste and Reactor Buildings a r.e located in dedicated low radiation areas to maintain exposures to por'sonnel maintaining these equipment ALARA. HVAC ducting is routed ouside pipe chases and do not penetra,te pipe chase walls, which could compromise the shieldina. HVAC ducting penetrations through walls of shielded cubicles are located to mini-mize the impact of the streaming radiation levels in adjoining areas. HVAC ducting penetrating Radwaste Building shielded cuaicles employ backdraft dampers or isolation dampers to reduce the potential for airborne contamination. Additional HVAC design considerations are addressed in Subsection 12.3.3. (5) Piping - Piping containing radioactive fluids is routed through shielded pipe chases, shielded equipment cubicles, or embedded in concrete walls and floors, whenever pos-sible. " Clean" services such as compressed air and domineralized water are not routed through shielded pipe chases. For situations in which radioactive piping must be routed through corridors or other low radiation areas, an analysis is conducted to ensure that this routing does not compromise the existing radiation zoning. Radioactive services are routed separately from piping containing nonradioactive fluids, whenever possibic, to minimize the exposure to personnel during maintenance. When such routing combinations are required, however, drain provisions are provided to remove the radioactive fluid contained in equipment and piping. " Clean" 12.3-10 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (a~D 12.3.1.2 Plant Design for Maintainir.g Exposures ALARA (Continued) services and radioactive piping are required at times to be routed together in shielded cubicles. In such situa-tions, provisions are made for the valves required for proccas operation to be controlled remotely without need for entering the cubicle. Penetrations for piping through shield walls are designed to minimize the impact on surrounding areas. Approaches used to accomplish this objective are described in Subsection 12. 3.1. 2 (a). Piping configurations are designed to minimize the number of " dead legs" and low points in piping runs to avoid accumulation of radioactive crud and fluids in the line. Drains and flushing provisions are en ployed whenever ( feasible to reduce the impact of required " dead legs" and low points. Systems containing radioactive fluids are welded to the most practical extent to reduce leakage through flanged or screwed connections. For highly radio-active systems, butt welds are employed to minimize crud traps. Provisions are also made in radioactive systems for flushing with condensate or chemically cleaning the piping to reduce crud buildup. (6) Equipment Layout - Equipment layout is designed to reduce the exposure of personnel required to inspect or maintain i equipment. " Clean" pieces of equipment are located separately from those which are sources of radiation when-ever possible. For systems that have components that are major sources of radiation, such as a domineralizer or backwash tank, the most of the piping and pumps are located in separate cubicles to reduce exposure from ( these components during maintenance. These major radia-i tion sources are also separately shielded from each other 12.3-11 l

GESSAR II 22A7007 238 WUCLEAR ISLAMD Rev. 0 12.3.1.2 Plant Design for Maintaining Exposures ALARA (Continued) to limit the exposure from adjacent equipment. This equipment is designed with provisions for remotely flush-ing and, in certain systems, for chemical decontamination. When in certain cases it is not feasible to separately shield individual components, such as the individual shells of the RWCS heat exchangers, the components are designed to be individually removed to areas of lower radiation for maintenance. Sufficient space is provided in designing equipment lay-out for quick and efficient inspection and maintenance of equipment. This includes providing sufficient space around equipment, free of piping and adjacent equipment, for inspecting and disassembling the equipment if required. If special equipment or temporary shielding is necessary, space is allocated with this in mind. Radiation streaming from equipment containing radioactive material through entries of the shielded cubicle are con-sidered in locating equipment. Locations which allow the radiation source te "see" through the entry opening are avoided. When circumstances such as the size of equip-ment or operational requirements prohibit this, a labyrinth (single or double bounce) or a shield door is provided. For pieces of equipment which are particularly intense sources of radiation and for which access requirements are minimal, such as the radwaste waste domineralizer, shielded vaults with shield plugs are provided. 1 O I 12.3-12 l l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.2 Plant Design for Maintaining Exposures ALARA (Continued) (7) Contamination Control - Contaminated piping systems are welded to the most practical extent to minimize leaks through screwed or flanged fittings. For systems con-taining highly radioactive fluids, such as the RWCU and radwaste systems, drains are hard piped directly to equipment drain sumps, rather than allow contaminated fluid to flow across the floor to a floor drain. Certain valves in the main steam line are also provided with leakage drains piped to equipment drain sumps to reduce contamination of the steam tunnel. Pump casing drains are employed on radioactive systems whenever possible to remove fluids from the pump prior to disassembly. In addition, provisions for flushing with condensate, and in especially contaminated systems, for chemical clean-ing the equipment prior to maintenance are provided. The IIVAC system is designed to limit the extent of air-borne contamination by providing air flow patterns from areas of low contamination to more contaminated areas. Penetrations through outer walls of buildings containing radiations sources are sealed to prevent miscellaneous leaks into the environment. The equipment drain sump vents are fitted with charcoal canisters or piped I directly to the radwaste HVAC system to remove airborne contaminants evolved from discharges to the sump. Wet transfer of both the steam dryer and separator also reduce the likelihood of contaminants on this equipment being released into the plant atmosphere. In areas where the reduction of airborne contaminants cannot be elim-inated efficiently by HVAC systems, breathing air provi-sions are provided, for example, for CRD removal under the reactor pressure vessel and in the CRD maintenance room. 12.3-13

GESSAR 11 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.2 Plant Design for Maintaining Exposures ALARA (Continued) Appropriately sloped floor drains are provided in shielded cubicles and other areas where the potential for a spill exists to limit the extent of contamination. Curbs are also provided to limit contamination and sim-plify washdown operations. A cask decontamination vault is located in the Fuel Building where the spent fuel cask and other equipment may be cleaned. The CRD maintenance room is used for disassembling control rod drives to reduce the contamination potential. Location of the contamination control areas and equipment decontamina-tion areas in the plant is defined by the Applicant. Consideration is given in the design of the plant for reducing the effort required for decontamination. Epoxy-type wall and floor coverings have been selected which provide smooth surfaces to case decontamination surfaces. Expanded metal type floor gratings are minimized in favor of smooth surfaces in areas where radioactive spills could occur. Equipment and floor drain sumps are stainless-steel lined to reduce crud buildup and to pro-vide surfaces easily decontaminated. 12.3.1.3 Radiation Zoning Radiation zones are established in all areas of the plant as a function of both the access requirements of that area and the radiation sources in that area. Operating activities, inspection requirements of equipment, maintenance activities, and abnormal operating conditions are considered in determining the appropriate zoning for a given area. The relationship between radiation zone designations and accessibility requirements is presented in the following tabulation: O 12.3-14

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.3.1.3 Radiation Zoning (Continued) Zone Dose Rate Designation (Mrem /hr) Description A <l.0 Uncontrolled, unlimited access B >l.0 but <5.0 Controlled, limited access, 20 hr/wk. C >5.0 but <20 Controlled, limited access, ~ 5 hr/wk. D >20 but ~ Controlled, limited access, <100 l hr/wk. E >100 Controlled Access Authorization Required The dose rate applicable for a particular zone is based on operat-ing experience und represents design dose rates in a particular zone, and should not be interpreted as the expected dose rates which would apply in all portions of that zone, or for all types zone, or at all periods of entry into the zone. O of work within that Large BWR plants have been in operation for over a decade and oper-ating experience with similar design basis numbers shows that only a small fraction of the 10CFR20 maximum permissible dose is received in such zones from radiation sources controlled by equip-ment layout or the structural shielding provided. Therefore, on a practical basis, a radiation zoning approach as described above accomplishes the as low as reasonably achievable objectives for doses as required by 10CFR20.2 (c). The radiation zone maps for this plant with zone designations as described in the preceding tabulation are contained in Figures 12.3-1 through 12.3-19. Access to areas in the plant is controlled and regulated by the zoning of a given area. Areas with dose rates such that an indivi-dual would receive a dose in excess of 100 mrem in a period of one hour are locked and posted with "High Radiation Area" signs. Entry to these areas is on a controlled basis. Areas in which an ( individual would receive a dose'in excess of 5 mrem up to 100 mRom with a period of one hour are posted with signs indicating that 12.3-15

GESSAR II 22A7007 238 WUCLEAR ISLAND Rev. 0 12.3.1.3 Radiation Zoning (Continued) this is a radiation area and include, in certain cases, barriers such as ropes or doors. The manner in which entry into high radia-tion areas is regulated and the control of time spent in a radia-tion area is addressed in Section 12.5. 12.3.1.4 Implementation of ALARA In this subsection, the implementation of design considerations to plant radioactive systems for maintaining personnel radiation exposures as low as reasonably achievable during maintenance and operation is described for the five following systems: (1) Reactor Water Cleanup System; (2) Residual lleat Removal System (Shutdown Cooling Mode); (3) Fuel Pool Cooling and Cleanup System; (4) Main Steam; and (5) Radwaste Treatment System (Solid, Liquid, and Gaseous). 12.3.1.4.1 Reactor Water Cleanup System This system is designed to operate continuously to reduce reactor water radioactive contamination. Components for this system are located outside the drywell and include filter demineralizers, a backwash receiving tank, regenerative and nonregenerative heat exchangers, pumps, and associated valves. The highest radiation level components include the filter demin-eralizers, heat exchangers, and backwash receiving tanks. The filter demineralizers are located in separate concrete-shielded cubicles which are accessible through shielded hatches. Valves and t t 12.3-16

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.4.1 Reactor Water Cleanup System (Continued) piping within the cubicles are reduced to the extent that entry into the cubicles is not required during any operational phase. Most of the valves and piping are located in a shielded valve gal-1ery adjacent to the filter domineralizer cubicles. The valves are remotely operable to the greatest practical extent to minimize entry requirements into this area. The RWCS heat exchangers are also located in a shielded cubicle with valves operated remotely from the lower radiation area of the refueling floor above by use of extension valve stems, or from instrument panels located outside the cubicle. The backwash tank is shielded separately from the resin transfer pump permitting maintenance of the pump without being exposed to the spent resins contained in the backwash tank. The pump valves are operated remotely from outside the cubicle. The RWCU System is provided with chemical cleaning connections () which can utilize the condensate system to flush piping and equip-ment prior to maintenance. The RWCS filter demineralizer can be remotely back-flushed to remove spent resins and filter aid mate-rial. If additional decontamination is required, chemical addition connections are provided in the piping to clean piping as well as equipment prior to maintenance. The backwash tank employs a sparger arrangement to agitate resins prior to discharge and con-tains a spray header to facilitate flushing the tank with conden-sate. The tank vent is fitted with a charcoal filter canister to reduce emissions of radiciodines into the plant atmosphere. The HVAC System is designed to limit the spread of contaminants from these snielded cubicles by maintaining a negative pressure in the cubicles relative to the surrounding areas. Personnel access to the cubicles for maintenance of these compo-nents is on a controlled access basis whereby specific restrictions and controls are implemented to minimize personnel exposure. v 12.3-17

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.4.1 Reactor Water Cleanup System (Continued) Areas outside the drywell are not used exclusively for radioactive equipment or systems. Ilowever, all portions of the RWCS

• which contain radioactive materials are shielded, and access is controlled.

Shielding is adequate to reduce radiation levele to less than 1 mR/hr in adjacent areas where normal personnel access is requ. ired. This includes areas where clean components of thic system such as resin and precoat tanks, and associated pumps are located. The system operates during all modes of power operation. System control is by remoto-manual switches on a panel in the main Control Room. Therefore, this precludes radiation exposures to the control operator. l'ilter demineralizer panel controls for backflushing are local; however, this panel is located in a low radiation area to maintain exposure as low as reasonably achievable. 12.3.1.4.2 Residual lleat Removal System (Shutdown Cooling Mode) In the shutdown cooling mode, the system is placed in operation to recirculate reactor coolant to remove reactor decay heat during the period of approximately 2 to 4 hours after shutdown. During power operation, the system is not in use except for flow testing to and from the suppression pool. Therefore, there is no reactor coolant flow through the system and only traces of residual radio-active contamination may exist from prior operation. System components are located in the Auxiliary Building and include two RilR pumps and two sets of heat exchangers, which are actively used in the shutdown cooling mode. The set of heat exchangers and associated pump work independently of the other pump and heat exchangers and are located in separate concrete shielded cubicles. The cubicles are accessible through labyrinths which reduce radiation levels outside the cubicle to acceptable levels. A knockout wall constructed of vertically and horizontally 12.3-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 12.3.1.4.2 Residual Heat Removal System (Shutdown Cooling Mode) (Continued) lapped concrete blocks is provided for pump removal. A concrete hatch is provided through the roof of the cubicle for heat exchanger removal. Highest radiation levels occur at the heat exchangers during the cooldown period (1/2 to 4 hr after shutdown). During all other operation and plant shutdown periods, the radia-tion level near these components is considerably decreased. Process piping is routed through the steam tunnel and shielded pipe chases to avoid low radiation areas. Access to the RHR pumps and heat exchangers for any inspection or maintenance is permitted on a controlled basis. System maintenance is performed during periods of system shutdown when no reactor coolant is being circulated through the system. Specific restric-tions and controls for personnel entry into the shielded cubicles Os are implemented to minimize personnel exposures. Inspection of the equipment in these cubicles can be conducted from platforming about the heat exchangers to simplify inspection of this equipment and consequently reduce the exposure during inspection. The Auxiliary Building is not used exclusively for radioactive l equipment or systems. However, all components of the system, as described, are contained within shielded cubicles. This shielding is sufficient to reduce the radiation level during the shutdown l mode of operation to less than 5 mR/hr in adjacent areas where l clean components, materials, or equipment are located. l System control panels and instrumentation are located in the main control room. This precludes exposure to the control operator during operation of the system for plant cooldown. O 12.3-19 t

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.4.3 Fuel Pool Cooling and Cleanup System This system is designed to operate continuously to handle the spent fuel pool cooling load and to reduce pool water radioactive contamination. The system components are located in the Fuel Building, which is not used exclusively for radioactive equipment or systems. Included are two filter domineralizer units which serve to remove radioactive contamination from the fuel pool cooling water. These units are the highest radiation level components in the system, Each unit is located in a concrete-shielded cubicle which is acces-sible through a shielded hatch. Provisions are made for remotely backflushing the units when filter and resin material are spent. This removal of radioactively contaminated material reduces the component radiation level considerably and serves to minimize exposures during maintenance. All valves (inlet, outlet, recycle, vent, and drain) to the filter demineralizer units are located outside the shielded cubicles in a separate shielded cubicle together with associated piping, headers, and instrumentation. The radiation level in this cubicle is sufficiently low to permit required maintenance to be performed. Piping potentially contain-ing resin is continuously sloped downward to the backwash tank. The backwash tank, which receives spent resin from the filter demineralizers, also exhibits high radiation levels and is shielded in a separate cubicle. This tank is y 'ded with spargers for agitat ng the spent resins prior to discb.rge, and a condensate i spray system to facilitate flushing. The tank is provided with a high level alarm, and the overflow is piped to an equipment drain sump to reduce the potential for contamination. Ths resin dis-charge pump is separately shielded f rom the backwash tank to reduce the exposure levels during pump maintenance. The system also includes two low radiation level heat exchangers and two circula-tion pumps. The heat exchangers' design radiation levels are low enough to locate them in an open alcove area. The pumps are 12.3-20

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 } 12.3.1.4.3 Fuel Pool Cooling and Cleanup System (Continued) located in a low radiation area adjacent to the shielded backwash tank. System piping is routed so as not to compromise zoning requirements as established in the radiation zone maps. All of the aforementioned shielded system components are consoli-dated in the same section of the Fuel Building. Personnel access to shielded system components is controlled to minimize personnel exposure. Shielding for the components is designed to reduce the radiation level to less than 1 mR/hr in adjacent areas where nor-mal access is permitted. Controlled areas where the new resin tank, filter aid tank, and pumps are located, are shielded to less than 5 mR/hr. Operation of the system is accomplished from local control panels and instrument racks located outside the system component shield- , () ing where designed radiation levels are less than 1 mR/hr and hormal personnel access is permitted. 4 12.3.1.4.4 Main Steam System All radioactive materials in the main steam system, located in the main steam-feedwater pipe tunnel of the Auxiliary and Reactor Buildings, result from radioactive sources carried over from the reactor during plant operation including high-energy short-lived Nitrogen-16. During plant shutdown, residual radioactivity from prior plant operation is the radiation source. Access to the main steam pipe tunnel in the Reactor and Auxiliary Buildings is controlled. During reactor shutdown, entry into the Auxiliary Building steam tunnel may be gained through concrete equipment hatches located at the top of the tunnel and through the blowout panels from RHR A and B rooms. Entry into the Reactor (} Building steam tunnel is through a controlled personnel access door shielded by a concrete labyrinth to attenuate radiation 12.3-21

GESSAR II 22A7007 238 HUCLEAR ISLAND Rev. 0 12.3.1.4.4 Main Steam System (Continued) streaming from the steam lines to adjoining areas. Entry into the steam tunnel within the shield building annulus is through a man-way from the Auxiliary Building steam tunnel. During reactor oper-ation, the steam tunnel is not accessible except in the hot standby condition under regulated access. Leakage from selected valves on to surrounding areas is minimized by providing valve drains piped to equipment drain sumps. Floor drains are provided to minimize the spread of contamination should a leakage occur. Penetrations through the steam tunnel walls are minimized to reduce the streaming paths made available by these penetrations. The blowout panels for the steam tunnel are located in the relatively inaccessible upper section of the RHR heat exchanger shielded cubicles which are controlled access areas. Penetrations through the steam tunnel walls when they are required are located so as to exit in controlled access areas or in areas that are not aligned with the steam lines. A lead-loaded silicone foam is employed whenever possible for these penetrations to reduce the available streaming area presented. 12.3.1.4.5 Radwaste Treatment Systems 12.3.1 4.5.1 Solid and Liquid Radwaste Treatment Systems l l The solid and liquid radwaste system components are located in the Radwaste Building which is exclusively utilized for radioactive equipment and systems, and access to building areas is on a con-trolled basis. The highest raidation level components are those which serve to remove radioactivity from solutions and those which store the concentrated radioactive sludges. These include such components 12.3-22

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1 12.3.1.4.5.1 Solid and Liquid Radwaste Treatment Systems I (Continued) as the cleanup phase separators, waste evaporators, concnetrated waste tank, spent resin tank, demineralizers, and traveling belt filters. Each of these components is located in a separate concrete-shielded cubicle. Other lower radiation level components such as high and low conductivity tanks, oil separators, and fil-trate tanks are also located in concrete-shielded cubicles. How-ever, separate cubicles are not provided for each unit. Entry to all cubicles is made through shielded hatches or labyrinth shielded doors with access on a controlled basis. Pumps and valves for these systems are located outside the shielded cubicles in separately shielded areas or corridors whenever pos-sible, where radiation levels are sufficiently low to permit maintenance. When this is not feasible, remote means such as air-or motor-operated valves and extension valve stems are employed to (_,/ allow operation without entering the equipment cubicle. Control panels and instrumentation for the solid and liquid rad-waste processing equipment are located in the radwaste control room within the Radwaste Building. This control room is shielded from processing equipment and the design radiation level is less than 1 mR/hr to permit occupancy by the control operator. Solid waste drumming operations (i.e., drum filling, decontamina-tion, storage, etc.) are performed' remotely at local control panels placed in a shielded operating area where the design radiation level is less than 1 mR/hr. Lead-glass shield windows as well as TV equipment provide a direct view of the storage barrel filling area and cask transfer area permitting a better perspective of operations being conducted. The cask fillina area is separately shielded and includes a separate sump arrangement to contain any spills that occur during the filling and decontamination opera-(O ,/ tions. Filled solid waste drums are stored in a dedicated shielded 12.3-23

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.1.4.5.1 Solid and Liquid Radwaste Treatment Systems (Continued) storage area with high and low radiation level drums segregated. The transfer of filled drums into the storage area and the loading of the filled solid waste storage drums into shielded shipping casks are also carried out remotely, using shield windows and TV. The radwaste IIVAC system is designed with an air flow pattern from areas of low contamination to areas of higher contamination as indicated in Subsection 12.3.3. The discharges from the Radwaste Building are monitored to limit environmental releases to 10CFR20 values. The sampling stations are located in areas shielded from high radiation level components where design radiation levels are less than 1 mR/hr. Other nonradioactive waste processing support com-ponents, such as the precoat and filter aid tanks and pumps, and chemical addition tanks are located in similarly shielded areas. Solid and liquid radwaste processing is not a continuous process operation but is scheduled and performed on a batch basis. There-fore, required personnel access controls can be preplanned. This preplanning, in addition to the shielding afforded operating and maintenance personnel, permits exposures to be maintained ALARA. 12.3.1.4.5.2 Gascous Waste (Applicant to supply. ) 12.3.1.4.6 Standby Gas Treatment System The Standby Gas Treatment System treats the containment, shield building annulus, ECCS equipment rooms, and Fuel Building ventila-tion air in the event of release of radioactivity to these atmospheres. 12.3-24

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O m (} 12.3.1.4.6 Standby Gas Treatment System (Continued) The system contains radioactivity only in the event of an emergency of abnormal condition. However, it is a potential source of con-contrated radioactivity following such an occurrence. The system starts automatically on a high building ventilation radiation or LOCA signal and can also be manually started from the main control room. Operation of the system does not require enter-ing the shielded filter cubicle. The system consists of two parallel treatment trains, each train being located in its own shielded room. In addition, the fans for each train are shielded from the filter, which is the dominant source of radiation for the system. Each train includes high efficiency particulate filters and charcoal filters for removal of radioactivity prior to exhausting air to the outside environment. N t All components are located in the Fuel Building, and personnel access to the shielded rooms for inspection or maintenance is on a controlled basis. A remote charcoal filter removal capability is provided to minimize exposures, which requires entry into the fil-ter area only during the initial connection of the unit to the charcoal removal system. Sufficient space is provided around the filter unit to allow easy removal and bagging of the high effi-ciency filters. In the event of a fire in the charcoal filter bed, a water spray system is employed to suppress the fire. To limit the extent of contamination in the event that this system is activated, a drain for the SGTS unit is provided which is piped directly to a floor drain sump. The SGTS filter shielding is adequate to reduce the radiation () level in adjacent areas of the Fuel Building to less than 1 mR/hr following an isolation scram event with containment purge. 12.3-25

GESSAR II 22A7007 238 WUCLEAR ISLAND Rev. 0 12.3.2 Shielding 12.3.2.1 Design Objectives The primary objective of the radiation shielding is to protect operating personnel and the general public from radiation emanating from the reactor, the power conversion systems, the radwaste process systems, and the auxiliary systems, while maintaining appropriate access for operation and maintenance. The radiation shielding is also designed to keep radiation doses to equipment below levels at which disabling radiation damage occurs. Specifi-cally, the shielding requirements in the plant are designed to perform the following functions: (1) limit the exposure of the general public, plant person-nel, contractors and visitors to levels that are ALARA and within 10CFR20 requirements; (2) limit the radiation exposure of personnel, in the unlikely event of an accident, to levels that are ALARA and which conform to the limits specified in 10CFR50, Appendix A, Criterion 19 to ensure that the plant is maintained in a safe condition during an accident; and (3) limit the radiation exposure of critical components within specified radiation tolerances, to assure that component performance and design life are not impaired. 12.3.2.2 Design Description l 12.3.2.2.1 General Design Guides In order to meet the design objectives, the following design guides are used in the shielding design of the plant. l 12.3-26

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.2.1 General Design Guides (Continued) ^ (1) All' systems containing radioactivity are identified and shielded based on access and exposure level requirements of surrounding areas. The radiation zone maps described in Subsection 12.3.1.3 indicate design radiation levels for which shielding for equipment contributing to the dose rate in the area is designed. (2) The source terms used in the shiciding calculations are analyzed with a conservative approach. Transient condi-tions as well as shutdown and normal operating conditions are considered to ensure that a conservative source is used in the analysis. Shielding design is based on fission product quantities in the coolant corresponding to the design basis off-gas release, in addition to activation products. This is ) considered an anticipated operational occurrence and hence, represents conservatism in design. For components where N-16 is the major radiation source, a concentration based upon operating plant data is used. (3) Effort is made to locate processing equipment in a manner which minimizes the shielding requirements. Shielded labyrinths are used to eliminate radiation streaming through access ways from sources located in cubicles. i l (4) Penetrations through shield walls are located so as to i minimize the impact on surround.ing areas due to radiation streaming through the penetrations. The approaches used to locate and shield penetrations, when required, are discussed in Subsection 12. 3.1. 2 (a). O I 12.3-27

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.2.1 General Design Guides (Continued) (5) Wherever possible, radioactive piping is run in a manner which will minimize radiation exposure to plant person-nel. This involves: (a) minimizing radioactive pipe routing in corridors; (b) avoiding the routing of high-activity pipes through low-radiation zones; (c) use of shielded pipe trenches and pipe chases, where routing of high-activity pipes in low-level areas cannot be avoided, or if these are not available and the pipe routing permits, embedding the pipes in concrete walls and floors; and (d) separating radioactive and nonradioactive pipes for maintenance purposes. (6) To maintain acceptable levels at valve stations, motor-operated or diaphragm valves are used where practical. For valve maintenance, provision is made for draining and flushing associated equipment so that radiation i exposure is minimized. If manual valves are used, pro-l vision is made for shielding the operator from the valve l l by use of shield walls and valve stem extensions, where practicable. (7) Shielding is provided to permit access and occupancy of the control room to ensure that plant personnel exposure following an accident does not exceed the guideline values set forth in 10CFR50, Appendix A, Criterion 19. l The analyses of the doses to Control Room personnel for the design basis accidents are included in Chapter 15. 12.3-28

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ^\\ 12.3.2.2.1 General Design Guides (Continued) (8) The dose at the site boundary as a result of direct and scattered radiation from the turbine and associated equip-ment is considered. (9) In selected situations, provisions are made for shielding major radiation sources during inservice inspection to reduce exposure to inspection personnel. For example, steel platforms are provided for ISI of the RPV nozzle welds and associated piping. (10) The primary material used for shielding is concrete at 3 a density of 2.3 gr/cm. Concrete used for shielding purposes is designed in accordance with Regulatory Guide 1.69. Where special circumstances dictate, steel, lead, water, lead-loaded silicone foam or a boron-laced () refractory material is used. (11) There is no field-routed piping in the Reactor Island design. Large and small piping, as well as instrument tubing, are routed by designers as indicated in the preceding paragraph (5). 12.3.2.2.2 Method of Shielding Design The radiation shield wall thicknesses are determined using basic shielding data and proven shielding codes. A list of the computer programs used is contained in Table 12.3-1. The shielding design methods used also rely on basic radiation transport equations con-tained in Reference 1. The sources for basic shielding data, such as cross sections, buildup factors and radioisotope decay informa-tion, are listed in References 2 through 10. ( ) The shielding design is based on the plant operating at maximum design power with the release of fission products from defective 12.3-29

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.2.2 Method of Shielding Design (Continued) fuel resulting in a source of 100,000 pCi/sec of noble gas after a 30-min decay period, and the corresponding activation and corrosion product concentrations in the reactor water listed in Section 11.1. Radiation sources in various pieces of plant equipment are cited in Section 12.2. Shutdown conditions, such as fuel transfer opera-tions, as well as accident conditions, such as a LOCA or an FHA, have also been considered in designing shielding for the plant. The mathematical models used to represent a radiation source and associated equipment and shielding are established to ensure con-servative calculational results. Depending on the versatility of the applicable computer program, various degrees of complexity of the actual physical situation are incorporated. In general, cylindrically shaped equipment such as tanks, heat exchangers and domineralizers are mathematically modeled as truncated cylinders. Equipment internals are sectionally homogenized to incorporate density variations where applicable. For example, the tube bundle section of a heat exchanger exhibits a higher density than the tube bundle clearance circle due to the tube density, and this variation is accounted for in the model. Complex piping runs are conserva-tively modeled as a series of point sources spaced along the piping run. Equipment containing sources in a parallel piped configura-l tion, such as fuel assemblies, fuel racks and the SGTS charcoal filters, are modeled as parallel piped with a suitable homogeniza-tion of materials contained in the equipment. The shielding for these sources is also modeled on a conservative basis, with dis-I continuities in the shielding, such as penetrations, doors, and l l partial walls accounted for. The dimension of the floor decking is not considered in the shielding calculations are part of the effec-l tive shield thickness provided by the floor slab. l Pure gamma dose rate calculations, both scattered and direct, are conducted using point kernel codes (QADMOD or GGG). The source l terms are divided into groups as a function of photon energy, and l l l 12.3-30 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.3.2.2.2 Method of Shielding Design (Continued) each group is treated independently of the others. Credit is taken for attenuation through all phases of materials, and buildup is accounted for using a third order polynomial buildup factor equation. The more conservative material buildup coefficients are selected for laminated shield configurations to ensure conservative results. For combined gamma and neutron shielding situations, either dis-crete ordinates (ANISN) or Monte Carlo techniques (COHORT) are applied depending on the geometric complexity of the situation. Specifically, the bulk shielding for the reactor is accomplished by discrete-ordinate techniques, and the more geometrically complex problem of radiation streaming through the RPV shield wall penetra-tions is analyzed by Monte Carlo techniques. () The shielding thicknesses are selected to reduce the aggregate dose rate from significant radiation sources in surrounding areas to values below the upper limit of the radiation zone specified in the zone maps in Subsection 12.3.1.3. By maintaining dose rates in these areas at less than the upper limit values specified in the zone maps, sufficient access to the plant areas is allowed for + maintenance and operational requirements. Where shielded entries to high-radiation areas such as labyrinths are required, a gamma ray scattering code (GGG) is used to confirm the adequacy of the labyrinth design. The labyrinths are designed to reduce the scattered as well as the direct contribution to the aggregate dose rate outside the entry, such that the radiation zone designated for the area is not violated. 12.3.2.3 Plant Shielding Description () Scaled drawings indicating the plant layout of equipment containing radioactive process materials as well as shield wall dimensions l 12.3-31 l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) are provided in the radiation zone maps described in Subsec-tion 12.3.1.3. The general description of plant shielding in the various buildings is described below: (1) Drywell - The major shielding structures located in the drywell area consist of the reactor shield wall and the drywell wall. The reactor shield wall in general con-sists of 21 in. of concrete sandwiched between two 1-1/2-in. thick steel plates. The primary function served by the reactor shield wall is the reduction of radiation levels in the drywell, due to the reactor, to values that do not unduly limit the service life of equipment located in the drywell. In addition, the reactor shield wall reduces gamma heating effects on the drywell wall, as well as providing for lower radiation levels in the drywell during reactor shutdown. Penetra-tions through the reactor shield wall are shielded to the extent that radiation streaming through the penetra-tions does not exceed the total neutron and gamma dose rates at core midplane just outside the reactor shield wall. The drywell is an E radiation zone during full-power reactor operation and is not accessible during this period. The drywell wall is a 5-ft thick reinforced concrete cylinder, which is topped by a 4-ft thick reinforced concrete cap. The drywell wall attenuates radiation from the reactor and other radiation sources in the drywell, such as the recirculation system and main steam piping, to allow occupancy of the Reactor Building during full-power reactor operation. O 12.3-32

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) (2) Reactor Building - The Reactor Building is defined in this context as the area located between the drywell wall and the Shield Building. In general, the shielding for the Reactor Building is designed to maintain open areas at dose rates less than 1 mR/hr. Penetrations of the drywell wall are shielded to reduce radiation streaming through the penetrations. Localized dose rates outside these penetrations are limited to less than 5 mR/hr. The penetrations through interior shield walls of the Reactor Building are shielded using a lead-loaded silicone sleeve to reduce the radiation streaming area made available by the penetrations. Penetrations are also located so as to minimize the impact of radia-tion streaming into surrounding areas. O The majority of the components of the Reactor Water Cleanup System (RWCS), except the pump, are located in the Reactor Building. Both the RWCU System regenerative and nonregenerative heat exchangers are located in a shielded cubicle separated from the other components of the system. A shielded labyrinth entry provides access to the cubicle, whic h need not be entered for system operation. Process piping between the heat exchangers and the filter demineralizers is routed through shielded areas or embed-ded in concrete to reduce the dose rate in surrounding areas. The two RWCS filter demineralizers are located in separate shielded cubicles, which allows maintenance of one unit while operating the other. The dose rate in the adjoining filter demineralizer cubicle from the operating unit is less than 6 mR/hr. Entry into the filter 12.3-33 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) domineralizer cubicle, which is infrequently required, is via a stepped shield plug at the top of the cubicle. The bulk of the piping and valves for the filter dominer-alizers is located in an adjacent shielded valve gallery. Backflushing and resin application of the filter domin-eralizers are controlled from an open corrodor, where dose rates are less than 1 mR/hr. The RWCS backwash receiving tank is also separately shielded from the other components of the RWCS, including the tank dis-charge pump, which allows maintenance of the pump without direct exposure to the spent resins contained in the backwash tank. A shielded labyrinth entry to the back-wash tank cubicle reduces the dose rates at the entry to less than 1 mR/hr. The fuel storage pool is shielded to maintain accessible areas around the pool at less than 1 mR/hr. Transfer of fuel from the Reactor Building to the Fuel Building is through an inclined fuel-transfer tube. A reinforced concrete structure surrounding the tube reduces radiation levels to less than 5 mR/hr during fuel transfer. Access to the fuel transfer tube is through a hatch shielded b) a stepped, composite concrete and lead shield plug. The main steamline, RWCS piping and RHR piping are routed l through the shielded steam tunnel. The steam tunnel l walls are designed to reduce the dose rates to less than 1 mR/hr in areas along the sides of the steam tunnel and less than 5 mR/hr both above and below the steam tunnel. Shielding of the Traversing Incore Probe (TIp) modules is provided by wing walls and a stepped, shielded floor at the (+)l1 ft 0 in. elevation above. During normal 12.3-34

CESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 12.3.2.3 Plant Shielding Description (Continued) operation, radiation levels near the modules are less than 5 mR/hr due to radiation streaming through the TIP drywell wall penetrations. In the unlikely event of an inadvertent and complete TIP retraction into the TIP module, a radiation monitor attached to the interior of the wing wall alarms to initiate egress from the TIP area. The shield walls previously mentioned limit the egress exposure to less than 100 mR. The Shield Building, which encloses the steel contain-ment, is in general a 3-ft thick reinforced concrete structure. As a radiation shield, the Shield Building is designed to reduce the radiation levels on the plant site and at the site boundary to 10CFR100 limits following a LOCA, during which the containment and shield annulus () contain a large inventory of radioisotopes released by the core. Radiation levels in open areas just outside the Shield Building during normal reactor operation are less than 1 mR/hr. (3) Auxiliary Building - The ECCS systems are located in separately shielded cubicles in the Auxiliary Building. Shield labyrinths are provided to gain entry into the cubicles, and equipment removal doors at the (-)32 ft 0 in. elevation are shielded with removable horizontally and vertically lapped concrete block. A reinforced con-creta monolith structure caps the roof hatch used for RHR o heat exchanger removal. Piping to and from the ECCS System is routed through shielded pipe chases. Access into the cubicles is not required to operate the systems. In general, the radiation levels in the open corridors of the Auxiliary Building are less than 1 mR/hr, except 12.3-35

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) g during RHR shutdown cooling mode operation, when radiation levels may temporarily range between 1 and 5 mR/hr in areas near the RHR cubicles. The RWCS pumps are located in a shielded cubicle designed to reduce the radiation levels in the adjoining open cor-ridor to less than 1 mR/ hr. The pumps are separated by shield walls to allow operation of one of the pumps while performing maintenance on the other. Dose rates at this pump due to the operating pump and piping are less than 5 mR/hr. A shielded valve gallery is employed to permit manual operation of the valves asscciated with the RWCU pumps without entering the pump arca. Piping for the pumps is directly routed from the steam tunnel to the RWCU pump area. The CRD maintenance room walls are designed to reduce dose rates in the adjoining corridor to less than 1 mR/hr during all CRD maintenance operations except CRD transfer, when doese rates in the corridor temporarily range between 1 and 5 mR/hr. The concrete CRD storagt vault located in the CRD Maintenance Room, which is used for storing clean and dirty CRD parts as well as assembled units, is designed to reduce dose rates in the room to less than 5 mR/hr. The main steamlines, routed through the Auxiliary Build-ing, are located in the' shielded steam tunucl. The steam tunnel reduces the dose rates from the steam lines to less than 1 mR/hr in all adjoining areas except the roof of the steam tunnel, which is less than 5 mR/hr. O 12.3-36

h GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 4 12.3.2.3 Plant Shielding Description (Continued) O (4) Fuel' Building - The 5-ft thick walls of the fuel storage pool reduce the dose rates in adjacent open corridors to less than 1 mR/ hr. The remaining pools are similarly shid1ded. During fuel transfer operations, sufficient water remains in the pools to maintain dose rates on the operating floor at less than 1 mR/hr. The decontamina-tion waste holding tank, which stores waste water from the fuel cask washdown operations, is' shielded to less than 1 mR/hr in the adjacent open corridor. i 'The components of the Fuel Pool Cooling and Cleanup (FPCC) System, employed for removing contaminants and cooling the fuel pool water, are located in the Fuel Building. Separately shielded cubicles house the two FPCC filter demineralizers to allow maintenance of one fi1ter domineralizer while operating the other. A stepped shield plug is employed to shield the access hatch because of the infrequent maintenance requirements of the filter demineralizers. The backwash receiving tank and the tank discharge pump are also separately shielded. Due to the relatively low activity levels of the fuel pool water, no special shielding is required for the FPCC circulation pumps, heat exchangers or drain tank. The suppression pool cleanup (SPCU) demineralizer is located in the Fuel Building and is shielded to maintain dose rates during normal reactor operation in adjacent open areas at less than 1 mR/hr. The spent resin trans-for piping, which is routed to the Radwaste Building, is embedded in the concrete floor for shielding. The SPCU circulation pumps are located in an open corridor at the (-) 32 f t 0 in, elevation. During normal O. 12.3-37

l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) operation, dose rates in the pump area are less than 1 mR/hr. During an isolation transient, however, dose rates in the area temporarily increase to 700 mR/hr. Due to the nature of the event, egress from the area can be accomplished well before dose rates reach this level. Access to equipment in this area is not required during this occurrence. The two redundant SGTS filter units are located in separately shielded cubicles in the Fuel Building. Shielding of the fan from the filter unit provides a lower radiation area for fan maintenance. Operation of the SGTS does not require entry into the SGTS filter area. (5) Control Room - The dose rate in the control room is much less than 1 mR/hr during normal reactor operating con-ditions. The outer walls of the Control Building are designed to attenaate radiation from radioactive materials contained within the Shield Building and from possible airborne radiation surrounding the Control Building following a LOCA. The walls provide sufficient shielding to limit the direct-shine exposure of control room per-sonnel tollowing a LOCA to a fraction of the 5-Rem limit as is required by 10CFR50, Appendix A, Criterion 19. Shielding for the outdoor air cleanup filters is also provided to allow temporary access to the mechanical equipment area of the Control Building following a LOCA, should it be required. (6) Radwaste Building - Shielding for the Radwiste Building is designed to limit radiation levels in the oren cor-ridors, control room and HVAC and Electrical Equipment 12.3-38

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.2.3 Plant Shielding Description (Continued) Rooms to 1 mR/hr. Radioactive piping is routed through pipe chases and shielded cubicles in order to avoid com-promising these areas. Highly radioactive components, such as the spent resin tank, the cleanup phase separators, the two sets of waste evaporators and heat-ing elements, and the waste filters are shielded separ-ately to reduce the exposure received during maintenance of adjacent equipment. Similarly, tank discharge pumps for these highly radioactive systems are also shielded from the components. Shield labyrinths are provided for all cubicles containing high-activity components to reduce the radiation streaming into other adjoining areas without restricting the accessibility of these cubicles. ( Waste barrel filling operations are conducted in a shielded cubicle. Control of these operations is from areas where radiation levels are less than 1 mR/hr. The filled waste barrels are stored in a shielded area at the (-)6 ft 10 in. elevation adjacent to the truck loading area. The truck loading area radiation level is less than 1 mR/hr for all periods except during the transfer of a waste barrel from the filling station into the storage area. (7) Turbine Building - (Applicant to supply.) (8) General Plant Areas - (Applicant to supply.) 12.3.3 Ventilation The HVAC systems for the various buildings in the plant are dis- /~ cussed in Section 9.4, including the design bases, system V) descriptions, and' evaluations with regard to the heating, cooling, 12.3-39

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.3 Ventilation (Continued) and ventilating capabilities of the systems. This section discusses the radiation control aspects of the HVAC systems. 12.3.3.1 Design Objectives The following design objectives apply to all building ventila-tion systems: (1) The systems shall be designed to make airborne radia-tion exposures to plant personnel and releases to the environment ALARA. To achieve this objective, the guidance provided in Regulatory Guide 8.8 shall be fallowed. (2) The concentration of radionuclides in the air in areas accessible to personnel for normal plant surveillance and maintenance shall be kept below the limits of 10CFR20 during normal power operation. 12.3.3.2 Design Description In the following sections, the design features of the various ventilation systems that achieve the radiation control design objectives are discussed. For all areas poter.tially having air-borne radioactivity, the ventilation syctems are designed such that during normal and maintenance operations, airflow between areas is always from an area of low potential contamination to an area of higher potential contamination. (Details of the air cleaning unit design are to be provided by the Applicant.) 12.3.3.2.1 Control Room Ventilation The Control Building atmosphere is maintained at a slightly posi-tive pressure (up to 0.5 in, wg) at all times, except if 12.3-40

~. GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 12.3.3.2.1 Control Room Ventilation (Continued) i exhausting or isolation are required, in order to prevent infiltra-l tion of contaminants. Fresh air is taken in via a dual inlet sys-l tem, which has an intake structure on the roof of the building and i an alternate intake structure on the side of the Auxiliary Build-ing. The inlets are arranged with respect to the SGTS exhaust stack such that a high probability exists that at least one of the l intakes is free of contamination after.a LOCA in the unit that the Control Building serves. Both inlets, however, can be submerged in contaminated air from a LOCA in an adjacent unit, but the calculated dose in the Control Room from such an eventuality is i still below the limit of Criterion 19 of 10CFR50, Appendix A. Outside air coming into the intakes is normally filtered by a particulate filter and a high-efficiency filter. If a high radia-I tion level in the air is detected by the airborne radiation moni-toring sytem, flow is automatically diverted to another filter O train (an outdoor air cleanup unit, that has: i '1 (1) a particulate filter; (2) a HEPA filter; (3) a charcoal filter; (4) another HEPA filter; and (5) an electric heater. J Two redundant, divisionally separated radiation monitors and filter l trains are provided (See Subsection 9.4.1 for detailed description i of the design.). Conservative calculations show that the filters keep the dose in the Control Room from a LOCA below the limits of Criterion 19 of 10CFR50, Appendix A. i The outdoor air cleanup units are located in individual, closed , () rooms that help prevent the spread of any radiation during 12.3-41

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.3.2.1 Control Room Ventilation (Continued) maintenance. Adequate space is provided for maintenance activities. The particulate and ilEPA filters can be bagged when being removed from the unit. Before removing the charcoal, any radioactivity is allowed to decay to minimal levels, and is then removed through a connection in the bottom of the filter by a pneumatic transfer system. Air used in the transfer system goes through a llEPA filter before being exhausted. Face masks can be worn during maintenance activities, if desired. 12.3.3.2.2 Drywell Access into the drywell is not permitted during normal operation. The ventilation system inside merely circulates, without filtering, the air. The only airflow out of the drywell. into accessible areas is minor leakage through the wall. During maintenance, the drywell air is purged before access is allowed. In addition, purified breathing air is supplied via a piped distribution system and quick-disconnect fittings to the area under the reactor pressure vessel. 12.3.3.2.3 Con tai nment The containment air is continually circulated at the rate of 84,000 cfm by a forced air circulation system. The air is mixed in almost all areas, thus preventing localized activity concentration build-ups or stratification. Extra fans are provided for the dome, RWCS heat exchanger room and RWCS valve room to supply thorough mixing and activity dilution in those areas. The containment air is continually purged to the outside air during normal operation by a 5,000 cfn airflow. Fresh air is drawn from O 12.3-42

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O {J) 12.3.3.2.3 Containment (Continued) outside environs, and exhaust air is drawn from the general con-tainment and from the RWCS Valve and Heat Exchanger Rooms. Radia-tion monitors continually monitor the containment exhaust air, and if high radioactivity is detected, the containment isolation valves are closed and the supply and exhaust fans are stopped to prevent any significant radioactive release. The containment can then be purged through the Standby Gas Treatment System. During normal operation, the containment air is mai.ntained at a slightly negative pressure with respect to the oubside atmosphere to prevent unmonitored venting or leakage. Dedicated airflows are provided at sample stations to direct any airborne radiation away from personnel. A curtain of air from the dome area is provided at the upper and lower air locks to reduce fs airborne contamination to personnel preparing to go through the \\- air lock during normal or abnormal conditions. During refueling, a high containment purge rate (25,000 cfm) using fresh air is established to keep airborne radiation in containment minimal. Most of the additional airflow is directed over the upper pools to exchaust vents nearby to minimize the spread of evolution from the pools. The exhaust flow is monitored for radioactivity, and if high activity is detected, the supply and exhaust valves close, and the supply and exhaust fans stop. The containment ventilation system is described in detail in Subsection 9.4.5. 12.3.3.2.4 Shield Annulus Exhaust System Air in the Shield Building annulus is exhausted and/or recirculated, as required, to maintain the pressure at (-) 5 in, wg. This V(N ensures that any leakage is into the building and not out of it. 12.3-43

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.3.2.4 Shield Annulus Exhaust System (Continued) The maximum exhaust flow rate is 1,500 cfm. The exhaust is con-tinually monitored for radiation. If a high radiation level is detected, the dampers to the plant vent automatically close, and the flow is directed to the SGTS. The annulus air pressure remains at (-)5 in, wg. 12.3.3.2.5 Auxiliary Building The Auxiliary Building HVAC System is divided into three zones, which are separated by leaktight, physical barriers. The zones include: (1) ECCS, RCIC and RWCS equipment rooms (These rooms contain equipment that is a potential source of radioactivity. If a leak occurs, the other accessible areas of the building are not contaminated.); (2) electrical equipment area, cable tunnels, cable spread-ing rooms, CRD maintenance room and remote control panel area, and the heating and ventilating equipment rooms; and (3) steam tunnel (This room also contains a potential source of radioactive material leakage.) Air pressure in the rooms in Zone 1 is maintained slightly below outside atmospheric pressure by a fresh air supply and exhaust system. The supply air is filtered by a particulate filter and a high-efficiency filter. The air is distributed to the corridor and then flows into the various rooms by infiltration. The exhaust stream is monitored for radioactivity, and if a high activity level is detected, the exhaust stream is diverted to the SGTS. 12.3-44

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O {v'h 12.3.3.2.5 Auxiliary Building (Continued) ) Normally, exhaust air is drawn from the corridor and various rooms. The exhaust duct in the corridor has two isolation valves in series and a radiation monitor. The valves isolate the corridor area from the rest of the system if high airborne radioactivity is detected by the radiation monitor. The area above the (-)6 ft 10 in. level of the Auxiliary Building, except for the CRD maintenance area, is maintained at a positive pressure during normal operation. The CRD maintenance area is maintained at a negative pressure and exhausted to atmosphere during normal operation. In the CRD maintenance room, a hood exhaust system draws con-ditioned air across the face of the cleaning tanks. The air is drawn through ductwork, through particulate filters and high-f') efficiency filters and discharged to the atmospher'e by the room exhaust fan. Also, a breathing air system is provided for personnel working in the room and wearing respiratory masks, helmets or suits. For a description of the Auxiliary Building HVAC System, see Subsection 9.4.3. 12.3.3.2.6 Fuel Building The areas within the Fuel Building are maintained at a negative static pressure, except when the railroad doors are open. One air-conditioning system is provided for the area above the operating floor, and a separate air conditioning system is pro-l l vided for the area below the operating floor. Exhaust air is drawn from areas that might be contaminated and from airflow across n~_- 12.3-45

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.3.2.6 Fuel Building (Continued) the pool. A radiation monitor is located on the exhaust flow line. If a high radiation level is detected, intake and exhaust flows are terminated, and the SGTS is started. The filter units for the SGTS are located in the Fuel Building. The design of the SGTS meets the requirements of 10CFR50, Appendix A and the guidelines of Regulatory Guide 1.52. Two fully redundant units are provided. The units are located in individual, closed rooms that prevent the spread of radioactivity during maintenance. See Subsection 6.5.1 for a detailed description of the SGTS. A radiation monitor installed in the exhaust ductwork alarms on indication of high radiation. This signal isolates the building and diverts the exhaust to the SGTS, as in the Auxiliary Building ECCS, RCIC and RWCS equipment rooms. Adequate space has been provided in the filter unit rooms for easy maintenance. Control panels for the units are located outside the rooms in a low-radiation area. The particulate filters and HEPA filters can be bagged when being removed from the unit. Before rcmoving the charcoal, any radioactivity is allowed to decay to minimal levels, and is then removed through a connection in the bottom of the filter by a pneumatic transfer system. Air used in the transfer sy3 tem goes through a IIEPA filter before being exhausted. 12.3.3.2.7 Radwaste Building The Radwaste Building is divided into two zones for ventilation purposes. The control room and unit substation are one zone, and the remainder of the building is the other zone. The air pressure 12.3-46

1 5 GLSSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 12.3.3.2.7 Radwaste Building (Continued) in'the first zone is maintained slightly above atmospheric, while j the air pressure in the second zone is maintained slightly below atmospheric. Air in the second zone is drawn from outside the building and distributed to various work areas within the building. Air flows from the work areas to the tank and pump rooms and is [ then discharged via the radwaste vent. An alarm sounds in the l control room if the exhaust fan fails. The exhaust flow is j monitored for radioactivity, and if a high activity level is detected, the potentially radioactive cells are automatically isolated, but airflow through the work areas continues, i f If the vent stack high-radiation alarm continues to annunciate after the tank and pump rooms are isolated, the work area branch exhaust ducts are selectively manually isolated to locate the I involved building area. Should this technique fail, because the airborno radiation has spread throughout the building, the control l room air conditioning continues, but the ' air conditioning for the balance of the building is shut down. i The silo, waste tilter room, oil separator room, and the mixing j and filling station exhaust air is drawn through a filter unit I consisting of a particulate filter, a HEPA filter, a charcoal filter, and then another HEPA filter, before being discharged to the atmosphere. The air is monitored for radioactivity, and if a high level is detected, supply and. exhaust is terminated. Maintenance provisions for the filters are similar to those for the SGTS and Control Building HVAC System. See Subsection 9.4.6 for a detailed discussion of the Radwaste Building HVAC System. i l 12.3-47

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 12.3.3.2.8 Diesel-Generator Buildings The Diesel-Generator Buildings do not contain sources of radio-activity, nor is access required during an event involving activity release from another building. Therefore, no radiation protection measures have been taken in these buildings. 12.3.3.2.9 Turbine Building Environmental System (Applicant to supply.) 12.3.3.2.10 Central Service Facility (Applicant to supply.) 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.4.1 Control Room Ventilation Radiation Monitoring System This system monitors the radiation level exterior to the inlet ducting of the control room ventilation system. The system con-sists of four channels identical to the channels in the contain-ment and drywell ventilation radiation monitoring system. The recorder is powered from 125 VDC bus B. Two-out-of-two upscale (high/high)/ inoperative trips in channels A and C initiate shutdown and inboard isolation valve closure of the control room ventilation system and initiate startup of the emergency air filtration fan (unit A). The same condition for channels B and D initiates shutdown and outboard isolation valve closure of the control room ventilation system and initiates startup of the emergency air filtration fan (unit B). (Additional information is to be supplied by the Applicant.) O 12.3-48

GESSAR II 22A7007 4 i 238 NUCLEAR ISLAND Rev. 0 ( 12.3.5 References i i 1. N. M. Schaeffer, " Reactor Shielding for Nuclear Engineers", TID-25951, U. S. Atomic Energy Commission (1973). 2. J. H. Hubbell, " Photon Cross Sections, Attenuation Coefficients, and Energy Absorption Coefficients from 10 KcV to 100 GeV", NSRDS-NBS29, U. S. Department of Commerce, August 1969. 3. " Radiological Health Handbook", U.S. Department of Health, Education, and Welfare, Revised edition, January 1970. 4. " Reactor Handbook',',' Volume III, Part B, E. P.

Blizzard, U.

S. Atomic Energy Commission (1962). 3 [ 5. Lederer, Hollander and Perlman, " Table of Isotopes", l Sixth Edition, (1968). l 1 6. M. A. Capo, " Polynomial Approximation of Gamma Ray Buildup Factors for a Point Isotropic Source", APEX- ) 510, November 1958. () 7. " Reactor Physics Constants", Second Edition, ANL-5800, U.S. Atomic Energy Commission, July 1963. 8. ENDF/B-III and ENDF/B-IV Cross Section Libraries, Brookhaven National Laboratory. 9. PDS-31 Cross Section Library, Oak Ridge National j Laboratory. 10. DLC-7, ENDF/B Photo Interaction Library. b !lO 12.3-49/12.3-50

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i Table 12.3-1 COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONS Computer Code Description 4 RIDST* Program for calculating radioisotope decay source terms (cumulated gamma yields in selected energy groups) for various decay times based on the input of initial isotopic activities. This pro-i gram is a condensed version of ISOSHLD-IIT.. QADMOD* A multigroup, multiregion, point kernel, gamma ray code for calculating the flux and dose rate at discrete locctions within a complex source-9 geometry configuration. () GGG* A multigroup, multiregion, point kernel code for calculating the contribution due to gamma ray scattering in a heterogeneous three-dimensional 1 space. ANISN* Multigroup, multiregion code for calculating the neutron and gamma flux for a one-dimensional, j cartesian, cylindrical or spherical geometry arrangement. 1 i i COHORT

• A multigroup, three-dimensional, Monte Carlo neutron and gamma ray transport code.

Additional codes to be added by Applicant.

• These computer programs meet the quality assurance standards for computer program certification described in Chapter 17 of this O

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i \\ / A j / / \\ l:n:r s \\ r. \\ i .s t s.ve.e,. --- - ,-.m \\l g - (.PUCIG#._-... ..-4 \\ Eh.- \\ --^ 1 ,... -r' (:] ..,i,, \\ a >= vu (_ \\ , 'i'i y\\ \\ Y ......T \\ / \\ / \\ / \\ \\ ~ ,l 8-..~ / \\ I /. \\l s. ..x........... l / k I Lc i i l

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y 10, o. & m...- ~ g r,,.ro.., Figure 12.3-4. Reactor, Aux & Fuel Bldg i' Radiation Zone Map Plan at El 28'6" d 12.3-59/12.3-60

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O 9

• UPPER

"'"d OL iIG 1/ 3 IS AL-( 1 m,... POOL e \\ f ?!,=' ll0 " VE S TiaijL t i

GESSAR II L 238 NUCLEAR ISLAND 22A7007 I Rev. O j F"

770*

ORf AELL HE AD STORAGE ~ g-= y ~% p / x x \\ (// \\ a CONT AIN ME NT 'g ( / g ,/ N f-- SMitLD BulLDING s N i ~ STAIRS N -rWi{g___f t \\ \\ ,/ \\.6 ) a.cu RtGEN.NO ,= j = AaEA 4 H A .h g.: b ff. &yY:... -.iT.h...- -2.f_ o' . x -. e -r/- r N W ~ m N g lq lN I $EPAR ATOR STORAGE ARE A i l I % '\\ / REFUELING PL ATFORM s s 's / / \\ 1 f ,/ AVN ILI ARY PL ATFORM R AIL REFUELING RL ATFORM RAIL siAnn. s \\g )/. : ) / d b y m-p* m,"IG?nli"o' a / n n== \\ 'EL_/ 'N. i r\\,/ w-

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GESSAR II I 238 NUCLEAR ISLAND 22A7007 Rev. 0 4 S

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I, d m f sG 12 313 I FIG 12 310 SM ELD 90'LDING e STEEL CCATA14WEss? E<. a r ..G,,,,, \\ r ..G,,,, =, - -l t FtG 12 3 9 h~ v ~ ( W FKi 12 3a -.t tou'8 RE - "' escas tE \\,

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c l l I i f, a r i C l t t i AIR INTAttE-e ["' STRUCTURE e J w $4G 12.3 Il

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• ..;. 2ONE C 5-20 mRom/hr

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i F' F F' SPENT REStN - SPf MT RE$iN TANE PUMP = \\ .s

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1 c. B [-- y !/~ ' ma % f f OE as t h O'a% fit L a f f DE WA tt tet s n tCg i _f -,....., e45 f f FILTt p$ GA Pi i _.. _ _ _. _m._,,} 3 g g 574145 A. fig 12 3 21 A f ~S L: { g u v m.< 1 .,g b c ev,

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