Text

Abb System 80+ Design Control Document - Volume 15
	ML22112A078
Person / Time
Site:	LaSalle, 05200002
Issue date:	01/31/1997
From:	; ABB Combustion Engineering
To:	; Office of Nuclear Reactor Regulation
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ML20148A597	List: LD-97-014, Transmits Final Version of Sys 80+ Design Control Document (DCD); LD-97-015, Informs That ABB-CE Has Completed Review of Effective Page Listing & Sys 80+ Design Control Document Revs Dtd Jan 1997; ML20148A821; ML22112A037; ML22112A038; ML22112A039; ML22112A042; ML22112A043; ML22112A044; ML22112A045; ML22112A046; ML22112A048; ML22112A049; ML22112A050; ML22112A051; ML22112A052; ML22112A075; ML22112A076; ML22112A077; ML22112A078; ML22112A079; ML22112A081; ML22112A085; ML22112A086; ML22112A088;
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{{#Wiki_filter:.-. -.- _.-- - -. ... .-..-__. - .. --_- --. 4 l I !O 1 i the l System 80+ standardplant i i I i i i l Design Control Document l iO l Volume 15 i a o Combustion Engineering, Inc. a unun 9%l FEE

9 Copyright C 1997 Combustion Engineering, Inc., All Rights Reserved. Warning, Legal Notice and Disclaimer of Liability The design, engineering and other information contained in this document have been prepared by or for Combustion Engineering, Inc. in connection with its application to the United States Nuclear Regulatory Commission (US NRC) for design certification of the System 80+ nuclear plant design pursuant to Title 10, Code of Federal Regulations Part 52. No use of any such information is authorized by Combustion Engineering, Inc. except for use by the US NRC and its contractors in connection with review and approval of such application. Combustion Engineering, Inc. hereby disclaims all responsibility and liability in connection with unauthorized use of such information. Neither Combustion Engineering, Inc. nor any other person or entity makes any warranty or representation to any person or entity (other than the US NRC in connection with its review of Combustion Engineering's application) concerning such information or its use, except to the extent an express warranty is made by Combustion Engineering, Inc. to W, its custo 2. in a written contract for the sale of the goods or services described i'l this document. Potential users are hereby wamed that any such information may be l unsuitt,,ble for use except in connection with the performance of such a written contract l by Combustion Engineering, Inc. l l Such information or its use are subject to copyright, patent, trademark or other rights of Combu. tion Engineering, Inc, or of others, and no license is granted with respect to such rights, 9xcept that the US NRC is authorized to make such copies as are necessary for the use of the US NRC and its contractors in connection with the Combustion Engineering, Inc. application for design certification. Publication, distribution or sale of this document does not constitute the performance of engineering or other professional services and does not create or establish any duty of care towards any recipient (other than the US NRC in connection with its review of Combustion Engineering's spplication) or towards any person affected by this document. For information address: Combustion Engineering, Inc., Nuclear Systems Licensing, 2000 Day Hill Road, Windsor, Connecticut 06095

1 Systzm 80+ orsign contr:I Document J r Introduction Certified Design Material

 ,              1.0        Introduction 2.0        System and Structure ITAAC                                                               !

3.0 Non-System ITAAC 4.0 Interface Requirements 5.0 Site Parameters i Approved Design Material - Design & Analysis l 1.0 General Plant Description 1

2.0 Site Characteristics !

3.0 Design of Systems, Structures & Components 4.0 Reactor : 4 5.0 RCS and Connected Systems l 6.0 Engineered Safety Features : . 7.0 Instrumentation and Control 8.0 Electric Power > 9.0 Auxiliary Systems I 10.0 Steam and Power Conversion 11.0 Radioactive Waste Management l' 12.0 Radiation Protection . 13.0 Conduct of Operations 14.0 Initial Test Program i 15.0 Accident Analyses 16.0 Technical Specifications 17.0 Quality Assurance 18.0 Human Factors 19.0 Probabilistic Risk Assessment i 20.0 Unresolved and Generic Safety Issues Approved Design Material - Emergency Operations Guidelines 1.0 Introduction 2.0 Standard Post-Trip Actions 3.0 Diagnostic Actions 4.0 Reactor Trip Recovery 5.0 Loss of Coolant Accident Recovery 6.0 Steam Generator Tube Rupture Recovery

7.0 Excess Steam Demand Event Recovery 8.0 Loss of All Feedwater Recovery 9.0 Loss of Offsite Power Recovery i 10.0 Station Blackout Recovery 11.0 Functional Recovery Guideline f3

. U Contents

!O i the i System 80+ standardplant i i \ 1 \ s i lb l V Approved Design Material Design & Analysis l J d b i f i O Combustion Engineering, Inc. a stit M EFEF

System 80+ os, contrat oocument iO Effective Page Listing Chapter 12 1 Pages Date l i, ii - 1/97 l iii 11/% i iv, v Original i i

.             12.1-1 through 12.1-6                            Original l              12.1-7, 12.1-8                                         11/%

12.2-1 through 12.211 Original 12.2-12, 12.2-13 11/% i 12.2-14 through 12.2-28 Original ! 12.2-29 11/% 12.2-30 Original 12.2-31 11/% 12.2-32 Original 12.2-33, 12.2-34 11/% 12.2-35 through 12.2-44 Original l 1 12.3-1 through 12.3-14 Original 12.3-15 1/97 12.3-16 through 12.3-87 Original 12.4-1 through 12.4-6 Original 12.4-7 2/95 ' 12.4-8 through 12.4-10 Original l 12.5-1 Original + 1 1 i O O Approwd Design Motoria!- Me6ebion Protection (1/97) PopeI,R

i l Syst:m 80 + De'ign CrntrolDocument O ca a'er 12 co te"*= ! Page 12.0 Radiation Protection .......... ..... .. .... ..... 12.1-1 ' 1 12.1 Ensuring That Occupational Radiation Exposures are As Low As l Reasonably Achievable . . . ...... . .. ..... ... . 12.1-1

l 12.1.1 Policy Considerations .... .. .. ...... ... ... . . . 12.1-1 ' 12.1.2 Design Considerations . . . . . ...... ..... ....... . .. .. 12.1-2 1 12.1.3 Operational Considerations . .... .. . ..... ... .. ... 12.1-5 l 12.2 Radiation Sources .... ..... .... .... . ... . ....... 12.2-1 l 12.2.1 Contained Sources ..... ........ . . .. . . . ... .. .. 12.2-1 l 12.2.2 Airborne Radioactive Material Sources .. .. ... . . . . . . . . . . . . . . 12. 2-9 12.2.3 Sources Used in NUREG-0737 Post-Accident Shielding Review ..... 12.2-12 12.3 Radiation Protection Design Features ..... ... .. .......... 12.3-1 j 12.3.1 Facility Design Features . . . . .... ..... .. .. . . . . . . 12.3-1 1 12.3.2 Shielding . . . .. .. .. . . ..... ........ . .. 12.3-13 12.3.3 Ventilation . . . ........... . ..... ...... . . 12.3-16 1 12.3.4 Area Radiation and Airborne Radiation Monitoring Instrumentation . ... .. 12.3-16 O 12.4 Dose Assessment 12.4-1 O 12.4.1 Methodology . . . ... ..... .. .. .. ....... . . .... . 12.4-1 12.4.2 Industry Average Occupational Exposure for 1988 .... ..,........ 12.4-1 12.4.3 PWR Reference Plant Data ' ................ . ... . -. . . . 12.4-2 l 12.4.4 System 80+ ALARA Dose Reduction Features .. .... ......... .. 12.4-2 ! 12.4.5 System 80+ Estimated Dose Assessment ...... ............... .. 12.4-5 l 12.4.6 Occupational Exposure Breakdown Among Workforce . . . . . . . . . . . . . . . . 12.4-6 12.4.7 System 80+ Individuals' Exposure Distribution . . . . . . . . . . . .... .... 12.4-6 12.5 Health Physics Program ....... ......... . ........... . 12.5-1 l Chapter 12 Tables Page l 12.2-1 Maximum Neutron Spectra Outside Reactor Vessel ... . .... . 12.2-14 12.2-2 Maximum Gamma Spectra Outside Reactor Vessel . ....... . .... 12.2-15 12.2-3 Shutdown Gamma Spectra Outside Reactor Vessel ...... . .. .... 12.2-16 12.2-4 Basis for Reactor Coolant Fission Product Activities . . . . .. . ... 12.2-17 12.2-5 Reactor Coolant Equilibrium Concentration . . . .. ... .... ., 12.2-18 12.2-6 Average Reactor Coolant Crud Activity . . . . .... . ..... . . 12.2-20 12.2-7 N-16 Activity . . . . ............... . . . . ... 12.3-20 12.2-8 Spent Fuel Gamma Source .. . .. ..... .. ....... .. 12.2-21 g 12.2-9 CVCS Heat Exchanger Inventories . . .. ... .. . . ...... 12.2-22 12.2-10 CVCS lon Exchanger and Evaporator in.entories . ... .. ... .... 12.2-25 Approved Design Meterial . Radiation Protection (11/96) Page lii

l l Srt m 80+ 0: sign centrot 0 cument Chapter 12 Tables (Cont'd.) g Page 12.2-11 CVCS Filter and Gas Stripper Inventories . .. .. . . 12.2-27 12.2-12 CVCS Tank Inventories . . . . . . .. 12.2-28 12.2-13 Shutdown Cooling System (SCS) Specific Source Strengths . . . .. . 12.2-30 , 12.2-14 Spent Fuel Pool Related Sources Radionuclide Specific Activities . . 12.2-31 l 12.2-15 Turbine Building Sources Radionuclide Specific Activities .. . 12.2-33 12.2-16 Nuclear Annex Sources Process Gas Specific Activities . 12.2-35 12.2-17 Radwaste Building Sources Liquid Waste Tank Specific Activities .. . . 12.2-36 12.2-18 Radwaste Building Sources Liquid Waste Process Equipment . . 12.2-38 12.2-19 Radwaste Building Sources Solid Waste Process Equipment . 12.2-40 12.2-20 NUREG-1465 Post-accident Shielding Source Terms . . . 12.2-42 l 12.3-1 Normal Operation Accessibility Zone Designations . .. . 12.3-18 ; 12.3-2 Normal Operation Radiation Zones . . ., . 12.3-19 ' 12.3-3 Post-accident Accessibility Zone Designations . . . 12.3-30 12.3-4 Post-accident Radiation Zones . . . . .. .. .. 12.3-31 12.3-5 Area Monitor Locations .. .. .. .. 12.3-43 12.3-6 List of Accessible Areas Potentially > 100 R/HR . . . 12.3-46 12.4-1 PWR Reference Plant Data .. .. . . 12.4-7 l 12.4-2 Dose Distribution by Work Group . . . .. . . ... 12.4-8 l 12.4-3 Individual Exposure Distribution for Reference Plants . . 12.4-9 l 12.4-4 System 80+ Estimated Annual Occupational Exposure . . .. ... 12.4-9 l 12.4-5 System 80+ Annual Expeeure Breakdown Among Workers .. 12.4-10 Chapter 12 Figures Page 12.2-1 Maximum Spent Fuel Assembly Dose Rates Versus Axial Distance in Refueling Pool . .. . . .. . . .. .. . . 12.2-43 12.2-2 Maximum Spent Fuel Assembly Dose Rates Versus Radial Distance in Refueling Pool . . .... . . . .. . . 12.2-44 12.3-1 Radiation Zones Normal Operating Status and Shutdown Plan at El. 50+0 12.3-47 12.3-2 Radiation Zones Normal Operating Status and Shutdown Plan at El. 70+0 12.3-49 12.3-3 Radiation Zones Normal Operating Status and Shutdown Plan at El. 81+0 12.3-51 12.3-4 Radiation Zones Normal Operating Status and Shutdown Plan at El. 91+9 . . 12.3-53 12.3-5 Radiation Zones Normal Operating Status and Shutdown Plan at El.115+6 . 12.3-55 12.3-6 Radiation Zones Normal Operating Status and Shutdown Plan at El.130+6 .. 12.3-57 12.3-7 Radiation Zones Normal Operating Status and Shutdown Plan at El.146+0 .. 12.3-59 12.3-8 Radiation Zones Normal Operating Status and Shutdown Plan at El.170+0 . 12.3-61 12.3-9 Radiation Zones Post-Accident Conditions Plan at El. 50+0 . . .. 12.3-63 12.3-10 Radiation Zones Post-Accident Conditions Plan at El. 70+0 12.3-65 12.3-11 Radiation Zones Post-Accident Conditions Plan at El. 81+0 . 12.3-67 12.3-12 Radiation Zones Post-Accident Conditions Plan at El. 91+9 . . .. 12.3-69 12.3-13 Radiation Zones Post-Accident Conditions Plan at El.115+6 .. . 12.3-71 12.3-14 Radiation Zones Post-Accident Conditions Plan at El.130+6 . . . 12.3-73 Approved Design Material- Radiation Protection Pageiv

Syst m 80 + D: sign C:ntrolDocument O Chapter 12 Figures (Cont'd.) Page 12.3-15 Radiation Zones Post-Accident Conditions Plan at El.146+0 .... .. . 12.3-75 12.3-16 Radiation Zones Post-Accident Conditions Plan at El.170+0 .... 12.3-77 12.3-17 Radiation Zones During Fuel Transport Fuel Transfer Tube Details . . . . . . . . 12.3-79 12.3-18 Radiation Zones Normal Operating Status and Shutdown Radwaste Building; Plan at El. 34 + 0 & 50 + 0 . . . . . . . . . . . ... .. ..... ..... . 12.3-81 12.3-19 Radiation Zones Normal Operating Status and Shutdown Radwaste Building; Plan at El. 70+0 . . . . . ... ..... . ... .... . .... .. 12.3-83 12.3-20 Radiation Zones Normal Operating Status and Shutdown Radwaste Building; Plan at El. 91 +9 . . . . ............... .. .. .. ... . 12.3-85 12.3-21 Radiation Zones Normal Operating Status and Shutdown Radwaste Building; Plan at El. 115 + 6 . . . . . . . . . . ... .. .. ..... ...... .. 12.3-87 O a V Ancroved Design Material- Radiation Protectron Page v l

System 80+ Design ControlDocument O O 12.0 Radiation Protection 1 l I l This chapter describes the radiation protection mexures of station design and operating policies to ensure that internal and external radiation exposures to stauon personnel, contractors, and the general population due to station conditions, including normal and anticipated operational occurrences, will be within applicable guidelines, and will also be as low as is reasonably achievable (ALARA). I Radiation protection measures include: separation of radioactive components into separately shielded cubicles; use of shielding designed to adequately attenuate radiation emanating from pipes and equipment which are sources of significant ionizing radiation; use of remotely operated valves or handwheel extensions; ventilation of areas by systems designed to minimize inhalation and submersion doses; installation of permanent radiation monitoring systems; control of access to the site and to restricted areas; training of personnel in radiation protection; and development and implementation of administrative policies and procedures to maintain exposures ALARA. l 12.1 Ensuring That Occupational Radiation Exposures are As Low As Reasonably Achievable 12.1.1 Policy Considerations It is management policy to keep occupational radiation exposures to personnel ALARA. Administrative j programs and procedures, in conjunction with facility design, ensure that the occupational radiation exposures to personnel will be kept ALARA. I ; Y 12.1.1.1 Design and Construction Policies The ALARA philosophy is applied in the initial design of the plant and will be implemented via internal design reviews and documentation. These reviews will be conducted and documented consistent with the recommendations of Regulatory Guide 8.8. The plant design will be reviewed, updated, and modified as necessary during the design phase. Plant I design will be reviewed to integrate the layout, shielding, ventilation, and monitoring designs with security, access control, maintenance, in-service inspection, and radiation protection aspects to ensure that the overall design produces a plant which will achieve exposures that are ALARA. 1 Piping containing radioact ive fluids is routed as part of the engineering design effort. This ensures that lines expected to contain significant radiation sources are adequately shielded and properly routed to minimize exposure to personnel. Onsite inspections will be conducted, as necessary during construction, to ensure that the shielding and piping layout meets established criteria. During construction, visual inspections will be made to ensure that there are no major defects or scattering and streaming paths for radiation in the shield walls as they are placed. During initial power operations, radiation surveys will be conducted to ensure that the shielding meets design reauirements during normal operation and maintenance of the plant. 12.1.1.2 Operation Policies t'N kj ((The station ALARA manual will be one of the primary means of officially expressing the operational ALARA policy. This policy is demonstrated in the radiation protection program, the training program, Approved Design Material Radiation Protection Page 12.11

Syet m 80+ D~ sign c ntrolDocument and station procedures, with specific details to be covered by the owner / operator.))! The ALARA program is established in conformance with the regtiirescents of Regulatory Guides 1.8, 8.8, and 8.10 and discussed in the site-specific SAR Section 13.1. Besides describing management's commitment to ALARA, the Radiation Protection section of the Station Manual designates the station personnel who have the responsibility and authority to implement ALARA. The Radiation Protection Manager has the responsibilities for the onsite Radiation Protection Program. The responsibilities and authority of the supervisory positions for the day-to-day operation of the site radiation protection programs are discussed in the site-specific SAR Sections 13.1.2 and 13.1.3. l Prior to startup of the unit, station procedures to be used for work which involves significant personnel radiation exposure will be reviewed to verify that the procedures adhere to the ALARA philosophy. Revisions to station procedures involving significant personnel exposure will also receive an ALARA review. Systems or station modifications affecting personnel radiation exposure will also be reviewed to see that the ALARA concept is applied. In addition, periodic reviews of the ALARA program, including review of radiation exposure records and operating procedures, are conducted by offsite Radiation Protection personnel. Personnel requiring access to the restricted area and/or radiological controlled areas will receive training as necessary to permit access to these areas. These personnel will be tested to evaluate each worker's knowledge, competency, and understanding relative to the training provided. Retraining and retesting will be conducted. 12.1.2 Design Considerations This section discusses the general design methods and features by which the policy considerations of Section 12.1.1 are applied. Provisions and designs for maintaining personnel exposures ALARA are presented in detail in Sections 12.3, and 12.5. Guidance for general design features to maintain personnel exposures ALARA is provided by written System 80+= <t) ALARA guidelines which provide guidance for equipment design and selection, shielding, contamination control and corrosion product production reduction techniques. This guidance is consistent with recommendations given in Regulatory Guides 8.8 and 8.10. This guidance also incorporates lessons learned from past nuclear plant designs into the System 80+ design. General arrangement and plant layout characteristics are discussed in Section 12.1.2.1. 12.1.2.1 General Design Considerations for Shielding and ALARA Exposures General design considerations, shielding, and methods employed to maintain in-plant radiation exposures ALARA con;istent with the recommendations of Regulatory Guide 8.8, Section C.3, have two objectives:

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

I COL information item; see DCD Introduction Section 3.2. i System 80+ is a trademark of Combustion Engineering, Inc. Approved Design Material- Radiation Protection Page 12.1-2 i i

i Svitem 80+ Deslan contalDocument t Plant operating personnel are protected as necessary by shielding wherever a potential radiation hazard may exist. The shielding performs the following additional functions: e Assists in maintaining radiation exposure to plant control room personnel within the limits of 10 CFR 50, Appendix A, Criterion 19 in the unlikely event of an accident. e- Protects certain components from excessive activation or excessive radiation exposure. 1 e- Facilitates access for maintenance of components. I Maintaining occupational exposures as low as reasonably achievable (ALARA) is a major design consideration in accordance with Section C.1 of Regulatory Guide 8.8. Although the station's design is intended to result in ALARA exposures during operation, thee design features will also facilitate deccmmissioning. Regulatory Guide 8.8 also provides guidance for both equipment selection and plani layout. Equipment selection for System 80+ plays an integral part for maintaining personnel expos. ire ALARA. Criteria ; used for equipment selection are discussed in written ALARA guidelines. Synem 80+ design includes: ; 1 e EnhancaA reliability of equipment which reduces the frequency of maintenance :md the personnel exposure associated with the maintenance. This is illustrated by the following:

1. Use of reliable extended service lamping in high radiation areas, as well as, location of lighting fixtures so that maintenance can be performed in a lower radiation area. These features are in accordance with Regulatory Guide 8.8 Position C.2.i.

2. Use of ion exchangers instead of evaporators except in the Chemical Volume and Control System. The System 80+ design will minimize the use of evaporators based on industry experience. Ion exchangers are simpler in design and are more reliable than evaporators.

e Careful attention to environmental qualification of equipment. This includes equipment , qualification for a variety of environmental conditions, such as radiation, humidity, and ] temperature. This is discussed in greater detail in Section 3.11. i Electrical components, containing radiation-sensitive materials, will be shielded or located in low-radiation areas. ((The COL applicant will establish environmental qualification criteria for equipment.))3 l e Material selection of piping and components, such as valve seats, in the prunary system. Material is selected with low cobalt or nickel impurities. This minimizes the production of corrosion products which is a significant contributor to personnel exposurc received during maintenance and operational activities. ( 1 COL information item; see DCD Introduction Section 3.2. Approved Design noenriet - noneekm proseceron Pese 12.13

System 80+ Design ControlDocument l l l

Maintainability which includes plant layout and equipment spacing.

l The plant layout is designed to maintain personnel exposures ALARA during normal and post- ) accident conditions. For example, the plant layout is designed to provide. l l

1. Adequate spacing to facilitate accessibility of equipment during maintenance activities,

2. Separation of non-radioactive systems from radioactive systems into quadrants minimizing the spread of contamination,

3. Separation of system components into cubicles or compartments based on: l l

frequency of maintenance, 1 operational characteristics, and l level of radioactivity. One example of this design technique is the location of ion exchangers in pits with spent resin tanks located below the ion exchanger. This design technique reduces the personnel l exposure resulting from maintenance on other equipment in the area. l l

4. Physical location of systems, such as the Chemical Volume and Control System, that I

      ,         produce radioactive waste in close proximity to radwaste processing systems discussed         l in Sections 11.2 to 11.4. This reduces the pipe length that must be routed through            I personnel access corridors and the spread of contamination.                                   l l

S. The ventilation systems are designed so that the air flow will be from clean areas to potentially contaminated areas. This will minimize potential for the spread of airborne contamination. For areas that have a potential for a high radiation release, such as the fuel storage area, a slightly negative pressure will be maintained. Ventilation systems are discussed in further detail in Section 9.4. The station design is reviewed by the radiation protection staff to assure the input of radiation protection professionals into the final station design. Design review not only entails examining layout and piping drawings, but also inspection of various stages of station design and construction. Radiation shielding personnel are kept aware of current or anticipated radiation protection problems by periodic visits to this and other sites. These station visits provide valuable feedback for use in reviewing the final station design. A formal operational feedback program is used to identify generic problems and implement design improvements. l ALARA exposures receive further attention through the training of designers and engineers in which, pipe layout, equipment selection and placement, radiation zone information and methods of minimizing crud build up in piping are covered. In addition close work with equipment vendors results in the purchase of low maintenance equipment with material properties suitable for minimizing corrosion, tanks are designed with sloped bottoms to prevent crud buildup. Those components with the potential of exposure from crud are provided with flushing capability from either demineralized water or chemical decontamination. Also, equipment is designed that separates highly radioactive portions from lower radiation level portions of a component. O l Approved Design Material Radiation Protection Page 12.14

I System 80+ Design ControlDocument I ((The' COL' applicant will submit a detailed shielding analysis and will ensure that the final design i incorporates lessons learned from previous nuclear power plant designs.))! i i 12.1.3 Operational Considerations i 4 j Consistent with the recommendations of Regulatory Guides 8.8 and 8.10, the radiation exposure of plant personnel will be kept ALARA by means of the radiation protection program discussed in Section 12.5. l Operational ALARA policy statements are formulated and expressed at the corporate stafflevel through

the training program discussed in Section 13.2 and through the Radiation Protection manual and the
ALARA manual discussed in Sections 12.1.1 and 12.5.

l [ 4 Personnel and job exposures trends are reviewed by management at the site and offsite radiation j protection personnel, and appropriate action is taken. Summary reports of occupational exposure are ; , provided that describe problem areas where high radiation doses are encountered and that identify which 1 work group is accumulating the highest doses. Recommendations are then made for changes in operating, ! j maintenance, and inspection procedures or for modifications to the station as appropriate to reduce dose, j ' j From industry experience it has been shown that the majority of exposure at operating plants is received

during plant outages from maintenance and inspection activities and not from normal operating activities.

? This is logical since operators can normally stay outside shield walls to read instruments or operate valves and have to enter cubicles containing radioactive equipment for short periods of time only to check i equipment, whereas maintenance and inspection personnel usually must go inside cubicles or behind

shield walls and must be in close proximity to the lines, valves, instruments, or other pieces of equipment c which are radiation sources.

2 The System 80+ design incorporates lessons learned from past designs, as well as the recommendations

from Regulatory Guide 8.8 and 8.10. Design features such as:

I { 1. - Addition of platforms around the steam generator and the reactor coolant pump seal

                          . canridges, i

i

2. Provisions for sufficient spacing for equipment laydown/ pull areas, 4

. 3. Use of removable insulation, 1-l ' 4. Reduction of length of welds by use of seamless piping, and

5. Integrated reactor head removal

{ , e i ensure that personnel exposure is maintained ALARA during maintenance and inspection activities. System 80+ design features are discussed in further detail in Section 12.3 and the System 80+ ALARA l 4 Guidelines Manual. l Areas which house systems and components subject to in-service inspections that are high radiation zones 3 are designed to permit prompt ingress and egress - System 80+ is also designed with adequate spacing, i g including laydown areas, between equipment to facilitate maintenance and inspection activities. Radiation g }- 3 COL information item; see DCD Introduction Section 3.2. 4prowd Design ntesenet Redesion Protec6 err Page 12.1-5 L _ _ , _

Syst~m 80 + D sign ControlD cument Protection personnel will perform surveys of areas requiring access for maintenance and inspection activities prior to entry and will perform periodic inspections during work activities. ((The COL applicant will provide Radiation Protection Procedures to provide survey requirements for work areas.))l In addition, System 80+ provides the following design features to maintain personnel exposure ALARA during ISI: e Pipe stops, snubbers, and pipe hangers near welds that require ISI or repair are carefully positioned to facilitate weld accessibility, e Integrally forged components and seamless pipe is selected whenever possible to avoid in-service weld inspections. This reduces the length of pipe requiring in-service weld inspections.

Blanket-type thermal insulation with velcro fasteners is selected for components and ,

I piping containing radioactive sources. This insulation is easily removable by one worker for in service weld inspection. Careful attention is paid to the potential for the creation of airborne contamination. Adequate containment or ventilation is provided to minimize the potential for the spread of airborne contamination and to ensure that the average concentration of airborne contamination is less than specified in 10 CFR 20, Appendix B of Sections 20.1001-20.2401, Table 1, Column 3. Section 12.2.2.1 specifies the methodology for determining the inplant concentration of airborne contamination. Section 12.3.1.4 ! provides further details regarding airborne contamination control for System 80+. Maintenance activities that could involve significant radiation exposure of personnel are carefully planned. They utilize any previous operating experience and are carried out using well trained personnel and proper equipment. Work permits for routine and non-routine operations are issued for each job, listing l Radiation Protection requirements that will be followed by all personnel working in the radiation control area. Where applicable, specific radiation exposure reduction techniques, such as those set out in Regulatory Guide 8.8 are evaluated and used. Procedures for such radiation exposure related operations l as maintenance, in-service inspection, radwaste handling, and refueling, are well planned and developed by cognizant groups, and are reviewed by the .uation radiation protection staff to ensure that exposures will be ALARA. Careful personnel radiatiw rnd contamination monitoring are integral parts of such maintenance activities. During and upon c : pletion of major maintenance jobs, personnel radiation exposures are evaluated and assessed relative to estimated exposures so that appropriate changes can be made in techniques or procedures as soon as practicable for future jobs. Some of the ALARA techniques that should be considered to reduce these exposures are discussed below. General ALARA Tecimiques

Permanent shielding is used where possible, by having workers stay behind walls or in areas of lower radiation levels when not actively involved in work in radiation areas. On some jobs temporary shielding is used. Temporary shielding will be used only if the total exposure, which includes the exposure received during installation and removal of the shielding is reduced.

O I COL information item; see DCD Introduction Section 3.2. Approved Design Material- Radiation Protection Page 12.1-6

Syztem 80+ Drsign ControlDocument (p)

Systems and major pieces of equipment which are subject to crud buildup have been equipped U

with connections which can be used for flushing. Prior to performing maintenance work, consideration will be given to the practicality and effectiveness of flushing and/or chemically decontaminating the system or piece of equipment in order to reduce the crud levels and personnel exposure.

On complex jobs involving exceptionally high radiation levels, " dry runs" may be made, and in some cases mockups may be used to familiarize the workers with the exact operations they must perform at the jobsite. This job preplanning will include estimates of the person-rem needed to complete the job. At the completion of the job, a debriefing session will be held with the people who actually performed the work in an effort to determine if the work could have been performed in a more efficient manner resulting in less exposure. This in'ormation, together with the procedures used and actual person-rem expended, will be recorded. The radiation, contamination, and airborne activity levels determined during working conditions will also be recorded. In addition, if any external body contamination or internal contamination was encountered during the job, this information will also be recorded. This information will be used to provide guidance at the preliminary stage of future similar operations. These techniques will assist in improving worker efficiency and thus will minimize the amount of time spent in the radiation field.
As much work as practical will be performed outside the radiation areas. This includes items such as reading instruction manuals or maintenance procedures, adjusting tools, repairing valve internals, and prefabricating components.
For long term repairjobs, consideration will be given to setting up communications systems, such V(3 as sound powered telephones or closed circuit television, so that supervisory personnel can check on the progress of work from a lower radiation area.
On some jobs, special tools may be used when their use would permit the job to be performed more efficiently or would have prevented errors, thus reducing the time in the radiation field.

Special tools may also be used if their use would increase the distance from the source to the worker, thus reducing the exposure rate. Unless special tools are necessary to accomplish thejob, special tools will be used only if the total exposure, which includes that received during installation and removal is reduced.

Entry and exit points will be set up in areas so that personnel are exposed to as low as a level of radiation as practical. This will be done because personnel may spend a significant amount of time changing protective clothing and respiratory equipment in these entry-exit areas. These entry and exit points are set up to limit the spread of contamnation from the work area.
Plastic glove boxes, which can be taped around valves or other fixed components, and plastic bags are used where practical so that personnel can work on equipment without being exposed to the contamination produced during the work, and to limit the spread of contamination.
Radiation levels in work areas will be posted at the entrance to the area and/or in the work area so that the areas of highest and lowest radiation level are clearly identifiable. Individuals will be instructed to stay in the lowest radiation area consistent with performing their jobs.

,.m

Personnel will wear self-reading dosimeters for work in high radiation areas so that they can V)

( determine their accumulated exposure at any time during the job. This is in addition to their monthly TLD, legal TLD and job TLD. Approved Design Material- Radiation Protection (11/96) Page 12.1 7

Syst ~m 80 + Design crntrolorcument

On jobs with exceptionally high radiation levels, a timekeeper, who knows the exposure rate of the radiation field, will keep track of the exposure using a timing device. This tecimique will ensure that personnel are not staying in a high radiation area longer than intended. The timekeeper will remain in the lowest possible exposure area consistent with performing this task.

e Robotics will be used, whenever practical, to perform maintenance and inspection activities (e.g., remote pipe welds and inspections), particularly in potentially high radiation areas (e.g., underwater surveillance robots and remote / slave manipulators, may be used to work in high radiation areas). References for Section 12.1

1. Regulatory Guide 1.8, " Personnel Selection and Training."

2. 10 CFR Parts 19 and 20.

3. Regulatory Guide 8.8, "Information Relevant to Insuring that Occupational Radiation Exposures at Nuclear Power Station will be as low as is Reasonably Achievable."

4. Palo Verde FSAR.

5. Catawba FSAR.

6. NUREG-0737, " Clarification of TMI Action Plan Requirements."

7. Regulatory Guide 8.19 " Occupational Radiation Dose Assessment in Light Water Reactor Power Plants Design State Man-Rem Estimates."

8. Standard Review Plan, NUREG-0800, Section 12.1.

9. Duke Power Company Design Engineering Department's Departmental ALARA Guide.

10. NURE , 0761, " Radiation Protection Plans for Nuclear Power Reactor Licensees."

11. Regulatory Guide 8.10, " Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable."

O Approved Design Material- Radiation Protection i11/96) Page 12.1-8

System 80+ De-Ign Cintrol DEcument 12.2 Radiation Sources This section discusses and identifies the sources of radiation that form the basis for shield design calculations and the sources of airborne radioactivity used for the design of personnel protective measures and for dose assessment. 12.2.1 Contained Sources ; l Source terms used for shielding design are based upon full power operation with 0.25% fuel claddmg defects. Sources in the primary coolant include fission products released from fuel cladding defects, activation products as well as corrosion ptoducts. Nitrogen-16 from activation products is the primary source for shielding design for most of the reactor coolant system. Conservative allowances have been j given for both transit decay and daughter product formation. 12.2.1.1 Containment 12.2.1.1.1 Reactor Core 1 i The primary radiation emanating from the reactor core during normal operat on are neutrons and gamma l rays. Tables 12.2-1 and 12.2-2 list neutron and gamma multi-group fluxes in the reactor cavity at the I side of the reactor vessel; these tables are based on nuclear parameters discussed in Chapter 4. Table 12.2-3 lists core gamma sources after shutdown for shielding requirements during shutdown and in-service inspection. O V 12.2.1.1.2 Reactor Coolant System Sources of radiation in the reactor coolant system are fission products released from fuel and activation and corrosion products. j l Table 12.2-6 lists the average expected activities due to crud deposits on steam generator tubing and l primary system piping. l 1 The activation product nitrogen-16 is the predominant activity in the reactor coolant pumps, steam l generators, and reactor coolant piping. The N-16 activity in each of the components depends on total transit time to the component. 12.2.1.1.2.1 Design Basis Source Terms 12.2.1.1.2.1.1 Maximum Fission Product Activities in Reactor Coolant Maximum normal operation fission product activities will be used as design basis source terms for shielding and facilities design. Source terms for calculating the consequences of postulated accidents are discussed in Chapter 15. The isotopes chosen for consideration in the maximum case are those which are significant for design purposes by reason of a combination of gamma energy, half-life or abundance. The mathematical model used to determine the concentration of nuclides in the Reactor Coolant System involves a group of linear, first order differential equations. These equations are obtained by applying mass balance for production and removal from the fuel pellet region as well as the coolant region. Approved Desigm Atatorial- RecGeaion Protection Page 12.21

Sy: tem 80+ D: sign ControlD?cument In the fuel pellet region, the mass balance includes fission product production by direct fission yield, by l parent fission product decay and by neutron activation and escape from the pellet. The computer code ORIGIN is utilized to calculate the fission product core inventories. In the coolant region, fission production is by escape from the fuel through the defective fuel rod cladding and parent decay. Removal is by decay, by coolant purification, by feed and bleed operation (for fuel burnup), by leakage and other feed and bleed operations such as startup and shutdowns as well as load i follow operation. ; 1 The calculation of the isotopic inventory in the reactor coolant is governed by the following equation: dN,,i/dt = (Ri Nn)/M c + (f. i K _i)N,,;.i i - {Ai + D, + (Qi /M )[Ui + (DF-1)/DF]}N,,i e i i Where the variables are identified as: l l R = vF v = Escape Rate Coefficient (1/sec) F = Fraction of fuel cladding failed assumed to be equal to 0.0025 Nr = Fission product inventory (Ci/gm) Me = Mass of the reactor coolant (gm) A = decay constant (1/sec) D = Dilution coefficient (1/sec) l D = b/(B(0)- b t) DF(i) b = Boron reduction rate (1/sec) = [B(0) - B(t)]/t B(0) = Initial boron concentration at t=0 (ppm) t = operating time (sec) l DF = Decontamination Factor of Purification IX l Qi = Letdown flow rate (gm/sec) U = Stripping Fraction f = Fraction of parent nuclide decay events that result in the formation of the daughter nuclide O Approved Design Material- Radiation Protection Page 12.2-2

4 Sy:t*m 80+ Dehn CintrolDocument j , and where the subscripts are identified as: ' 1 l 1 = i* isotope f I-1 = precursor to the i* isotope for decay 4 Escape rate coefficients are used to represent the overall release from the fuel pellets to the gap. The ! ' escape rate coefficient is an empirical value which was derived from experiments initiated by Bettis and " run in the NRX and MTR reactors (Reference 4). The escape rate coefficients were obtained from test i rods which were operated at high linear heat rates. The linear heat rates were uniform over the test ' sections of 10.25 inches in length. The exact linear heat rates were precisely known but post-irradiation i { inspection showed that some test specimens had experienced centerline melting. Later tests were done in Canada to determine the effect of rod length on the release of fission gases and iodines from the . : defective fuel rods (Reference 5). A by-product of these experiments was the relationship between linear : heat rate and the escape rate coefficient. ;

4. ;
The average heat rate for a fuel rod is well below the values that correspond to the selected escape rate

.. coefficients for halogens and noble gases. The presently used escape rate coefficients are based on a linear heat rate of 21 kw/ft. Shown in Table 12.2-4 are the values of bounding parameters used to evaluate the maximum reactor j coolant fission product activities. : 1 i The maximum activities are presented in Table 12.2-5. I i i i 12.2.1.1.2.2 Deposited Crud Activities , i j Radionuclide activity for circulating activated corrosion products (crud) from NUREG-0017 is assumed. i The development of NUREG-0017 radionuclide concentrations are based on ANSI N237 methodology; !

however, they differ due to the use of current industry operating data.

9 i , The development of improved Reactor Coolant System coolant chemistry and the specification of corrosion resistant materials in the Reactor Coolant System will assure very low circulating activated 4 crud. Therefore, the activity of circulating crud in the reactor coolant system can be maintained within ! NUREG-0017 concentrations listed in Table 12.2-6. ) 12.2.1.1.2.3 ' Neutron Activation Products 12.2.1.1.2.3.1 Nitrogen-16 Activity Nitrogen-16 is produced by the 160(n,p)i6N reaction. Nitrogen-16 decays by beta emission and high energy gamma emission 78% of the time. The gamma energies are 6.13 MeV, 73 % of the time and 7.10 j MeV,5% of the time. The nitrogen-16 half life is 7.13 seconds. The threshold energy for the reaction is 10.2 MeV. , 4 The nitrogen-16 activity at the pressure vessel outlet nozzle is 5.76 x 106 disintegrations /cm3 -sec. This i activity is based on the following expression and reactor parameters: j Amuoveet Design atesonio!- llediation Protoceien Page 12.2 3 4-

l System 80+ Design ControlDocument l Activity (disintegrations /cm3 -sec) = # II - * )* (1 - e 9 ) Where: 7 Ed is the reaction rate (4.72 x 10 d/cm3 - sec), te is the core transit time (0.79 sec), t, is the total primary loop time (8.6 sec), t, is the time from the active core outlet to the point of interest (0.69 see to outlet nozzle) and A is the decay constant (0.097 sec-1) 12.2.1.1.2.3.2 Carbon-14 Production Carbon 14 is produced in the RCS by activation of 017 and Nld isotopes. The greatest amount of C id l4 is produced by the 017 (n, a) C 14 reaction, a lesser amount of C is praiuced by the N 4 (n, p) C 4 l4 reaction. The production of C from both sources can be calculated by using the following equation: Q = N oo, & m t p s Where: No= atom concentration in the RCS water, (atoms /kg H2O) 2 ao= thermal cross section (cm ) & = thermal neutron flux,5.5 x 1013n/cm 2-s m = mass of core water,2.47 x 104 kg t = conversion factor (sec/yr) p = plant capacity factor,0.8 s = 1.03 x 10-22 Ci/ atom Q = production rate, Ci/ year For C-14 production from 017activation, N o = 1.3 x 1022atoms O l7/kg (H O) 2 and oo = 2.4 x 2 1 0-25 cm are used in the above equation. The production rate is 11.0 curies / year. For carbon-14 production from Nl ' activation N 2 2 o = 2.75 x 10 atoms N /kg (H 2O) and oo= 1.8 x 10-25 cm are used 34 in the above equation. The production rate is 1.8 curies / year. l The annual production of C34from these sources will be 12.8 curies. i O' Approved Design Material- Radiation Protection Page 12.2-4 l J

System 80+ Design c7ntr-1 Document l' (s 12.2.1.1.3 Main Steam Supply System ) , j i The rate of Steam Generator tube leakage is assumed to be 1 gal / min. This is assumed to be concurrent !

with the previously assumed 0.25% fuel cladding defects. A blowdown rate of 0.2% of main steam rate )

, is assumed. No credi' was taken for the condensate polisher demineralizers. )

12.2.1.1.4 Spent Fuel Handling Transfer I + The spent fuel assemblies are the predominant long-term source of radiation in the containment after plant

shutdown for refueling. A reactor operating time necessary to establish near-equilibrium fission product buildup for the reactor at rated power is used in determining the source strength. The initial fuel

;       composition that produced the maximum decay source is used. The spent fuel decay gamma source is given in Table 12.2-8. Fuel assembly dose rates as a function of distance in water in the refueling pool and time after shutdown are shown in Figures 12.2-1 and 12.2-2.

12.2.1.1.5 Processing Systems i i 12.2.1.1.5.1 Chemical and Volume Control System (CVCS) l The shielding design is based on the maximum expected activity in each component. These sources are i listed in Tables 12.2-9 through 12.2-12.

- ai Heat Exchangers (Table 12.2-9)

C% -

! V               Activities are provided on a concentration basis.

< I e Ion Exchangers (Table 12.2-10) i

1. Purification Ion Exchanger Total curie inventory is based on a resin buildup of 1.2 effective years. This ion exchanger is used for lithium removal and normal purification of RCS letdown. When it is used for lithium removal it is on line an average of 58 days prior to placing it in service as a purification ion exchanger for 300 days.

Radionuclide removal efficiencies consistent with NUREG-0017 guidance are used.

2. Deborating Ion Exchanger The total curie inventory is based on resin buildup of 14 days (2 bed volumes assumed).

This ion exchanger is used to reduce reactor coolant boron concentration toward the end of cycle length. Boron control in the CVCS is detailed further in Section 9.3.4. Radionuclide remeval efficiencies consistent with NUREG-0017 guidelines are used.

3. Preholdup lon Exchanger Total curie inventory is based on resin buildup of 1.2 effective year (438 days),

b Radionuclide removal efficiencies consistent with NUREG-0017 guidance are used. Approved Design Meterial . RoaVation Protection Page 12.2-5

System 80+ Design ControlDocument Sources provided by the prehold-up lon-exchanger include: 8.5 x 105 gallons (1931 gpd) of letdown previously processed through the purification Ion exchanger and purification filter, 200 gpd from the Reactor Drain Tank (RDT) and 50 gpd from the Equipment Drain Tank (EDT).

4. Boric Acid Condensate lon Exchanger Total curie inventory is based on resin buildup of 1.2 effective year (438 days).

Radionuclide removal efficiencies consistent with NUREG-0017 guidance are used.

Filters (Table 12.2-11)

Total curie inventories on all CVCS filters are based on crud buildup of 62 days. All CVCS filters removed crud with a decontamination factor of 10 (i.e., an efficiency of 90%). e Tanks (Table 12.2-12) Activities are provided on a concentration basis. 12.2.1.1.5.2 Steam Generator Blowdown System Radiation sources in the steam generator blowdown system are shown in Table 12.2-15. These are based on the primary-to-secondary leakage and failed fuel percentage of section 12.2.1.1.3. The blowdown rate is assumed to be 0.2% of main steam rate. In the event of a primary-to-secondary leak, radioactive Steam Generator Blowdown System Resins will be disposed of through the radioactive Solid Waste Management System. 12.2.1.1.S.3 Condensate Polishing System Space has been provided for shielding around the condensate polishing system however it has not been detailed. The need fo shielding only arises when tube leaks are occurring concurrently in both the steam generator and the cor denser. Inventories are located in Table 12.2-15. 12.2.i.2 Reactor Building Subsphere and Nuclear Annex 12.2.1.2.1 Shutdown Cooling System The pumps, heat exchangers, and associated piping of the Shutdown Cooling System (SCS) are potential carriers of radioactive materials. For plant shutdown, pumps and heat exchangers sources of radioactivity result from the radioactive isotopes carried in the reactor coolant, discussed in Section 12.2.1.1.2, after 4 hours of decay following shutdown and dilution. Table 12.2-13 provides a listing of the maximum specific source strengths in the SCS. 12.2.1.2.2 Component Cooling Water System The Component Cooling Water System (CCWS) is a closed loop demineralized water system which can potentially become contaminated by heat exchanger leakage from other radioactive systems being cooled by CCWS. The CCWS is designed to detect leakage of radioactive water into the CCWS. The detection of leakage is sensitive to a level of 1x104 Ci/ml gross gamma activity in the CCWS cooling water. In Approved Design Material . Radiation Protect % Page 12.2-6

l l }. Svat:m 80+ Design CrntrolDocument case of a major leak in one of the CCWS trains, that train is removed from service and the other train used. .In this way, radioactive leakage to the CCWS is controlled to lower than detectable levels.

12.2.1.3 Fuel Building l' 12.2.1.3.1 Spent Fuel Storage and Transfer

[' Spent fuel assemblies and associated crud are the primary source of radiation in the spent fuel storage and { transfer area. Shielding design assumes the maximum number of fuel assemblies in storage. Of these, l l 241 spent fuel assemblies are assumed to be from unloading the full core with 72 hours decay, and 81 } assemblies are assumed to be from previous refueling operations with at least 90 days of decay. 12.2.1.3.2 Spent Fuel Pool Cooling and Cleanup System 1

                                                                                                                               .I j                 Activity levels in the spent fuel pool cooling and cleanup system are determined based on the activities        j

[ present in the spent fuel pool. Normal and design basis spent fuel pool specific activities are summarized ' ! in Table 12.2-14. Spent fuel pool cooling and cleanup system filter and demineralizer maximum specific j l activities are also provided in Table 12.2-14. ! i .

i. Spent fuel pool normal and maximum fission and corrosion product specific activities are evaluated for !

. the start of the refueling period. It is assumed that upon shutdown for refueling, the Reactor Coolant l l System is cooled down for a period of approximately two days. During this period, the primary coolant !

is letdown through the purification filter, purification ion exchanger, gas stripper and volume control tank. This serves two purposes: 1) removing the noble gases in the gas stripper avoids large activity {( j releases to the containment following reactor vessel head removal, and 2) the ion exchange and filtration

              - reduces dissolved fission and corrosion products in the coolant which would otherwise enter the spent fuel pool and refueling water cavity. At the end of this period, the coolant above the reactor. vessel flange l:                is partially drained. The reactor vessel head is unbolted and the refueling water cavity is filled with         i

[ approximately 493,000 gallons of water from the IRWST. The remaining reactor coolant volume is then

mixed with water in the refueling cavity and the spent fuel pool. After refueling, the spent fuel pool is isolated and the water in the refueling cavity is returned to the IRWST. This series of events determines the total activity in the spent fuel pool. Table 12.2-14 radionuclide concentrations are calculated based l upon a total mixed volume of 800,000 gallons. Normal spent fuel pool concentrations'are calculated using Table 11.1.1-2 normal reactor coolant specific activities. Table 12.2-14 design basis spent fuel pool concentrations are calculated using Table 12.2-5 design basis reactor coolant specific activities. The spent
fuel pool activities are subsequently reduced by decay during refueling as well as by operation of the fuel pool cooling and cleanup system. There is no contribetion from defective fuel elements because of low power and temperature during storage and " degassing" during plant shutdown operations.
Maximum spent fuel pool filter and demineralizer inventories are calculated assuming decreasing water
activity and 30 days service following refueling.

i ! 12.2.1.4 Turbine Building ! Significant radiation sources in the turbine building are limited mainly to the steam generator blowdown i system. Activity levels for all turbine building related sources are summarized in Table 12.2-15. The specific activities provided in Table 12.2-15 are based on normal operation reactor coolant activity levels j and primary-to-secondary leakage conditions. Filter and ion exchange media are assumed to provide 1 j year of service and radionuclide removal efficiencies consistent with NUREG-0017 guidance are used. i L ANuoved Design niew - Magniet\ ion ProtecDon Page 12.2 7 ) i

                                                                               , -                 -    e

Sy' tem 80 + Design C*ntrolDicument 12.2.1.5 Nuclear Annax Nuclear Annex sources are generally confined to the CVCS sources addressed in Section 12.2.1.1.5. Gaseous waste management system source terms are provided in Table 12.2-16. The process gas specific activities are calculated using the design basis equilibrium reactor coolant radionuclide concentrations provided in Table 12.2-5. Activity build-up on the process gas charcoal beds is modeled assuming maximum design basis holdup times for noble gases calculated consistent with NUREG-0017 (i.e.,30 days for Xenon and 3.0 days for Krypton gases). 12.2.1.6 Radwaste Building Radwaste building tanks and process component source terms are summarized in Tables 11.2-17 through , 12.2-19. The radwaste building source terms provided include waste fluid and ion exchange resin specific activities calculated for CVCS and condensate cleanup components as well as calculated radwaste process equipment source terms. Equipment waste and floor drain tank fluid specific activities are calculated using Table 12.2-5 degassed reactor coolant equilibrium radionuclide concentrations and Table 11.2-2 activity fraction assumptions (i.e., equipment and floor drain tanks receive fluids with average primary coolant activity fractions of 0.2 and 0.02, respectively). The laundry and hot shower tank specific activities are calculated using NUREG-0017 annual detergent waste radionuclide release projections and assuming 540 gallon per day of detergent wastes are treated and released. The chemical waste tank will receive fluids of varying ; radioactive contamination levels and is shielded assuming relatively high levels (consistent with the equipment waste tank) may be received. The waste monitor tank source term is calculated using equipment waste tank radionuclide specific activities and an assumption that liquid waste processing equipment achieves an overall decontamination factor of 1000. Speific activity source terms for waste process filters and demineralizers are calculated using an activity build-up and decay model. Process flow rate assumptions consistent with Table 11.2-2 and process fluid , activity levels provided in Table 12.2-17 are used. For the purposes of the source term calculttion, waste l process filters and resin beds are assumed to have a 3 month useful service life. Although radwaste ! process filtration media source terms and useful service life will realistically vary, component sources will be controlled (i.e., media replacement based on elevated dose rate levels, if necessary) to assure occupational exposures asscciated with radwaste system operations remain ALARA. Specific activities for the high activity spent resin storage tanks are the same as calculated for the CVCS purification demineralizer resins presented in Table 12.2-10. The low activity spent resin storage tank source terms are taken from Table 12.2-19 values for waste process demineralizer resins. 12.2.1.7 Sources Resulting from Design Basis Accidents I Accident parameter and sources are discussed in Chapter 15. l r 12.2.1.8 Stored Radioactivity Tanks (holdup, reactor makeup water, boric acid and condensate storage) are the principle sources of activity located outside. The Condensate Storage Tank is expected to contain sources to yield a surface Approved Design Material- Radiation Protection Page 12.2-8

System 80+ Design C:ntrolDocument

 '} '

dose rate ofless than 0.2 mrem /hr. Section 15.7.3 analysis evaluated the consequences of a failure of holdup, reactor nakeup water and boric acid tanks. Spent fuel is stored in the spent fuel pool until it is placed in the spent fuel shipping cask for transport offsite. Storage space is allocated in the radwaste building for storage of spent filter cartridges and solidified spent resins, evaporator bottoms, and chemical wastes. Radioactive wastes stored inside plant structures are shielded so that there is design radiation Zone II access outside the structure. If radiation i levels outside the structure exceed the design radiation zone limit, or it is necessary to temporarily store radioactive waste outside plant structures, radiation protection measures are taken by the radiation protection staff to assure compliance with 10 CFR 20 and to be consistent with the recommendation of Regulatory Guide 8.8. 12.2.1.9 Field Run Pipe Routing Radioactive piping is not field routed. Procedures are followed when radioactive pipe is routed.

These procedures are utilized to provide guidance so that personnel exposure is maintained ALARA in i accordance with reconunendations in Regulatory Guide 8.8. Criteria for routing radioactive piping include

4

1. Radioactive piping is routed through shielded pipe chases.

2. Systems containing radioactive liquids, gases, or slurries are physically located in close proximity r3 to interfacing systems. This reduces the pipe length and minimizes the possibility of routing i

Q radioactive piping through personnel access corridors. Pipe routing through personnel access corridors is avoided.

3. Stagnant runs of piping are avoided to minimize the potential for crud traps. Flushing and decontamination ctpabilities are also provided as necessary.

12.2.2 Airborne Radioactive Material Sources Airborne radioactive material is introduced within the plant principally through: (1) leakage of i radioactive fluids from equipment (e.g., va've stems and pump sea's), (2) evaporation of tritiated water, and (3) recirculation of contaminated air discharged from the plant. System 80+ design limits leakage of udioactive fluids as much as reasonable. However, leakage of radioactive fluids through pump seals, valve stems, and flanges can not be eliminated entirely. Therefore, i utilizing recommendations and guidance provided by NUREG-0017, a source term is estimated based on the type of equipment, number of volves, flanges, and level of radioactivity in the fluid stream. Evaporation of tritiated water from s;uls or large bodies of water, such as the spent fuel pool, are significant sources of airborne tritium. The rate of evaporation is dependent on the pool temperature, l air velocity across the pool, and relative humidity. ) Intake louvers for ventilation systems are located on the exterior of buildings draw outside air into the plant. This air may be contaminated. The concentration of the radionuclides in the air at the intake is O a function of site specific characteristics, such as the atmospheric dispersion coefficient (x/Q), the wake effect from the surrounding structure. The release of low level of radioactivity from the unit vent is a continuous process. In general the airborne material will rise, due to the momentum and the buoyancy Approved Design Material . Redetion Protection Page 12.2-9

System 80+ Design CrntrolDocument of the effluent of the exhaust, and will be carried away by wind currents. However, fumigation of the effluent may occur due to inversion of the plume compounded by wake effects of nearby structures. This may cause the effluent to linger around the plant ventilation intakes where it can be drawn into the plant recirculating contaminated air discharged from the plant. 12.2.2.1 In-plant Concentrations The levels of airborne radioactivity within the plant during normal operation are calculated based on the following:

1. Activity of fluid stream, based on 0.25% fuel defects for fission products and ANSI N237 for coolant and corrosion activation product source term.

2. Characteristic leakage rate of equipment (e.g., valves).

3. Ventilation flow rate.

4. Evaporation rete of large pool (e.g., spent fuel pool).

The above information will be used to calculate the airborne concentration of radioisotopes in cubicles. It is assumed that in areas where there are no potential sources of radioactive leakage or evaporation, the concentration of radioactivity is equal to the concentration in the air external to the ventilation intakes. This is reasonable because the design of ventilation systems is such that airflows from areas of lower potential airborne radioactivity to areas of higher airborne radioactivity. For those areas with sources ofleakage or evaporation, the ececentration is calculated by: C=C+ 9 o 60xF where: I C = Room Concentration (pCi/ml) l Co e Outside air concentration for the appropriate ventilation system (pCi/ml) Q = Activity source term for room (pCi-ft3 /ml-hr) Q = LxPFxAo L = Leakage or evaporation rate (ft3 /hr liquid phase basis) PF = Partition Factor (The ratio of the quantity of a nuclide in the gas phase to the total quantity in both the liquid and gas phases when the liquid and the gas are at equilibrium.) Ao e Initial activity of fluid stream source ( Ci/ml) , I 3 F e Room exhaust flow rate (ft / min) 60 m Conversion factor (min /hr) ' Aptwoved Design Material- Radiation Protection Page 12.210

_ _ _ _ _ . - _ . - - - _ _ - . _ . _ . ~ .t Svit:m 80+ Derkn controlDocanent h < % Credit for decay has been neglected for conservatism. The airborne concentrations in rooms or cubicles

accessible to pusonnel throughout the plant will be maintained within inplant concentrations prescribed in 10 CFR 20. The design acceptance criteria necessary to assure plant design compliance with 10 CFR l j

20 requirements are specified in the Radiation Protection Design Acceptance Criteria document. The 4 design acceptance criteria specified include:

Maintain inplant airborne concentrations of radioisotopes within 10 CFR 20, Appendix B of I Sections 20.1001-20.2402, Table 1, Column 3 limits for areas requiring infrequent access.
Maintain inplant airborne concentrations of radioisotopes well within or a small fraction (i.e.,

0.25) of 10 CFR 20, Appendix B of Sections 20.1001-20.2402, Table 1, Column 3 limits for areas requiring continuous or frequent access. Inplant airborne concentrations will be calculated as described in Section 12.2.2.

Provide sufficient containment and ventilation capability to prevent the spread of airborne contamination.
Provide airborne radiation monitors in areas normally occupied and where a potential of airborne contammation exists. The airborne radiation monitoring system is discussed in Section 11.5.

((The COL Applicant will evaluate the plant design in accordance with this Design Acceptance Criteria document to assure airborne concentrations are maintained less than 10 CFR 20.1204 limits.))l The occupational exposure to personnel in areas with airborne radionuclides is dependent on: o concentration of airborne contamination in the cubicle, calculated as shown above,- o personnel breathing rate, and

occupancy time in the area.

Inhalation dose is calculated as follows: Dr = { C(i) x DCF(i) j x BR x t x CF Where: Dr = TotalInhalation Dose (Rem) C(i) = Concentration of the ith isotope in the air to which the worker is exposed (pCi/ml) BR e Breathing Rate (m3 /sec) = 3.47E-4 DCF(i)j = Dose conversion factor for the jth organ of interest (mrem /pCi) (ICRP 30) 1 COL infonution item: see DCD Introduction Section 3.2. Appnpuoef Deeepn nietenef Realfwh%m Prom Page 12.2-11

Svatem 80+ Design ControlDocument t = Occupancy time (hrs) 3 CF = Conversion Factor (ml-rem-pCi-sec/m - mrem-pCi-hr) = 3.6E+12 Whole body external dose from immersion in a semi-infinite cloud of noble gases is calculated as follows: Dw3 = C(i) x DFB(i) x t x CF x K Where: Dws e External whole body dose due immersion in a semi-infinite cloud (rem) C(i) e Concentration of the ith isotope in the cubicle (pCi/ml) DFB(i) m Whole body dose factor (mrem-m3 /pCi-yr) (Regulatory Guide 1.109) CF = Conversion factor (yr-pCi-rem-ml/sec-pCi- mrem-m3 ) = 31.69 t e Occupancy time (sec) K a Correction factor from semi-infinite cloud to a non-infinite cloud geometry The personnel inhalation dose and external whole body dose will be maintained within 10 CFR 20.101 limits. 12.2.3 Sources Used in NUREG-0737 Post-Accident Shielding Review Item II.B.2 of NUREG-0737 clarifies the requirement for ensuring that areas which require post-accident personnel access or contain safety-related equipment are adequately shielded in the vicinity of systems which may contain highly radioactive materials as a result of the Design Basis Accident. ((A radiation and shielding design review of the System 80+ Standard Design in accordance with Item II.B.2 of NUREG-0737 amended by the guidance provided in Draft NUREG-1465 relative to post LOCA source tenns is perfonrad during the detailed design phase of the plant.))l The review of systems that, as a result of an accident, contain highly radioactive materials was performed using the same methodology described in Section 12.3.2. Initial core releases are used which are equivalent to those recommended in Draf; NUREG-1465. The source terms are presented in Table 12.2-20. The sources are characteristic of a depressurized system. Plant areas requiring post-accident occupation (" vital areas), and the duration of occupation are identified. These source terms are used to evaluate the adequacy of shielding in post-accident conditions using shielding codes, discussed in Section 12.3, to verify that: e Vital areas are accessible to operators to take the mitigative actions during post-accident conditions, and I O COL information item; see DCD Introduction Section 3.2. Approved Design Material- Radiation Protection (11/96) Page 12.212

1 Syntim 80+ Deskn CentrolDocumart e Safety related equipment are qualified for the radiation area in which they are located. For vital areas requiring irregular access, sufficient shielding will be provided to ensure the whole body dose, or its equivalent, is limited to 5 rem for the duration of the accident in accordance with 10 CFR 50, Appendix A (General Design Criterion 19). For vital areas requiring continuous occupancy, such ! as the control room, the local radiation levels are limited to 15 mrem /hr averaged over 30 days per j NUREG-0737, Section II.B.2.3. j l References for Section 12.2 l I

1. M. E. Meek, B. F. Rider, " Summary of Fission Product Yields," NEDO12154, January 1972. )

i

2. " Chart of Nuclides," USAEC, Modified by Battelle-Northwest, May 1%9 and May 1970.

1

3. " Neutron Cross Sections," BNL 325 Supplement No. 2, May 1964.

l J

4. J. D. Eichenberg, " Effects of Irradiation on Bulk UO2," WAPD-183, October 1957.

5. G. M. Allison and H. K. Rae, "The Release of Fission Gases and lodines from Defected UO2 -

Fuel Elements of Different Lengths," AECL-2206, June 1%5. 24i

6. M E. Meek, B. F. Rider, " Summary of Fission Product Yields for U2", U23s, Pu2 ", and Pu ,

at Thermal, Fission Spectrum and 14 MeV Neutron Energies," APED-5398, Class 1, March 1, A 1968. V 7. ANL-7450 Chemical Engineering Division Research Highlights, May 1%7-April 1%8.

8. E. P. Lippincott, A. L. Pitner and L. S. Kellog, " Measurement of ' B (n,t) Cross Section in a ;

Fast Neutron Spectrum," HEDL-TME-73-49, May 1973. [

9. " Neutron Cross Sections," BNL 325 Supplement No. 2, May 1964. i

10. Point Beach Semi-Annual Reports, 601-1n4. l

11. H. B. Robinson Semi-Annual Reports, 6/71-165, 3

12. Ginna Semi-Annual Reports,601-In5.

13. Source Term Data for Westinghouse Pressurized Water Reactors, WCAP8253, May 1974.

14. P. J. Grant et. al., "Oconee Radiochemistry Survey Program,". RDTPL-754, May 1975. l

15. Omaha Semi-Annual Reports, 1973-1975. j l

16. U.S. Nuclear Regulatory Commission, " Calculation of Releases of Radioactive Materials in l i

Gaseous and Liquid Effluents from Pressurized Water Reactors," NUREG-0017, April 1985. O 17. Draft NUREG 1465, " Accident Source Terms for Light Water Nuclear Power Plants," June 1992. 4promt oeste noneuw . nenetia, proneesia, is tiss) roe 12.2-ss

System 80+ Dwign ControlDocument Table 12.2-1 Maximum Neutron Spectra Outside Reactor Vessel h Average Neutron Energy (Mev) Neutron Spectra (neutrons /cm -s)Di 2 13.60 5.90 x 10+6 11.10 1.86 x 10+' 9.10 3.79 x 10+' 7.27 6.87 x 10*' 5.66 1.08 x 10+' 4.51 9.08 x 10+' 3.53 1.58 x 10+' 2.73 2.01 x 10+' 2.40 6.69 x 10+' 2.09 3.86 x 10+' l.47 1.48 x 10+' 8.30 x 17' 5.20 x 10+' 3.30 x la' l.47 x 10" 5.70 x 10-2 1.05 x 10+' I 1.96 x 103 3.24 x 10+' 3.42 x 104 2.96 x 10+' 6.50 x 105 2.01 x 10+' 1 1.98 x 10-8 1.27 x 10+' l 6.90 x 104 1.44 x 10+' 2.09 x 104 1.09 x 10+' 7.60 x 10-' 9.67 x 10+: 2.50 x 10-8 (thermal) 6.22 x 10+' O DI At core midplane, one half foot from vessel surface Approved Design Material Radiation Protection Page 12.214

Sy't m 80 + Design CEntrolDocument i [ Table 12.2-2 Maximum Gamma Spectra Outside Reactor Vessel Average Gamma Energy (Mev) Gamma Spectra (Gamma /cm2 3)tij 9.00 2.17 x 10+8 7.25 1.27 x 10+' 5.75 8.42 x 10+: 4.50 7.21 x 10+: 3.50 9.90 x 10+: 2.75 6.60 x 10+s 2.25 1.11 x 10+' 1.83 8.19 x 10+: 1.50 8.43 x 10+8 1.16 1.07 x 10+' O.90 8.06 x 10+8 0.70 1.03 x 10+' O.50 2.46 x 10+' t 0.35 1.56 x 10+' O.25 2.60 x 10+' O.15 4.14 x 10+' i 0.075 1.05 x 10+' O.025 4.99 x 10+' i i1 At core midplane, one half foot from vessel surface. j Approved Desipt Material- Radiation Protection Page 12.2-16 1

l Sy ~t m 80 + De:ign controlDocument Table 12.2-3 Shutdown Gamma Spectra Outside Reactor Vessel Average Gamma Energy (Mev) Decay Gamma Material Activation 2 2 (Gamma /cm -s)ttl (Gamma /cm .3)til 2.75 1.18 x 10+4 - 2.25 3.72 x 10+4 -

i 1.83 8.24 x 10+4 - 1 I 1.50 1.37 x 10+5 . 1.16 1.92 x 10+5 7.39 x 10+5 0.90 1.48 x 10+5 8.50 x 10+4 0.70 1.84 x 10+5 1.00 x 10+5 0.50 2.44 x 10+5 1.48 x 10+5 0.35 2.08 x 10+5 1.29 x 10+5 0.25 3.81 x 10+5 2.63 x 10+8 0.15 5.86 x 10+5 4.06 x 10+5 0.075 1.38 x 10+5 9.80 x 10+' O.025 6.89 x 10+2 5.00 x 10+2 i I 1 l l l l l l l l l O Ill At core midplan, one half-foot from vessel surface. At 48 hours after shutdown. Approved Design Material- Radiation Protection Page 12.2-16

i l Syst~m 80 + De'ign Crntrol Drcument

 ,.m                                                                                                                  i V)

Table 12.2-4 Basis for Reactor Coolant Fission Product Activities I Parameter Maximum Performancelu Core Power Level (MWt) 4100 3931 Duration of Reactor Operation (core cycles) 5 5 Equilibrium Fuel Cycle (Equivalent Full Power Days) 438 438 Average Thermal Fission Rate (Fission /MW-second) 3.10x10 68 3.10X10 16 Thermal Neutron Flux - average (n/cm 2-second) 5.50x10 3 5.50X1013 Fraction of Failed Fuel 0.0025 0.0025 Reactor Coolmt Mass including pressurizer (Pounds) 5.752x10 5 5.752x105 Core Coolant Volume to Reactor Coolant Volume Ratio 0.0723 0.0723 Purification Flow (gpm) 72 72 Purification Flow, yearly average for boron control (gpm) 0.48 0.48 I Boron Concentration Reduction Rate (ppm /second) 4.51x10-5 4.51x10-5 Beg! aings of Life Boron Concentration (ppm) 1200 1200 Ion Exchanger and Gas Stripper removal efficiency j CVCS Purification Ion Exchanger l Noble gas, tritium 0 0 Cs,Rb 0.5 0.5 1 1

All others 0.9 0.9 1 I [] C' CVCS Lithium Removal lon Exchangert2j Noble gas, tritium 0 0 All others 0.9 0.9 CVCS Gas Stripper Removal Efficiency Noble gas 0.999 0.999 All others 0 0 CVCS Gas Stripper Operation Continuous None , Fission Product Escape Rate Coefficients (sec'!) Noble Gases 6.5 x 10.s 6.5 X 10.s l Halogens 1.3 x 10-8 1.3 X 10.s CS 1.3 x 10r s 1.3 X 10-8 4 Te, Mo 1.0 x 10 1.0 X 104 All Others 1.6 x 10-12 1.6 X 10-12 IU Conditions for use in system operational performance evaluations. p C) [2] Nuclides are also removed from the letdown flow via the CVCS Lithium Removal lon Exchanger. This ion exchnager is used in series with the CVCS Purification lon Exchanger during approximately 20% of the core cycle. Approved Design Material- Radiation Protection Page 12.2-17 u

l Syotem 80+ Deoign ControlDocument Table 12.2-5 Reactor Coolant Equilibrium Concentration Maximum Values (pCi/gm) Isotope Concentrationto Br-83 5.34E-04 Br-84 5.65E-03 Br-85 6.83E44 Br-86 1.20E-05 Rb-88 6.66E41 Rb-89 3.53E-02 Sr-90 6. I 1 E-03 St-89 9.42E-04 Sr-90 3.30E-05 Sr-91 1.39E-03 Y-90 9.38E-06 Y-91 1.35E-04 Y-91m 8.10E-04 Y-93 3.34E-05 Zr-95 1.47E44 Nb-95 1.46E-04 Tc-99m 4.67E-02 Mo-99 8.13E-02 Ru-103 5.02E-05 Ru-106 1.83E-05 1-130 2.06E-04 1-131 7. l lE-01 1-132 1.92E-01 1-133 1.02E +00 1-134 1.20E-01 1-135 5.74 E-01 Sn-125 3.44 E-07 Sb-125 1.48E-07 Sb-127 3.65E-07 Te-125m 8.19E 4)5 Te-127 1.38E44 Te-127m 4.20E-05 Te-129 1.82 E-03 Te-129m 1.71 E-03 Te-131 3.18E-03 Te-131m 8.12E-03 Te-132 5.65 E-02 , Te-133 1.65E-03 Te-133m 4.83E-03 Te-134 6.40E43 Cs-134 4.99E-02 Approved Design Material- Radiation Protection Page 12.2-18

( Sv t:m 80 + Design ControlDocument 1 ( i Table 12.2-5 Reactor Coolant Equilibrium Concentration (Cont'd.) i { i j Maximum Values ; j (pCl/gm) '

Isotope Concentration"' I Cs-13 79E Cs-138 2.09E-01 !

Ba-137m 8.29E-02 ! . Ba-140 1.15E-03 , Ba-141 6.93E-06 La-140 3.88E-04

La-141 1.63E-05 Cc-141 4.33E-05

! Ce-143 1.21E-04 Cc-144 1.10E-04 ) Pr-144 1.10E-04 l- Kr-83m 3.06E-03 Kr-85 1.20E+00 j Kr-85m 3.03E-01 , Kr-87 2.35E-01 .! j , Kr-88 . 6.60E-01

Kr-89 1,76E-02

{

. Kr-90 3.38E-03 Xe-131m 1.41E+00 Xe-133 8.55E+01 l )

l .Xe-133m 7.92E-02 !- Xc-135 1.79E+00 i Xe-135m 1.71E-01 2 Xe-137 3.98E-02 >- Xe-138 1.45E-01 , Mn-54 1.90E-03 . Co-58 5.47E-03 Co-60 6.31E-04

Fe-59 3.57E-04 j- Cr-51 3.68E-03 i H-3 2.50E+00m i

s i 5: i l ji Il3 These isotopic inventories are applicable to the CVCS Letdown and Regenerative Heat Exchangers. A The Maximum concentration of tritium in the reactor coolant was obtained from Section 11.1.1. ,' . 4prend Deengo nietwini Medinaion Protecaion Page 12.2-19 I !

i. . -

I Spt ~'m 80 + D: sign C?ntr:IDocwnent ! Table 12.2-6 Average Reactor Coolant Crud Activity Isotope Activity (pCi/gm) Mn-54 1.90E-03 4 Co-58 5.47E-03 Co-60 6.31E-04 Fe 59 3.57E-04 Cr-51 3.68E-03 Table 12.2-7 N-16 Activity Location Activity (disintegrations /cm 3-s) Vessel Outlet Nozzle 5.76 x 10+6 Vessel Outlet Line (midpoint) 5.69 x 10+6 Steam Generator (midpoint) 4.61 x 10+6 Pump (midpoint) 3.71 x 10+6 Vessel Inlet Line (midpoint) 3.49 x 10+6 l l l l l l l l l O Approved Design Material Radiation Protection Page 12.2 20

I i Sy.? tim 80 + Drian C ntr:1 Document l 4 Table 12.2-8 Spent Fuel Gamma Source l i Gamma Source (Mev/ watt-s) Time After Shutdown

Mean Energy (Mev) 50 hr 200 hr 500 hr 1000 hr

, 1.500E4)2 .275E +09 .996E+08 .521E+08 .384E+08 ' 2.500E-02 .660E+08 .364E+08 .225E+08 .158E +08 i I 3.750E-02 .120E +09 .644E+08 .376E+08 .251E+08 l 5.750E-02 .118E+09 .639E+08 .3%E+08 .288E+08

8.500E-02 .319E +09 .119E+09 .490E+08 .303E+08

, 1.250E-01 .110E + 10 .335E+09 .146E+09 .985E+08 4 2.250E-01 .142E+ 10 .353E+09 .10$E+09 .677E+08 . 3.750E-01 .802E+09 .425E+09 .198E+09 .872E+08 5.750E-01 .354E+ 10 .201E+10 .132E+ 10 .924E+09 I 8.500E-01 .540E + 10 .388E+ 10 .309E+ 10 .251E+ 10 1.250E+00 .838E +09 .356E+09 .169E+09 .991E+08 1.750E+00 .301E+ 10 .221E+ 10 .113E+ 10 .372E+09 l 2.250E+00 .201E+09 .134E+09 .791E+08 .434E+08

2.750E+00 .175E+09 .130E+09 .662E +08 .216E+08 3.500E+00 .187E +07 .141E+07 .745E +06 .277E+06 5.000E +00 .376E+02 .437E +01 .425E+01 .405E+01 7.000E+00 .715E+00 .704E+00 .684 E+00 .653E+00 1.100E+01 .129E+00 .128E+00 .124E+00 .118E+00 l

v Approved Design Material- Radiation Protection Page 12.2-21

System 80+ Design ControlDocument Table 12.2-9 CVCS IIeat Exchanger Inventories Maximum Values ( Ci/gm) Isotope Regenerative Letdown Injection Seal 3 Sr-83 5.34E 04 5.34E-04 5.33E-06 Br-84 5.65E-03 5.65E-03 5.64E-05 Br-85 6.83E-04 6.83E-04 6.82E-06 Rb-86 1.20E-05 1.20E-05 6.00E-06 Rb-88 6.66E-01 6.66E-01 3.32E-01 Rb-89 3.53E-02 3.53E-02 1.76E-02 Rb-90 6.I1E-03 6.I1E-03 3.05E-03 St-89 9.42E-04 9.42E-04 1.88E-05 Sr-90 3.30E-05 3.30E-05 6.58E-07 Sr-91 1.39E-03 1.39E-03 2.78E-05 Y-90 9.38E-06 9.38E-06 1.87E-07 Y-91 1.35E-04 1.35E-04 2.70E-06 Y-91m 8.10E-04 8.10E-04 1.62E-05 Y-93 3.34E-05 3.34E-05 6.66E-07 Zr-95 1.47E-04 1.47E-G4 2.94E-06 Nb-95 1.46E-04 1.46E44 2.92E-06 Tc-99m 4.67E-02 4.67E-02 9.32E-04 Mo-99 8.13E-02 8.13E-02 1.62E-03 Ru-103 5.02E-05 5.02E-05 1.00E-06 Ru-106 1.83E-05 1.83E-05 3.66E-07 I-130 2.06E-04 2.06E44 2.06E-06 l I-131 7.11E-01 7.11E-01 7.10E-03 1-132 1.92E-01 1.92E-01 1.92E-03 I-133 1.02E +00 1.02E +00 1.02E-02 I-134 1.20E-01 1.20E-01 1.20E-03 l I-135 5.74E-01 5.74E-01 5.73E-03 Sn-125 3.44E-07 3.44E-07 6.86E-09 Approved Design Material- Radiation Protection Page 12.2-22

Syst m 80 + Design CrntrolDocument

  ,o U)

I Table 12.2-9 CVCS Heat Exchanger Inventories (Cont'd.) Maximum Values (pCilgm) Isotope Regenerative Letdown Injection Seal Sb-125 1.48E-07 1.48E-07 2.96E-09 Sb-127 3.65E-07 3.65E-07 7.28E-09 Te-125m 8.19E-05 8.19E-05 1.63E-06 Te-127 1.38E-M 1.38E-M 2.76E-06 Te-127m 4.20E-05 4.20E-05 8.38E-07 Te-129 1.82E-03 1.82E-03 3.64E-05 Te-129m 1.71E-03 1.71E-03 3.42E-05 Te-131 3.I8E-03 3.I8E-03 6.34E-05 Te-131m 8.12E-03 S.12E43 1.62E-04 Te-132 5.65E-02 5.65E42 1.13E-03 (~~x, Te-133 1.65E-03 1.65E-03 3.30E-05 t J

 'v'             Te-133m                              4.83E-03           4.83E-03            9.64E-05 Te-134                              6.40E-03           6.40E-03             1.28E-04 Cs-134                              4.99E-02           4.99E42             2.49E-02 Cs-136                               1.35E-02           1.35E-02           6.75E-03 Cs-137                              8.79E-02           8.79E-02            4.39E-02 Cs-138                              2.09E-01           2.09E-01             1.NE-01 Ba-137m                              8.29E-02           8.29E-02             1.65E-03      )

Ba-140 1.15E-03 1.15E-03 2.30E-05 Ba-141 6.93E-06 6.93E-06 1.38E-07 La-140 3.88E-04 3.88E-04 7.74E-06 l La-141 1.63E-05 1.63E-05 3.26E-07 i Ce-141 4.33E-05 4.33E-05 8.64E-07 Ce-143 1.21 E-04 1.21E-04 2.42E-06 Cc-144 1.10E-04 1.10E-04 2.20E-06 73 Pr-144 1.10E-04 1.10E-N 2.20E-06 Kr-83m 3.06E-03 3.06E-03 5.18E-04 Kr-85 1.20E+00 1.20E+00 1.20E+00 Approved Design Material- Radiation Protection Page 12.2-23

Sy:t m 80+ D=ign C?ntrol D:cument Table 12.2-9 CVCS Heat Exchanger Inventories (Cont'd.) Maximum Values ( Ci/gm) Isotope Regenerative Letdown Injection Seal Kr-85m 3.03E-01 3.03E-01 9.97E-02 Kr-87 2.35E-01 2.35E-01 2.82E-02 Kr-88 6.60E-01 6.60E-01 1.58E-01 Kr-89 1.76E-02 1.76E-02 1.76E-04 Kr-90 3.38E-03 3.38E-03 --- Xe-131m 1.41E+00 1.41E +00 1.38E +00 Xe-133 8.55E +01 8.55E+01 8.10E+01 Xe-133m 7.92E-02 7.92E-02 7.03E-02 Xe-135 1.79E+00 1.79E+00 1.06E+00 Xe-135m 1.71E-01 1.71E-01 6.84E-03 Xe-137 3.98E-02 3.98E-02 3.97E-04 Xe-13S 1.45 E-01 1.45E-01 5.80E-03 Mn-54 1.90E-03 1.90E-03 3.80E-06 Co-58 5.47E-03 5.47E-03 1.09E-05 Co-60 6.31E-04 6.31E-04 1.26E-06 Fe-59 3.57E-04 3.57E-04 7.12E-07 Cr-51 3.68E-03 3.68E-03 7.34E-06 O Approved Design Material- Radiation Protection Page 12.2-24

                                                                     = _ _    .      - _    .. . . -          - - . . . .

Syst~m 80+ Drsign Crntrol Document Il v Table 12.2-10 CVCS Ion Exchanger and Evaporator Inventories Maximum Values (pCi/ml) Boric Acid Purification IX Deborating IX Pre-Holdup IX Concentrator Boric Acid Cond. IX Isotope ( Ci/ml) (pCi/ml)DI (pCi/ml) (pCi/ml) (pCi/ml) 3 3 3 Volume Basis 32ft 64ft 32ft VR = 110 32ft 3 Br-83 3.30E-02 1.65E-04 1.03E-05 6.77E-09 9.16E-08 Br-84 7.70E-02 3.85E-04 2.40E-05 - 2.14E-07 Br-85 8.41E-04 4.20E-06 2.62E-07 - 2.34E-09 Rb-86 6.99E-02 - 1.08E-04 3.46E-04 -- Rb-88 2.57E+00 - 3.98E-03 - -- Rb-89 1.19E-01 -- 1.85E-G4 -- -- Rb-90 3.38E-03 - 5.06E-06 - -- Sr-89 2.90E+01 - 9.80E-03 1.54E-03 9.88E-06 St-90 6.02E+00 - 2.03E-03 4.71E-05 2.05E-06 Sr-91 3.46E-01 - 1.17E-04 1.16E-04 1.18E-07 Y-90 1.53E-02 - 5.17E-06 2.52E-05 5.21E-09 Y-91 4.81E+00 - 1.62E-03 2.03E-04 . 1.63E-06 Y-91m 1.71E-02 5.76E-06 7.%E-05 5.81E-09 g - i Y-93 8.12E-03 - 2.74E46 2.62E46 2.76E-09 Zr-95 5.84E+00 - 1.97E-03 2.07E-04 1.99E-06 Nb-95 3.14E+00 - 1.06E-03 2.09E-04 1.07E-06 Tc-99m 7.17E+00 - 2.42E-03 7.22E-02 2.44E-06 Mo-99 1.38E+02 - 4.66E-02 7.65E-02 4.70E-05 Ru-103 1.21E +00 - 4.10E-04 6.96E-05 4.13E-07 Ru-106 2.31E+00 - 7.80E-04 2.61E-05 7.87E-07 l-130 6.58E42 3.29E-04 2.05E-05 2.91E-05 1.83E-07 I-131 3.53E+03 1.24E+01 1.10E+00 8.13E-01 9.81E-03 I-132 1.13E+01 5.64E-02 3.51E-03 5.82E-02 3.13E-05 I-133 5.52E+02 2.76E+00 1.72E-01 3.60E-01 1.53E-03 l I I-134 2.74E+00 1.37E-02 8.54E-04 5.05E-15 7.62E-06 I-135 9.89E +01 4.95E-01 3.086-02 1.20E-02 2.75E-04 Sn-125 2.02E-03 - 6.82E-07 4.35E-07 6.88E-10 Sb-125 2.37E-02 -- 8.01E-06 2.12E-07 8.07E-09 Sb-127 8.58E-04 - 2.89E-07 3.86E-07 2.92E-10 Te-125m 2.89E +00 - 9.73E-04 1.15E-04 9.81E-07 Te-127 3.31E-02 - 1.12E-05 6.57E-05 1.13E-08 Te-127m 2.62E+00 -- 8.85E-04 5.93E-05 8.92E-07 Te-129 5.34E-02 1.80E-05 1.50E-03 1.82E-08 g --

 'y/       Te-129m            3.51E+01               -            1.18E-02    2.36E-03           1.19E-05 i

Approved Design Material Radiation Pmtection Page 12.2-25

Syst:m 30+ Maximum Values (pCi/ml) Isotope (pCi/mt) Volume Basis 1.14E-05 1.20E-03 2.09E-03 4.60E-03 3.79E-02 5.65E-02 2.95E-06 1.04E-16 3.83E-05 6.20E-16 3.83E-05 -- 3.86E-08 , Cs-134 3.88E+03 6.02E +00 1.53E +00 Cs-136 5.48E+01 8.50E-02 3.79E-01 Te-133 8.74E-03 2.95E-06 1.04E-16 3.83E-05 6.20E-16 Te-134 3.83E-05 - 3.86E-08 l Cs-134 6.02E +00 1.53E+00 Cs-136 5.48E +01 8.50E-02 3.79E-01 Cs-137 8.19E +03 1.27E+01 2.69E+00 Cs-138 1.51E +00 2.35E-03 - l Ba-137m 3.02E-05 2.53E+00 3.04E-03 1.50E-03 1.81E-08 --- 1.83E-Il La-140 1.34 E-04 1.06E-03 5.52E-07 2.11E-08 2.90E-04 5.97E-05 3.44E-05 7.47E-05 4.24E-03 1.57E-04 2.73E-07 1.57E46 7.68E-03 2.08E-04 7.99E-03 5.91E-04 3.68E-03 6.93E-05 3.30E-04 3.82E-05 2.12E-03 3.87E-04 1.23E-04 1.70E-04 Approved Design Design Crntr-lDocument Table 12.2-10 CVCS Ion Exchanger and Evaporator Inventories (Cont'd.) Purification IX Deborating IX Pre-Holdup IX Boric Acid Boric Acid Cond. IX (pCi/ml) (pCi/ml)DI (pCi/ml) Concentrator (pCi/ml) 32ft 3 64ft 3 32ft 3 VR= 110 32ft 3 Te-131 3.37E-02 -- 1.15E-08 Te-131m 6.20E+00 - 2.llE-06 Te-132 1.12E+02 - 3.82E-05 Te-133 8.74E-03 -- 2.97E-09 Te-133m 1.13E-01 - 3.86E-08 i Te-134 1.14E-01 - -- - - -- - 2.97E-09 Te-133m 1.13E-01 --- 3.86E-08 ! 1.14E-01 - 3.88E+03 - - - - - - -- - 8.97E-02 -- 3.05E-08 Ba-140 9.01E+00 - 3.06E-06 Ba-141 5.37E-05 - 3.97E-01 - 1.35E-07 La-141 1.64E-03 - 5.57E-10 Ce-141 8.58E-01 --- 2.92E-07 Cc-143 1.02E-01 - 3.47E-08 Ce-144 1.26E +01 -- 4.27E-06 Pr-144 8.09E-04 -- 2.75E-10 Mn-54 2.07E +01 --- 7.68E-06 Co-58 2.35E +01 - 7.99E-06 Co-60 1.08E +01 --- 3.68E-06 Fe-59 9.71E-01 -- 3.30E-07 Cr-51 6.23E+00 - 2.12E-06 Np-239 3.63E-01 --- 1.23E-07 UI Deborating IX assumed to be operating downstream of Purification IX. Shielding design shall bc based on Purification IX source term to provide flexibility to operate Deborating IX as a Purification IX. Material- Radiation Protection Page 12.2-26

Sy~ tem 80+ D sign C*ntrol Drcument j ( j) t,_ Table 12.2-11 CVCS Filter and Gas Stripper Inventories Maximum Values (pC1/ml) Isotope Purification Seal Reactor Gas Stripper Boric Acid Reactor l Filter Injection Drain Filter Exhaust Filter Makeup I ( Ci/ml) Filter (pCl/ml) ( Ci/ml) (pCi/ml) Filter (pCi/ml) (pCl/ml) Volume Basis 2.7 ft3 0.4 ft3 1.3 ft3 0.3 scfm 2.7 ft3 2.7 ft3 1 1 Kr-83m -- -- --- 9.14E-02 - - Kr-85 --- - - 3.60E+01 - - Kr-85m -- -- -- 9.05E+00 --- - Kr-87 - -- -- 7.04E+00 --- - Kr-88 - - --- 1.98E+01 --- - Kr-89 -- - -- 5.27E-01 - -- Kr-90 - - - 1.01E-01 --- - Xe-131m - - - 4.23E+01 - -- Xe-133 -- - -- 2.56E+03 - --- Xe-133m - -- - 2.37E+00 -- -- (') C'# Xe-135 --- - - 5.36E+01 - --- Xe-135m - - - 5.12E+00 -- - Xe-137 - - - 1.19E+00 - - Xe-138 -- -- -- 4.35E + 00 - --- Mn-54 5.58E+02 7.77E+00 3.22E +00 - 1.74E-02 2.06E-06 Co-58 1.29E+03 6.40E+00 7.42E +00 - 4.01E-02 4.74E-06 Co-60 1.96E tt,2 9.74E-01 1.13E+00 - 6.11E-03 7.22E-07 Fe-59 7.16E+01 3.56E-01 4.13E-01 - 2.23E-03 2.64E-07 Cr-51 5.84E+02 2.90E +00 3.37E +00 - 1.82E-02 2.15E-06 Np-239 4.30E+01 2.14E-01 2.48E-01 - 1.34E-03 1.58E-07

 ,r~.,

Lv) Approved Design Material . Radiation Protection Page 12.2-27

l Sy t m 80 + D ~ sign Control Document Table 12.2-12 CVCS Tank Inventories hiaximum Values (pCi/gm) Volume Control Reactor and Equip. Holdup Boric hiakeup Drain Acid Isotope Liquid Gas Liquid Gas Liquid Gas Liquid Liquid (pCl/gm) (pCi/ml) (pCi/gm) (pCi/ml) ( Ci/gm) ( Ci/ml) ( Cilgm) (pCi/ml) Br-83 5.33E-06 - 5.33E-04 -- 6.15E-11 - 6.77E-09 1.28E-13 Br-84 5.64E-05 -- 5.64E.03 --- - - - - Br-85 6.82E-06 - 6.82E-04 --- -- - - - Rb-86 6.00E-06 - 1.20E-05 -- 3.15E-06 -- 3.46E-04 3.15E-09 Rb-88 3.32E41 -- 6.65E-01 --- - --- - --- Rb-89 1.76E-02 - 3.52E-02 --- - - -- --- Rb-90 3.05E-03 - 6.10E-03 --- - -- -- -- St-89 1.88E-05 --- 9.40E-04 --- 1.40E-05 --- 1.54E-03 2.98E-09 St-90 6.58E-07 - 3.29E-05 -- 4.28E-07 - 4.71E-05 3.32E-11 Sr-91 2.78E-05 - 1.39E-03 -- 1.05E-06 - 1.16E-N 8.09E-11 I Y-90 1.87E-07 -- 9.36E-06 - 2.29E-07 - 2.52E-05 1.76E-11 Y-91 2.70E-06 -- 1.35E-04 - 1.84E-06 - 2.03 E-04 1.42E-10 ; Y-91m 1.62E-05 -- 8.08E-N --- 7.24E-07 - 7.96E-05 5.57E-11 l Y-93 6.66E-07 - 3.33E-05 - 2.38E-08 - 2.62E-06 1.83E-12 Zr-95 2.94E-06 - 1.47E-04 -- 1.88E-06 - 2.07E-04 1.44E-10 I Nb-95 2.92E-06 --- 1.46E-04 - 1.90E-06 - 2.09E-04 1.46E-10 Tc-99m 9.32E-04 -- 4.66E-02 - 6.56E-N - 7.22E-02 5.05E-08 j Mo-99 1.62E-03 - 8.11E-02 - 6.96E-04 - 7.65E-02 5.35E-08 ] Ru-103 1.00E-06 -- 5.01E-05 - 6.33E-07 -- 6.96E-05 4.87E-11 Ru-106 3.66E-07 - 1.83E-05 - 2.37E-07 - 2.61E-05 1.82E-11 1-130 2.06E-06 -- 2.06E-04 - 2.64E-07 --- 2.91E-05 5.50E-10 l I-131 7.10E-03 - 7.10E-01 - 7.39E-03 -- 8.13E-01 1.54E-05 I-132 1.92E-03 - 1.92E-01 - 5.29E-N - 5.82E-02 4.07E-08 I-133 1.02E-02 - 1.02E+ 00 - - - 3.27E-03 - 3.60E-01 6.81E-06 I-134 1.20E-03 - 1.20E-01 - 4.59E-17 --- 5.05E-15 - I-135 5.73E-03 - 5.73E-01 --- 1.09E-04 --- 1.20E-02 2.28E-07 Sn-125 6.86E-09 - 3.43E-07 - 3.95E-09 --- 4.35E-07 3.04E-13 Sb-125 2.96E-09 -- 1.48E-07 -- 1.93E-09 -- 2.12E-07 1.48E-13 Sb-127 7.28E-09 - 3.64E-07 - 3.51E-09 --- 3.86E-07 2.70E-13 Te-125m 1.63E-06 - 8.17E-05 -- 1.04E-06 - 1.15E-N 8.01E-Il Te-127 2.76E-06 -- 1.38E-04 --- 5.97E-07 - 6.57E.05 4.59E-11 Te-127m 8.38E-07 -- 4.19E-05 --- 5.39E-07 - 5.93E-05 4.15E-11 Te-129 3.ME-05 -- 1.82E-03 - 1.36E-05 --- 1.50E-03 1.05E-09 Te-129m 3.42E-05 -- 1.71E-03 --- 2.15E-05 -- 2.36E-03 1.65E-09 Approved Design Material- Radiation Protection Page 12.2-28

Synt:m 80+ Dvign ControlDocument gs (h Table 12.2-12 CVCS Tank Inventories (Cont'd.) Maximum Values (pCi/gm) Volume Control Reactor and Equip. Holdup Boric Makeup Drain Acid Liquid Gas Liquid Gas Liquid Gas Liquid Liquid Isotope ( Cilgm) (pCi/ml) (pCi/gm) (pCi/ml) (pCi/gm) (pCi/ml) (pCi/gm) ( Ci/ml) Te-131 6.34 E-05 --- 3.17E-03 -- 1.09E-05 -- 1.20E-03 8.38E-10 Te-131m 1.62E-04 -- 8.10E-03 -- 4.18E-05 - 4.60E-03 3.21E-09 Te-132 1.13 E-02 --- 5.64E-02 -- 5.14E-04 - 5.65E-02 3.95E-08 Te-133 3.30E-05 -- 1.65E-03 -- 9.45E-19 - 1.04E-16 -- Te-133m 9.64 E-05 - 4.82E-03 --- 5.63E-18 - 6.20E-16 --- Te-134 1.28E-04 --- 6.39E-03 --- -- --- -- -- Cs-134 3.49E-02 -- 4.98E-02 - 1.39E-02 -- 1.53E+00 1.39E-05 Cs-136 6.75 E-03 -- 1.35E.02 --- 3.45E-03 -- 3.79E-01 3.45E-06 Cs-137 4.39E-02 - 8.77E-02 - 2.45E-02 - 2.69E+00 2.45E-05 Cs-138 1.04E-01 - 2.09E.01 - -- - -- -- Ba-137m 1.65E-03 -- 8.27E-02 --- 2.30E-02 --- 2.53E +00 2.30E-05 Ba-140 2.30E-05 --- 1.15E-03 --- 1.37E-05 --- 1.50E-03 1.05E-09 Ba-141 1.38E-07 - 6.92E-06 -- - --- - -

  ,s    La-140     7.74E-06          ---

3.87 E-04 --- 9.61E-06 - 1.06E-03 7.39E-10 I I La-141 3.26E-07 -- 1.63E-05 - 1.92E-10 - 2.11E-08 1.48 E-14 Ce-141 8.64E-07 --- 4.32E-05 -- 5.43E-07 -- 5.97E-05 4.18E-Il Ce-143 2.42E-06 - 1.21 E-04 -- 6.79E-07 - 7.47E-05 5.22E-11 Ce.144 2.20E-06 - 1.10E-04 -- 1.42E-06 -- 1.57E-04 1.10E-10 Pr-144 2.20E-06 - 1.10E-04 - 1.42E-06 --- 1.57E-04 1.10E-10 Kr-83m 5.18E-04 1.12E-02 - 1.04 E-05 2.64E-10 1.81E-08 2.90E-08 5.49E-13 Kr-85 1.20E + 00 9.89E-01 1.07 +00 1.74 E-01 8.81E-13 1.06E-04 9.69E-11 1.84E-15 l Kr-85m 9.97E-02 2.16E + 00 -- 2.49E-03 1.81E-10 4.31E-06 2.00E-08 3.78E-13 Kr-87 2.82E-02 6.26E-01 - 5.49E-04 -- 9.54 E-07 -- - Kr-88 1.58E-01 3.41 E +00 - 3.46E-03 -- 6.01 E-06 -- --- Kr-89 1.76E-04 2.18E-03 - 1.70E-06 - 2.%E-09 -- --- Kr-90 0.00E +00 7.20E-05 -- 5.56E-08 - 9.66E-11 - --- Xe-131m 1.38E +00 2.01E-01 5.64E-02 6.53E-01 1.10E-05 1.19E-04 1.21E-03 2.29E-08 Xe-133 8.10E+01 1.21E +03 1.71E +00 1.94 E +01 1.27E-03 6.78E-03 1.40E-01 2.66E-06 Xe-133m 7.03E-02 1.12E +00 7.%E-014 7.79E-03 7.46E-03 5.47E-06 8.21E-03 1.55E-07 l Xe-135 1.06E +00 1.60E +01 - 3.01E-02 6.11E-04 4.98E-05 6.72E-02 1.27E-06 Xe-135m 6.84E-03 1.02E-01 - 8.17E-05 1.65E-05 1.42E-07 1.82E-03 3.44 E-08 Xe-137 3.97E-04 5.97 E-03 - 4.65 E-06 - 8.09E-09 - --- Xe-138 5.80E-03 9.39E-02 - 7.55E-05 - 1.31 E-07 -- - Mn-54 3.80E-06 -- 1.90E-03 -- 1.89E-06 - 2.08E-05 -- Co-58 1.09E-05 --- 5.46E-03 - 5.37E-06 -- 5.91 E-05 -- Co-60 1.26E-06 - 6.30E-04 --- 6.30E-07 -- 6.93E-06 --- ['w/ Fe-59 Cr-51 7.12E-07 7.34E-06 -

                                        ---     3.56E-04 3.67E-03 3.47E-07 3.52E-06 3.82E-06 3.87E-05 Approved Design Material Radiation Protection                                                 (11/96) Page 12.2-29

Syst~m 80 + Design Contr:I Document Table 12.2-13 Shutdown Cooling System (SCS) Specific Source Strengths Maximum Values (MeV/ gram-sec) Decay Energy (Mev) Time (gm) 0.3 0.63 1.10 1.55 1.99 2.38 2.75 3.25 3.70 1 3.3(+4)i0 2.4(+5) 6.7(+4) 1.9(+4) 4.7(+3) 3.4( +2) 1.6(+2) 9.9(+ 1) 1.2( + 2) 10 2.5(+4) 1.2(+5) 2.9(+4) 7.5(+ 3) 2.2(+3) 2.9(+ 1) 6.7(-1) 6.2(-1) 8.9(-3) 100 1.8(+4) 4.4(+4) 6.3(+3) 2.4(+3) 3.5(+2) 2.2(+ 1) 2.7(-2) 8.7(-3) - l l 9 l O in Number in parentheses denoted powers of ten. Approved Design Material- Radiation Protection Page 12.2-30 l l

System 80+ Design Control Document (3 Table 12.2-14 Spent Fuel Pool Related Sources Radionuclide Specific Activities L) . Isotopes Nonnal Pool Design Basis Pool Cooling and Cleanup Cooling and Cleanup Water Water ( Ci/gm) Filter (pCi/ml) IX ( Ci/ml) ( Ci/gm) Volume Basis: - - 2.7 ft3 32 ft3 Br-83 --- 2.05E-12 -- --- Rr-86 - 1.96E-07 - - Sr-89 1.14E-06 3.81E-06 - 8.3E-03 Sr-90 1.01E-07 1.37E-07 -- 4.54E-04 St91 2.34E-07 1.90E-07 -- - 1 Y-90 -- 2.32E-08 --- 3.23E-08 Y-91 4.2SE-08 5.47E-07 -- 1.27E-03 - Y-93 9.65E-07 4.27E-09 - -- Zr-95 3.19E-06 5.98E-07 -- 7.71E-03 l Nb-95 2.24E-06 5.83E-07 - 4.llE-03 i Tc-99m 1.37E-07 7.81E-07 - - l Mo-99 3.15E-05 2.05E-04 - 3.84E-04 1 Ru-103 6.04E-05 2.01E-07 - 1.19E-01 { Ru-106 7.51E-04 7.58E-08 2.35E +00 7- - Q I-130 - 5.68E-08 - -- I-131 3.06E-04 2.41E-03 - 6.02E-01 ; I-132 6.60E-10 3.56E-10 - - I-133 2.18E-04 8.45E-04 -- 1.34E-10 I 135 1.29E-05 1.61E-05 - -- Sn-125 - 1.24E-09 - 4.75E-07 Sb-125 - 6.14E-10 - 2.00E-06 Sb-127 -- 1.06E-09 - 1.59E-08 j Te-125m - 3.33E-07 - 7.72E-04 l Te-127 - 1.67E-08 - --- Te 127m - 1.72E-07 - 4.72E-04 l Te-129 4.50E-17 2.06E-18 -- --- Te-129m 1.52E-06 6.82E-06 -- 1.22E-02 l Te-131m 3.90E-06 1.11E-05 - 2.19E-09 l Te-132 9.01E-06 1.53E-04 - 8.45E-04 Cs-134 2.23E-04 8.77E-04 - 2.82E +00 Cs-136 2.41E-05 2.13E-04 - 1.42E-01 Cs-137 2.95E-04 1.55E-03 - 5.11E +00 (O Ba-140 9.65E-05 4.29E-06 - 6.30E-06 La-140 8.72E-05 7.05E-07 - 1.19E46 Approved Design Material Radiation Protection (11/96) Page 12.2-31

Syst~m 80 + Design ControlDocument Table 12.2-14 Spent Fuel Pool Related Sources Radionuclide Specific Activities (Cont'd.) Isotopes Normal Pool Design Basis Pool Cooling and Cleanup Cooling and Cleanup Water Water Filter IX (pCilgm) (pCi/gm) ( Ci/ml) (pCi/ml) Volume Basis: - - 2.7 ft3 32 ft3 La-141 -- 1.45E-11 -- -- Ce-141 1.20E-06 1.72E-07 - 2.10E-03 Ce-143 8.07E-06 1.83E-07 --- 7.27E-09 Cc-144 3.25E-05 4.55E-07 - 1.00E-01 Mn-54 1.27E-05 7.45E-06 4.67E-01 2.57E-02 Co-58 3.60E-05 2.11E-05 1.06E-00 5.83E-02 Co-60 4.23E-06 2.48E-06 1.65E-01 9.07E-03 Fe-59 2.32E-06 1.36E-06 5.72E-02 3.15E-03 Cr-51 2.34E-05 1.37E-05 4.35E-01 2.39E-02 O O Approved Design Material- Radiation Protection Page 12.2 32

Syst:m 80+ Design Contr IDocument i r Table 12.2-15 Turbine Building Sources Radionuclide Specific Activities Spent Blow- Condensate Resin Neutrali-SGm down Blowdown Condensate Second Decanting zation Water Condensate Steam Filter IX' Lead IXA IXm TankW TankM Isotope ( Ci/gm) (pCi/gm) (pCi/gm) ( Ci/ml) (pCl/m!) (pCi/ml) (pCl/ml) 4Ci/ml) 4Ci/ml) Volume 1 ff 30 ft' 275 ff 280 ft' 555 ft' 92,000 Basis gallons

Sr-89 1.2E-08 5.9E-11 5.9E-I l -

3.7E-04 0.0E +00 6.58E-06 3.32E-06 1.20E-07 l Sr-90 1.0E-09 5.2E-12 5.2E-12 -- 1.6E-04 0.0E +00 2.86E-06 1.44 E-06 5.21E-08 l Sr-91 0.0E+00 0.0E+00 0.0E +00 - 1.lE-05 0.0E+00 0.0E+00 0.0E +00 0.0E +00 l Y-91 4.4 E-10 2.2E-12 2.2E-12 - 1.6E-05 0.0E +00 2.82E-07 1.42E-07 5.14E-09 Y-91m 0.0E +00 0.0E+ 00 0.0E +00 - 0.0E +00 0.0E +00 0.0E+00 0.0E +00 0.0E +00 l Y-93 1.9E-07 9.4E-10 9.4E-10 -- 4.7E-05 0.0E+00 8.31E-07 4.19E-07 1.51E-08 Zr-95 3.4E-08 1.7E-10 1.7E-10 - 1.3E-03 0.0E +00 2.39E-05 1.21E-05 4.35E-07 l Nb-95 2.3E-08 1.2E-10 1.2E-10 --- 5.0E-04 0.0E+00 8.95E-06 4.52E-06 1.63E-07 l Tc-99m 1.6E-07 7.9E-10 7.9E-10 -- 2.5E-05 0.0E +00 4.38E-07 2.21E-07 7.98E-09 l Mo-99 4.9E-07 2.5E-09 2.5E-09 --- 8.5 E-04 0.0E +00 1.51E-05 7.62E-06 2.75E-07 l bV Ru-103 6.5E-07 3.2E-09 3.2E-09 - 1.6E-02 0.0E+00 2.84E-04 1.43E-04 5.17E-06 l Ru-106 7.8E-06 3.9E-08 3.9E-08 --- 8.9E-01 0.0E + 00 1.58E-02 7.97E-03 2.88E-04 l I-131 2.4 E-06 2.4E-08 2.4 E-08 -- 1.2E-02 0.0E+00 4.32E-04 2.18E-04 7.87E-06 1-132 1.3 E-07 1.3E-09 1.3E-09 --- 2.0E-04 0.0E +00 7.04E-06 3.55E-06 1.28E-07 I-133 3.3E-06 3.3E-08 3.3E-08 -- 3.3E-03 0.0E +00 1.16E-04 5.85E-05 2.llE-06 1-135 7.6E +06 7.6E +04 7.6E+04 --- 1.3E-03 0.0E +00 4.70E-05 2.37E-05 8.56E-07 Te-129 2.5 E-07 1.2E-09 1.2E-09 -- 7.5E-06 0.0E +00 1.33E-07 6.71E-08 2.42E-09 l Te-129m 1.6E-08 8.2E-11 8.2E-11 --- 3.4E-04 0.0E +00 6.06E-06 3.06E-06 1.10E-07 l Te-131 3. l E-08 1.5E-10 1.5E-10 -- 3.3E-07 0.0E +00 5.89E-09 2.97E-09 1.07E-10 l Te-131m 9.9E-08 4.9E-10 4.9E-10 -- 7.7E-05 0.0E +00 1.37E-06 6.91E-07 2.50E-08 l Te-132 1.3E-07 6.6E-10 6.6E-10 -- 2.7E-04 0.CE +00 4.73E-06 2.39E-06 8.61E-08 l Cs-134 7.5E-07 3.8E-09 3.8E-09 -- 9.2E-02 1.76E-03 9.96E-05 4.84E-01 1.75E-02 l Cs-136 8.9E-08 4.4E-10 4.4E-10 -- 6.5E-04 1.26E-05 7.10E-07 3.47E-03 1.25E-04 l Cs-137 1.0E-06 5.0E-09 5.0E-09 -- 1.4E-01 2.73E-03 1.54E-04 7.51E-01 2.71E-02 l La-140 1.8E-06 8.8E-09 8.8E-09 -- 1.8E-03 0.0E + 00 3.26E-05 1.64E-05 5.94E-07 l Cc-141 1.3 E-08 6.4E-11 6.4 E-Il - 2.6E-04 0.0E +00 4.57E-06 2.31E-06 8.32E-08 l Ce-143 1.9E-07 9.3E-10 9.3E-10 - 1.6E-04 0.0E +00 2.82E-06 1.42E-06 5.14E.08 l { Cc-144 3.4E-07 1.7E-09 1.7E-09 - 3.5E-02 0.0E + 00 6.25E-04 3.15E-04 1.14E-05 l l Approved Design Material Radiation Protection (11/96) Page 12.2 33 ]

System 80+ D: sign C ntrolDocument l Table 12.2-15 Turbine Building Sources Radionuclide Specific Activities (Cont'd.) Spent Blow- Condensate Resin Neutrali-SG"' down Blowdown Condensate Second Decanting zation Water Condensate Steam Filter IX Lead IXA IXW TankH1 Tankm Isotope ( Ci/gm) (pCi/gm) (pCi/gm) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) Volume 1 ft' 30 ft' 275 ft' 280 ft' 555 ft' 92,000 Basis gallonsW l Kr-85 - - 1.2E-09 -- -- -- - -- -- l Kr-85m - -- 2.3 E-08 - -- -- --- - -- l Kr-87 -- - 2.5E-08 -- .- - - - - - -- l Kr-88 - --- 4.5E-08 --- -- - - - -- --- -- l Xe-131m -- -- 7.2E-09 --- --- -- --- -- -- l xe-133 - --- 4.7E-08 - --- --- --- - - l Xe-133m - - - -- 2.6E-09 2.6E-09 --- -- --- -- -- l Xe-135 -- -- 9.3E-08 9.3E-08 --- - -- -- - l xe-135m --- -- 2.5E-08 -- --- - -- - - l Xe-137 -- - 6.7E-09 -- - -- --- -- - l Xe-138 --- -- 2.3E-08 - --- - --- - - l Mn-54 1.4 E-07 6.9E-10 6.9E-10 4.6E-01 1.5E-03 0.0E +00 2.64E-04 1.33E-04 4.81E-06 Co-58 4.0E-07 2.0E-09 2.0E-09 5.3E-01 1.7E-03 0.0E +00 3.06E-04 1.54 E-04 5.57E-06 l l Co-60 4.6E-08 2.3E-10 2.3 E-10 2. l E-01 6.9E-04 0.0E+00 1.22E-04 6.15E-05 2.22E-06 l Fe-59 2.5E48 1.3E-10 1.3 E-10 2.2E-02 7.0E-05 0.0E+00 1.24E-05 6.26E-06 2.26E-07 l Cr-51 2.7E-07 1.4E-09 1.4E-09 1.5E-01 4.7E-04 0.0E +00 8.37E-05 4.22E-05 1.52E-06 Ul Also source assumed for SG Blowdown and SG Drain Tank Fluid systems. A The lead bed of the condensate cleanup system plishers is a cation bed ion exchanger. m The second bed of the condensate cleanup system polishers is a mixed bed ion exchanger. Hi The activity in the spent resin decanting tank is the sum of the lead and second bed condensate cleanup system ion exchangers. m The activity in the neutralization is equal to the total activity accumulated on tne condenste polishers prior to regeneration. All activity is assumed to be removed during regeneration and is collected in the neutralization tank. W 92,000 Gallons represents the useable volume (0.8

115,000 gallons).

Approved Design Material- Radiation Protection (11/96) Page 12.2-34

_m ._ _ _ ~ . . _ _. _ - . _ _ _ _ __ _ __. Syst:m 80+ Design contet Document j . Table 12.2-16 Nuclear Annex Sources Process Gas Specific Activities Isotope Process Header 10 (pCi/ml) Charcoal Bed (pCi/ml) Volume Basis: 450 ft3 ! Kr-83m 9.14E+02 1.05E-02 Kr-85 3.60E+01 1.36E+02 Kr-85m 9.05E+00 2.51E+00 Kr-87 7.04E+00 5.54E-01 Kr-88 1.98E+01 3.49E+00 Kr-89 5.27E-01 1.72E-03 Kr-90 1.01E-01 5.60E-05 Xe-131m 4.23E+01 6.16E +02 Xe-133 2.56E+03 1.97E +04 Xe-133m 2.37E +00 7.93E+00 Xe-135 5.36E+0! 3.04E+01 Xe-135m 5.12E+00 8.23E-02 Xe-137 1.19E +00 4.69E-03 Xe-138 4.35E+00 7.61E-02 O IU Also reflects gas stripper gas phase source. Approved Design Material- Radiation Protection Page 12.2-35

Srtem 80+ Design C'ntrol Document Table 12.2-17 Radwaste Building Sources Liquid Waste Tank Specific Activities Isotope Equipment Waste Floor Drain Laundry and Chemical Waste Waste Monitor Tank (pCi/ml) Tank (pCi/ml) Hot Shower Tank ( Ci/ml) Tank (pCi/ml) Tank (pCi/ml) Br-83 5.3 E-04 5.3E-05 - 5.3E-04 5.3 E-07 Br-84 5.7E-03 5.7E-04 - 5.7E-03 5.7E-06 Br-85 6.8E.04 6.8E-05 1 - 6.8E44 6.8E-07 Rb-88 6.7E-01 6.7E-02 -- 6.7E-01 6.7E44 Rb-89 3.5E-02 3.5E-03 - 3 5E-02 3.5E-05 Sr-89 9.4E-04 9.4E-05 1.2E-07 9.4E-04 9.4E-07 Sr-90 3.3E-05 3.3E-06 3 1.7E-08 3.3E-05 3.3E-08 St-91 1.4E-03 1.4E-04 --- 1.4E-03 1.4E-06 Y-90 9.4E-06 9.4 E-07 - 9.4E-06 9.4E-09 Y-91 1.4E-04 1.4E-05 1.1E-07 1.4 E-04 1.4E-07 Y-91m 8.1E-04 8. l E-05 --- 8.1E-04 8. l E-07 Y-93 3.3E-05 3.3E-06 - 3.3E-05 3.3E-08 Zr-95 1.5E-04 1.5E-05 1.5E-06 1.5E-04 1.5E-07 Nb-95 1.5 E-04 1.5E-05 2.5 E-06 1.5E-04 1.5E-07 Tc-99m 4.7E-02 4.7E-03 -- 4.7E-02 4.7E-05 Mo-99 8.1E-02 8.1E-03 8.0E-08 8.1E-02 8.1 E-05 Ru-103 5.0E-05 5.0E-06 3.9E-07 5.0E-05 5.0E-08 Ru-106 1.8E-05 1.8E-06 1.2E-05 1.8E-N 1.8E-08 1-131 7.1 E-01 7.1E-02 2. l E-06 7. LEE 7.lE 04 I-132 1.9E-01 1.9E-02 --- 1.9E-01 1.9E-04 I-133 1.0E-00 1.0E-01 --- 1.0E-00 1.0E-03 1-134 1.2E-01 1.2E-02 - 1.2E4)1 1.2E-04 I-135 5.7E-01 5.7E-02 - 5.7E-01 5.7E-04 i Te-129 1.8E-03 1.8E-04 --- 1.RE-03 1.8E-06 Te-129m 1.7E4 1.7E-04 - 03 1.7E-06 Te-131 3.2E-Oi 3.2E-04 --- 3..!E-03 3.2E-06 Te-131m 8. lE-03 P.1E-04 --- 8.1E-03 8.1E-06 Te-132 5.7E-02 . 7E-03 - 5.7E-02 5.7E-05 ' Te-134 6.4E-03 6.4E-04 -- 6.4E.03 6.4E-06 Cs-134 5.0E-02 5.0E-03 1.5 E-05 5.0E-02 5.0E-05 1 Cs-136 14.35 1.4E-03 5.0E-07 1.4 E-02 1.4E-05 Cs-137 i.T. .-02 8.8E-03 2.1 E-05 8.8E-02 8.8E-05 Cs-138 2.!E-01 2. lE-02 - 2.1E-01 2.lE-04 Ba-137m 8.3E-02 8.3E-03 - 8.3 E-02 8.3E-05 Ba-140 1.2E-03 1.2E-04 1.2E-06 1.2 E-03 1.2E-06 Lal40 3.9E-04 3.9E-05 - 3.9E-04 3.9E-07 Ce-141 4.3E-05 4.3E-06 3. lE-07 4.3 E-05 4.3 E-08 Ce-143 1.2E-04 1.2E-05 - 1.2E-04 1.2E-07 ApprovedDesign Afsterial Radiation Protectmn Page 12.2-36 1 i

Sv? tem 80+ Design controlD cument } Table 12.2-17 Radwaste Building Sources Liquid Waste Tank Specific Activities (Cont'd.) Isotope Equipment Waste Floor Drain Laundry and Chemical Waste Waste Monitor Tank (pCi/ml) Tank (pCl/ml) Hot Shower Tank (pCl/ml) Tank (pCi/ml) Tank (pCl/ml) Cc-144 1.1E-04 1.1E-05 5.2E-06 1.1E-04 1.1E-07 , Pr-144 1.lE-04 1.lE-05 - 1.1E-04 1.1E-07. ! Mn-54 1.9E-03 1.9E-04 5.lE-06 1.9E-03 1.9E-06 Co-58 5.5E-03 5.5E-04 1. lE-05 5.5E-03 5.5E-06 Co-60 6.3 E-04 6.3E-05 1.9E-05 6.3E-04 6.3E-07 Fe-59 3.6E-04 3.6E-05 2.9E-06 3.6E-04 3.6E-07 Cr-51 3.7E-03 3.7E-04 6.3E-06 3.7E-03 3.7E-06 ) P-32 - - 2.4E-07 - --- Fe-55 - -- 9.7E-06 --- -- l l Ni-63 - . 2.3E-06 -- -- Ag-!10m - - 1.6E-06 -- - Sb-124 - - 5.8E-07 -- -- t . () : l l l l O l Approved Desfors Noterial Ro6etiers Protectiers Page 12.2-37

System 80+ De*ign C*ntrol Document Table 12.2-18 Radwaste Building Sources Liquid Waste Process Equipment Isotope Waste Process Filter Waste Process IX Detergent Waste Filter (pCihn!) (pCi/ml) (pCi/ml) Volume Basis 7,.1 ft' 30 ft' O.1 ft' Br-83 --- 1.18E-04 - Br-84 -- 2.76E-04 - Br-85 - 3.02E-06 - Rb-88 --- 1.82E-02 - Rb-89 --- 8.46E-04 - S$89 --- 7.53E-02 -- Sr"-90 --- 4.60E-03 -- Sr-91 -- 1.25E-03 - Sr-92 - 0.00E +00 - Y-90 - 5.55E-05 - Y-91 -- 1.16E-02 - Y-91m - 6.19E-05 - Y-92 - 0.00E +00 - Y-93 - 2.94E-05 -- Zr-95 - 1.32E-02 - Nb-95 - 9.50E-03 -- Tc-99m -- 2.60E-02 Mo-99 -- 5.00E-01 -- Rh-103m -- 0.00E-00 - Ru-10? - 3.51E-03 - Ru-106 -- 2.36E-03 - 1-131 -- 1.27E +01 - I 1-132 - 4.05E-02 - l-133 - 1.98E+00 -- 1-134 -- 4.84E-03 --- 1-135 - " ' .55E-0, - Te-129 -

                                                                . 94E44                 --                 i Te-129m                              -

1.08E-01 -- Te-131 - 1.22E-04 --- Te-131m - 2.25E42 - Te-132 --- 4.07E-01 - Te-134 - 4.12E-04 - - - Cs-134 - 6.70E+00 - - -

Cs-136 -- 3.86E-01 -- l Cs-137 - 1.23E+01 - - - 1 Cs-138 - 1.08E-02 --- Ba-137m -- 3.25E-04 Approved Design Material- Radietion Protection Page 12.2 38

                          . _ .    -       . _ -        . . _ ~  .. - - .-                                 __  _ - . . .

4 Sy~ tem 80 + Design CrntrolDocument Table 12.2-18 (] Radwaste Building Sources Liquid Waste Process Equipment (Cont'd.) Isotope Waste Process Hiter Waste Process IX Detergent Waste Filter (pCi/ml) (pCi/ml) (pCi/ml) Volume Basis 0.1 ft' 30 ft' O.1 ft' Ba-140 -- 3.24E-02 -- La-140 -- 1.44E-03 - Cc-141 --- 2.67E-03 -- Cc-143 -- 3.69E-04 -- Cc-144 --- 1.38E-02 --- Pr-144 -- 2.93E-06 --- M mS4 7.11E+01 4.83E-03 5.78E-01 , Co-58 1.50E+02 1.02E-02 7.09E-01 Co-60 2.56E+01 1.74 E-03 2.53E+00 i Fe-59 7.86E+00 5.35E-04 1.37E-01 Cr-51 5.97E+01 4.06E-03 1.88E-01

O L) i i l i k Approved Design Material Radiation Protection Pope 12.2-39

Sy t~m 80 + Design Control Drcument Table 12.2-19 Radwaste Building Sources Solid Waste Process Equipment h Isotope IIA Spent Resin Tank LA Spent Resin Tank (pCi/ml) ( Ci/ml) Br-83 3.30E-02 1.18E-04 Br-84 7.70E-02 2.76E-04 Br-85 8.41E-04 3.02E-06 Rb-86 6.99E-02 4.80E-04 Rb-88 2.57E+00 1.82E-02 Rb-89 1.19E-01 8.46E-04 Sr-89 3.38E-03 7.53E-02 Sr-90 2.90E +01 4.60E-03 Sr-91 6.02E +00 1.25E-03 Sr-92 3.46E-01 --- Y-90 1.53E-02 5.55E-05 Y-91 4.81 E +00 1.16E-02 Y-91m 1.71E-02 6.19E-05 Y-93 8.12E-03 2.94 E-05 Zr-95 5.84E+00 1.32E-02 Nb-95 3.14E +00 9.50E-03 Tc-99m 7.17E +00 2.60E-02 Mo-99 1.38E + 02 5.00E-01 Ru-103 1.21E +00 3.51E-03 Ru-106 2.31 E+00 2.36E-03 I-130 6.58E-02 2.36E-04 I-131 3.53E+03 1.27E+01 1-132 1.13E +01 4.05E-02 1-133 5.52E +02 1.98E+00 l I-134 2.74 E +00 9.84E-03 1-135 9.89E+01 3.55E-01 Sn-125 2.02E-03 7.32E-06 Sb-125 2.37E-02 2.01E-05 Sb-127 S.58E-04 3. llE-06 Te-125m 2.89E +00 6.98E43 Te-127 3.31E42 1.20E-04 Te-127m 2.62E+00 4.46E-03 Tc-129 5.34 E-02 1.94E-04 Te.129m 3.51 E +01 1.08E-01 Tc-131 3.37 E-02 1.22E-04 Te-131m 6.20E +00 2.25E-02 Te-132 1.12E+02 4.07E-01 Te-133 8.'.75 03 3.17E-05 Te-133m 1.13E-01 4.llE-04 Approved Design Material- Radiation Protection Page 12.2-40

Sy' tem 80+ De-ign C'ntrolDocument b)l q Table 12.2-19 Radwaste Building Sources Solid Waste Process Equipment (Cont'd.) Isotope HA Spent Resin Tank LA Spent Resin Tank (pCihn!) ( Ci/ml) l l Te-134 1.14 E-01 4.12E-04 l Cs-134 3.88E+03 6.70E+00 Cs-136 5.48E +01 3.86E-01 Cs-137 8.19E+03 1.23E +01 Cs-138 1.51E + 00 1.08E-02 Ba-137m 8.97E-02 3.25E-04 Ba-140 9.01E +00 3.24E-02 Ba-141 5.37E-05 1.95E-07 La-140 3.97E-01 1.44 E-03 l , La-141 1.64 E-03 5.94E-06 Ce-141 8.58E-01 2.67E-03 Ce-143 1.02E-01 3.69E-04 Ce-144 1.26E+01 1.38E-02 i Pr-144 8.09E-04 2.93E46 Mn-54 2.07E+01 4.83 E-03 l Co-58 2.35E+01 1.02E-02 l ! ,O Co-60 1.08E +01 1.74E-03 l Fe-59 9.71E-01 5.35E-04 Cr-51 6.23E+00 4.06E-03 l l l I 1 l l

Approved Design Material- Radiation Protection Page 12.2-41 i

Sy-tem 80 + Design C?ntrolDocument Table 12.2-20 NUREG-1465 Post-accident Shielding Source Termstil g Gap Release Fuel plus Gap Release Liquid Containing Systems: 5% of Core Groupt2) 40% of Core Group 2 isotopes 2 and 3 isotopes 30% of Core Group 3 isotopes 15% of Core Group 4 isotopes 3% of Core Group 5 isotopes 0.8% of Core Group 6 isotopes 0.2% of Core Group 7 isotopes 1% of Core Group 8 isotopes 4% of Core Group 9 isotopes Containment Airborne: 5% of Core Group [2) 100% of Core Group 1 isotopes 1,2 and 3 isotopes 40% of Core Group 2 isotopes 30% of Core Group 3 isotopes 15% of Core Group 4 isotopes 3% of Core Group 5 isotopes 0.8% of Core Group 6 isotopes 0.2% of Core Group 7 isotopes 1% of Core Group 8 isotopes 4% of Core Group 9 isotopes O IU The above source terms represent the initial releases from the core into the containment atmosphere and sump water. [2] Group 1 : Xe, Kr, Rn Group 2 : 1 Br Group 3 : Cs, Rb Group 4 : Te, Sb, Se, Ag, As, Cd, Ga, Ge, In, Sn, Zn Group 5 : Sr. Ra Group 6 : Co, Ru, Rh, Pd, Mo, Tc Group 7 : Am, Cm, Gd, Ho, La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Tb, Dy, Y Group 8 : Ce, Pu, Np, Th, U, Pa, Cf. Ac Group 9 : Ba Approved Design Material- Radiation Protection Page 12.2-42

Systrm 80+ Design crntrolDocument rs U 1 l l 5 4 Notet effect of the Shielding' Refueling Machine is included 3 - T l E 2 - 2 E i I

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             .3
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                ~3   ~

114.5 144.5 174.5 204.5 234.5 264.5 294.5 324.5 354.5 384.5 Distance from Top of Active Fuel (em) O 72 hr-post-shutdown + 200 hr 6 1000 hr A 2000 hr

  ;     Maximum Spent Fuel Assembly Dose Rates Versus Axial Distance in                                Figure 12.2-1
   \    Refueling Pool Approved Design Material- Radiation Protection                                                               Page 12.2-43

                      -     -                                    ._.       --              .=.

Syst:m 80+ Design centrolDocument l 9 9 8 l 7 . C6 - E 5 - e

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                                                                                      ^

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      -2        e        i        i        i     i     i     i l           i       e       i 53.4       1    113.4        l   173.4   1   233.4     1   293.4   1    353.4      I 83.4             143.4        203.4         263.4       323.4           383.4 Distance from Edge of Active Fuel (cm)

D 72 hr-post-shutdown + 200 hr o 1000 hr a 2000 hr Maximum Spent Fuel Assembly Dose Rates Venus Radial Distance in Figure 12.2-2 Refueling Pool Approved Design Material . Radiati'" Protection Page 12.2-44

Sy~t m 80 + oesign controlDocument I

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Q} 12.3 Radiation Protection Design Features i 12.3.1 Facility Design Features i The System 80+ design incorporates ALARA principles per Regulatory Guide 8.8 and 8.10 to minimize i the onsite exposure to plant personnel and operators during normal operation and maintenance activities. Section 12.1.2.1 details ALARA principles incorporated into the plant layout, component locations, and i materia! selection. Plant layout is designed so that access to higher radiation areas is from lower radiation i areas. This is ensured through specific design features, such as location of radiation systems and equipment and pipe routing. The following section details specific design features to ensure operational

and maintenance exposure is ALARA I 1

12.3.1.1 General Arrangement Design Features 1

Location of Radioactive Systems and Equipment Nonradioactive systems are separated from radioactive systems. This helps control the spread of contamination and minimize the necessity for routing piping containing radioactive fluids or slurries through personnel corridors. This also facilitates radiation area access control.

Radioactive equipment are separated into compartments whenever possible. Equipment is compartmentalized based on frequency of access required, operational characteristics, and radiation level. For example, ion exchangers containers resin beads are typically located in a O separate compartment from active components, such as pumps and valves. Valves are typically 3 located in valve galleries. Ion exchangers are located in pits with their associated spent resin j service tanks located directly below the ion exchanger to minimize the pipe lengths and the j general area radiation. The compartment walls provide shielding which enables personnel to perform operation and maintenance activities in a lower radiation area. l

Pipe Routing Pipe lengths of radioactive systems are minimized by locating interfacing systems in close proximity. Piping for these systems are then routed through shielded pipe chases. The number of active components located in pipe chases are minimized to reduce the frequency of access ,

required into the pipe chase for maintenance activities.

Spacing The System 80+ Standard Design ic designed to provide adequate spacing around equipment for easy access of equipment for maintenance and inspection. This includes provisions for adequate laydown area or equipment pull area, as well as transport paths for removal or replacement of equipment. Rigging and lifting equipment are also provided to facilitate the removal, transport, or replacement of equipment or portable shielding during maintenance activities.
Hot Tool Cribs and Hot Machine Shops p Hot tool cribs are located in low radiation areas adjacent to maintenance areas to minimize

(") waiting times in high radiation areas, to help pvent the spread of contamination, and to decrease the amount of decontamination work required to be perfonned. This reduces the radioactive wastes generated and personnel exposure. Approved Desue n Atatorint. Radiation Protection Pope 12.3-1

i Srtem 80+ 0: sign crntrolD cument The provision of a hot machine shop adjacent to the equipment hatch enables personnel to remove l equipment from containment and perform maintenance in a lower radiation area. Access to the I hot machine shop is also provided from the truck bays and maintenance areas for ease of i equipment movement. l

Staging Areas l

Large staging areas inside and outside the equipment hatch and personnel airlocks allow pre- 1 staging prior to the start of an outage, as well as provide space for efficient radiation controls for i moving equipment in and out of containment. l

Personnel Decontamination and Change Areas 1

Two personnel decontamination areas are provided in the System 80+ design. One is located ; within the radiation access control area (RCA) and the other is located adjacent to RCA access I point. Protective clothing, respirators, shower and toilet facilities, lockers, and containers for contaminated clothing are provided in these areas. Change areas are located near airlocks to minimize personnel traffic flow, distance travelled, and the potential for the spread of contamination.

Radiation Control Area (RCA)

The System 80+ design provides for a single point access into the RCA on elevation 91+9; however cruergency egress is provided on all elevations. The access area to the RCA provides l a flexible and adaptable layout, a large area (40' x 100') sufficient to accommodate outage work , crews and enhance the availability of immediate interaction with radiation protection personnel stationed at this point.

Accessways and Entrances to High Radiation Areas Labyrinths or shield doors are provided at the entrances to high radiation areas to minimize the exposure due to scatter and streaming of radiation through entrances.

Shield plugs are provided as necessary to provide shielding during normal operation for adjacent corridors. These shield plugs are removable to permit components, such as heat exchangers, and their internals to be pulled during maintenance activities. High radiation areas are provided with locked doors to prevent inadvertent access by plant personnel. 12.3.1.2 Equipment and System Design Features for Control of Onsite Exposure System 80+ specifies the use of more reliable and simplistic equipment. This reduces the frequency of maintenance and the radiation exposure to plant personnel. The following section discusses equipment design characteristics utilized in radioactive systems.

Pumps

1. Pumps and associated piping are flanged to facilitate pump remova! w a lower radiation area for maintenance or repair. Pump internals are also removaMe.

9

Approved Design Material- Radiation Protection Page 12,3-2

          . _   _     _        __      .          _._.._...____._.-.____....-.m                                   .  -____m           .-._.m._ . ,

. ~

i h Sy? tem 80+ Cz'.n ControlDocument

2. All pump casings are provided with drain connections to facilitate decontamination. The drain connection are free of internal crevices to minimize accumulation of radioactive

_ corrosion products (crud).

3. Pump seals are easily serviceable without removal of the entire pump or motor. The 4

reactor coolant pump seals are a cartridge type to facilitate removal for maintenance or repair.

Ion Exchangers (Demineralizers)

                               .1.        Ion exchangers are designed for complete drainage.

2. ' Spent resin removal is designed to be done remotely by hydraulic flushing from the vessel to the Solid Warte Management System (SWMS).

a. ]. 3. Piping, strainers and resin screens are flushable so that all spent resin is removed in the l flush mode. L . :

4. Fresh resin addition is accomplished from a low radiation area above the shielded t

compartment housing the ion exchanger. ] 1 4 ( 5. Internal crevices are minimized to prevent accumuhtion of radioactive crud. i- ;

6. . Ion exchanger manways are easily accessible. Internal components are easily removed j(

                  ,                      through manways requiring muumal disassembly.

[ 3

Liquid Filters .

j 1. Filter housings are provided with vents connections and are designed for complete i j drainage. i ' , 2. Filter housings are designed with a minimum of internal crevices to minimize the i } . accumulation of radioactive crud. t

3. Filter housings and cartridges are designed to permit remote removal of illter elements. $
- Cartridge filter seals are an integral part of the filter cartridge so that seal removal is ;

i accomplished during cartridge removal. ! [

                             ~ 4.       Cartridge filter housing closure heads are designed to swing free for the unobstructed g                                        removal of the cartridge.                                                                                     !

Tanks i

1. Tanks are designed for complete drainage; free of internal crevices, and pockets. The l drain line is connected to the bottom. ,

I

2. Tanks are provided with at least one of the following means of decontaminating the tank

    .                                   internals:
                                                        . Ample space is provided to permit decontamination of the tank manway.

4-1- AMnownf DenQn Adeseniel- Re6\ntion J400echien Page 12.3-3 } i' J k 1 . ._ ,_. _ _- -

System 80+ Design ControlDocument Internal spray nozzles are provided on potentially highly contaminated tanks for internal decontamination. i i Back l lush capability is provided for tank inlet screens.

3. Tanks are designed d'i a convex or sloped bottom to facilitate drainage and minimize ;

the accumulation of cruo. l

4. Tanks are provided with vents to facilitate the removal of potentially radioactive gases during maintenance.

5. Non pressurized tanks are provided with overflows, routed to a floor drain pump or other suitable collection point to avoid spillage of radioactive fluids onto the floor or ground.

The floor drain system is connected to the Liquid Waste Management System for funher processing prior to release to the environment. e Valves

1. The following discussion summarizes valves specifications that minimize valve leakage, as well as extend valve design life.

Except for modulating valve applications, packless valves are used on all valves two inches and under in diameter. Modulating valves and valves greater than two inches in diameter use live loading of the packing by conical spring washers or equivalent means to maintain a compressive force on the packing where possible. Double stem packing with a leak-off between the packing is used for valves four inches and larger, as well as normally open valves two to four inches in diameter. Stem leakage is piped to an appropriate drain sump or tank. Valves utilizing stem packing are provided with backseat capability. Radiation resistant seals, gaskets, and elastomers are utilized, when practicable, to extend the design life and reduce maintenance requirements. Valves located in high radiation areas will be equipped with reach rods or motor i operators to minimize radiation exposure. l

2. Fully ported valves are used to minimize internal accumulation of crud.

3. Valves requiring removal during maintenance and inspection activities are flanged.

4. Internal valve surfaces are designed free of crevices to minimize the accumulation of crud.

5. Valve wetted parts are made of austenitic stainless steel or corrosion resistant material, l

Approved Design Material- Radiation Protection Page 12.3 4

r i~ System 80+ Derkn contratDocument

 '.                  6.           Valves are designed so that they may be repacked without removing the yoke or j'                                 topworks.                                                                                                     t

Piping and Penetrations :

1. There is no field run piping.

2. Resin and concentrate piping is designed as follows: 4 The length of pipe runs are minimized.

Piping is routed through shielded pipe chases whenever possible to minimize routing through personnel access corridors. '

Large diameter piping (> 5 pipe diameters) is utilized to minimize the potential { for clogging during slurry or resin transfer without violating minimum flow j requirements. . The number of pipe fittings (e.g., elbows, tees, etc.) are minimized to reduce the i potential for radioactive crud accumulation. ) Low points, deadlegs, and vertical pipe runs are minimized. 1 7 Pipe runs are sloped and gravitational flow is used where practicable. Crevices on piping internal surfaces are minimized by the use of butt welds instead of socket welds. Socket welds are known crud traps in radioactive systems. The use of butt welds , generally result in smoother internal surfaces reducing cmd buildup. Therefore, ! butt welds will be used in lieu of socket welds, whenever possible, to minimize crud traps in piping in radioactive systems. Flushing capability is provided to facilitate decontamination of piping. Penetrations are located so that the source and the penetration are not in a direct line of sight. This minimizes the potential for personnel exposure due to streaming. e Heat Exchangers

1. Heat exchangers are designed with vents and for complete drainage.

2. Internal wetted surfaces.are designed crevices free to minimize the potential for accumulation of radioactive crud on internal surfaces.

3. Corrosion resistant materials are utilized to minimize the need for replacement and reduce the frequency of maintenance required.

_ _. _ _ ~. rma

1 Srt m 80+ De:ign controlDocument e Reactor Vessel Head Vent A vent nozzle and line is provided on the reactor vessel head. Utilization of this design feature ! will allow a reduction of exposure during the head removal process by minimizing the gases discharged directly to the containment atmosphere while the head is being removed. e Reactor Coolant System Leakage Control Exposures from airborne radionuclides to personnel entering the containment will be minimized by controlling the amount of reactor coolant leakage to the containment atmosphere. Examples of such controlled leakage are listed below:

1. Primary pressurizer safety and safety depressurization system valve leakage is directed to the IRWST.

2. Valves larger than 2" in diameter are provided with a double-packed stem with an intermediate lantern ring with a leak-off connection to the Reactor Drain Tank.

3. Instrumentation is provided to detect abnormal reactor coolant pump seal leakage. The reactor coolant pumps are equipped with two stages of seals plus a vapor or backup seal as described in Section 5.5. The vapor or backup seal with prevent leakage to the containment atmosphere and allow sufficient pressure to be maintained to direct the controlled seal leakage to the Volume Control and Reactor Drain Tanks. The vapor seal is designed to withstand full Reactor Coolant System pressure in the event of failure of any or all of the two prirnary seals.

e Refueling Equipment

1. All spent fuel transfer and storage operations are designed to be conducted underwater to insure adequate shielding and to limit the maximum continuous radiation levels in working areas.

2. Equipment is designA to prevent the fuel from being lifted above the minimum safe water depth, thereby limiting personnel exposure and avoiding fuel damage.

3. The equipment design limits the possibility of inadvertent fuel drops which could cause fuel damage and personnel exposure.

4. The refueling equipment design will facilitate the transfer of new and spent fuel at the same time to reduce overall fuel handling time and, therefore, personnel exposures during refueling.

5. Underwater cameras are used to facilitate safe handling and visual control, thus minimizing errors and potential exposures.

6. Portable hydraulic cutters are provided to cut expended Control Element Assemblies and in-core instrumentation leads. The cutters allow underwater handling of these items.

7. Equipment is provided to allow for the underwater inspection of fuel elements to determine leakers.

Apsvoved Design Material- Radiation Protection Page 12.3-6

Sy~ tem 80+ Design C'ntrolDocument ^

  ,m t

In-Service Inspection Equipment v

Inspection of the reactor coolant pressure boundary can be done with remote equipment to minimize personnel exposures. ((A detailed discussion of the In-Service Inspection Program is provided by the COL Applicant.))l

Remote Instrumentation All systems (e.g., RCS, CVCS, LWMS, etc.) containing radioactive fluids are designed to be controlled remotely to the maximum extent practical. This will minimize personnel exposures from the normal operation of these systems.
In-Service Inspection of Reactor Vessel Nozzle Welds The design of welds joining the reactor vessel nozzle to reactor coolant pipe permits in-service inspection to be accomplished from the inside diameter of the reactor vessel. Automated equipment, operated remotely normally used for reactor vessel pressure boundary inspections, can be utilized in this area.

In the event that in-service inspection of this area is performed from the outside, insulation for the reactor vessel and reactor coolant piping utilizes removable sections for access. These i removable sections are lightweight and are held in place mainly by quick actuation type buckle fasteners. After the necessary panels are removed, remote equipment can be utilized to perform em the required inspections. b

Blanket Type Thermal Insulation System 80 + will use blanket type thermal insulation, wherever practical, held in place by velcro fasteners for components on systems containing radioactive fluids. A metal jacket around the insulation will be provided. The jacket is held in place by quick actuation type buckle fasteners.

This insulation will be easily removable and will facilitate the performance of in-service weld inspections. This will minimize personnel exposures associated with in-service inspections. , 1

Electrical Service and Lighting l The System 80+ design provides good lighting and convenient electrical service. This will facilitate maintenance and inspection activities and reduce the anticipated personnel exposure.

Reliable extended service lamping in high radiation areas will be used, whenever possible, to minimize the frequency of maintenance required. The lighting fixtures are located to minimize i personnel exposure during maintenance. These features are in accordance with Regulatory Guide 8.8, Position C.2.i guidance.

Spent Fuel Pool Decontamination System 80+ provides the capability to use high pressure demineralized water for the decontamination the spent fuel pool. Alternative methods of decontamination, such as use of a strippable coating, may be evaluated by the operator, as practical.

(o) v 8 COL information item; see DCD Introduction Section 3.2. Approved Design Material- Radiation Protection Page 12.3-7

System 80+ Design Crntrol Document

Snubbers Mechanical snubbers rather than hydraulic snubbers will be used in radiation areas to minimize l I

the frequency of maintenance and inspections required. 12.3.1.3 Source Term Control i Source term control is an important aspect of the System 80+ design. The following design features reduce the overall dose due to operation, maintenance, and inspection activities. e Fuel Performance The System 80+ design features assure low primary system sources with impreved fuel clad leakage performance of less than 0.1% fuel clad failures, as well as an extended fuel cycle. e Corrosion Product Control l System 80+ design includes design features that reduce corrosion product production in the primary system. I

1. Primary System Materials The System 80+ design specifies primary system materials with low corrosion rates and very low cobalt impurities (0.05% by weight for equipment in direct contact with the primary coolant) except where no proven alternative exists.

The presence of antimony in RCP bearings has presented a problem with hot particles in the current generation of nuclear plants. In the System 80+ design, the reactor coolant pump bearings will be designed to minimize the presence of antimony. Steam generator tubes are fabricated to relieve stresses to reduce stress corrosion cracking. This will reduce the probability of tube plugging activities and further reduce i maintenance exposures. j Control rod drive materials are specified with low cobalt alloys, if no proven alternative exists to reduce RCS exposures.

2. Primary System Chemistry Increased pH in the range of 6.9 to 7.4 reduces equilibrium corrosion rates and buildup of activated corrosion products on primary system surfaces.

12.3.1.4 Airborne Contamination Control In the System 80+ design, plant ventilation systems are designed so that flow is from areas of lower to areas of higher potential activity. This design minimizes the potential for the spread of contamination. In addition, the following conf'mement devices are utf.ized to minimize the spread of contamination: O Approved Design Material- Radiation Protection Page 12.3-8

i.

              . - . . . . - .            .--- ._ ..- -.--.-.-~ - .                                     _ - - . - . .          . - . . - .

4 l Systrm 80+ Desian control Document I . i

Drip Containment

( ' Drip containment devices are used to collect equipment leakage and prevent suspension of p radioactive particulate into the air or volatile radioisotopes, such as noble gases and radioiodines. i j

Glove Bags ;

Glove bags are used to perform maintenance activities, such as valve refurbishments, in an ' j enclosed area. {! '* Tents Tents provide a large enclosed area to perform work such as grinding or maintenance on large

equipment. These tents are provided with ventilation capabilities and essentially provide for a .!

j local hot machine shop. l l

Hot Machine and htstrument Shops 1 1

! These swas provide a dedicated area where maintenance can be performed on radioactive and l_ contaminated equipment. a

Loop Seals

} ;( Water filled loop seals are provided in the Floor Drain System, discussed in Section 9.3.3, to i preclude the flow of contaminated air from one area / floor to another. i 12.3.1.5 Equipment Improvements j e The System 80+ RCPs incorporate a cartridge type of RCP seal which is a proven, reliable and , ! easily replaceable seal design. The replacement is also facilitated by the addition of platforms j

around the RCPs. This design allows the seal to be removed and repaired outside the crane wall ;

[ or other low dose area. Therefore, the time required to perform maintenance on the RCP seals ] l and maintenance exposure is reduced. i

I

?

Steam Generator Maintenance j System 80+ design includes several features which enhance accessibility during maintenance and .

inspection. 'Ihese features, described in Section 5.4.2, reduce the overall exposure to personnel ! 4 during these activities. These features include:

1. Use of automatic /robotic equipment for inspection and maintenance activities

2. Adequate pull and laydown areas

3. Platforms

4. Handholes

5. Increased size of manways to 21" www areeuw.neneswo prosecuan !*oe 1239

Sy:t m 80 + Design ControlDocument

6. Use of removable insulation to facilitate weld inspection 1

7. Use of Inconel 690 for tubes to reduce corrosion product production.

Also, included in the System 80+ design are features which are important to achieving ALARA goals. These include:

1. Considerations for equipment reliability, maintainability, and accessibility j i

2. Component design, i.e., tank design, piping design and instrument design to minimize l particulate deposition

3. System flushing and decontamination capability i 1

4. Radwaste handling operations (also discussed in Sections 11.2 - 11.4)

5. Isolation of contaminated components and proper shielding !

i

6. Controlled access to high radiation area via locked doors l

7. Piping containing radioactive liquid, resins, of gases are routed through shielded pipe chases. !

In order to maintain exposure ALARA and to aid in the layout and shielding design, the station is divided into radiation zones. These zones indicate maximum dose rates based on design activities only. The zone limits are summarized in Table 12.3-1. 12.3.1.6 Radiation Zone Designation The radiation zones for the Nuclear Island for normal operating conditions are designated in Table 12.3-2, , as well as the associated Radiation Zone Maps illustrated in Figures 12.3-1 through 12.3-8. The i Radiation Zone maps for the Radwaste Building are illustrated in Figures 12.3-18 through 12.3-21. The turbine building is generally a low radiation area. The only potentially radioactive areas are the blowdown filter and demineralizer, condensate polisher, and the flash tank rooms during normal and anticipated operational occurrences. The radiation zone designations for normal and anticipated operational occurrences are illustrated in Table 12.3-2. Following significant primary-to-secondary steam generator leakage events, the secondary side of the plant may become contaminated. The condenser air removal system radiation monitor will provide indication of a steam generator tube leak. It is assumed that the operator will isolate the damaged steam generator within 30 minutes following indication of tube leakage in excess of technical specifications or following a reactor and turbine trip indicative of a steam generator tube rupture. Therefore, the contamir ; tion of the secondary side systems is minimized. Since control systems and operating procedures limit the duration that the condensate and blowdown systems process secondary water following steam generator leakage events, the activity of the spent resins and filter cartridges would not be significantly impacted by abnormal primary-to-secondary leakage or steam generator tube rupture events. The radiation zone designations for post-accident conditions are illustrated in Table 12.3-4. Approved Design Material-Itadiation Protection Page 12.310

            .   - .                 .     -     . - . - . - - . - . - . . - . - . - . . . . -       - . . . -                   - - ~ . ~ - , -.-

a i l Svat:m 80+ oerian centrolDocument The station service building is not expected to contain any radiation sources. Figure 1.2 20 provides j.- layout de4ign details. ; i The .radwaste building houses the Liquid Waste Management System (LWMS) and Solid Waste i Management System (SWMS) components. These systems are not used to process post-accident waste i streams; therefore, the radiation zone designations are not expected to be adversely impacted during post-i accident conditions. The anticipated post-accident radiation zone designations are illustrated in Table l 12.3-4. 1 12.3.1.7 General Design Considerations to Keep Post-Accident Exposures ALARA l [ Direct and airborne sources are considered when determining that access can be provided to those vital { areas necessary for the control of the plant. The plant design ensures that personnel exposures will meet - j GDC 19 and NUREG-0737 guidelines. f Sampling capabilities with exposures kept ALARA will incorporate a post-accident sampling system that meets the requirements of NUREG-0737 and Regulatory Guide 1.97. ; 1 l 12.3.1.8 Post-Accident Radiation Zones i Radiation Zone maps were developed in accordance with NUREG-0737 (amended by the guidance provided in Draft NUREG 1465 relative to post-LOCA source terms) to review access throughout the plant following a DBA. The layout assists in keeping occupational doses ALARA even following a DBA. O Source terms are discussed in Section 12.2.2. Continuous access will be provided during post-accident conditions with dose rates less than or equal to 15 mrem /hr to the following Advanced Control Complex vital area:

Main Control Room
Technical Support Center
Remote Shutdown Panel
Computer system area e Rooms housing Instrument and Control systems and equipment i Required access to the following vital areas and . systems will not exceed 5 rem:
Hydrogen Reco.mbiner Rooms

                    *.           Process Control Sarapling Panels

Primary Chemistry Labs Generic plant emergency procedures were reviewed and the above areas were identified as vital areas.

The following systems were considered for post-accident access; however, they do not constitute vital

                  . areas.-

Approved Design ninterief - Redesion Protection !*90 12 3-11

i Sy't m 80 + Design C*ntrol Document i l

Annulus Ventilation System o Safety Injection System I
Containment Spray System o Shutdown Cooling System i
Chemical and Volume Control System i
Post-Accident Sampling I
Subsphere Ventilation System

((A complete list of vital areas will be developed by the Owner Operator based on site specific Emergency Procedures.))l l The zone limits are summarized for the DBA LOCA in Table 12.3-3. The Radiation Zone designations, I as well as the Radiation Zone Maps are shown in Table 12.3-4 and Figures 12.3-9 through 12.3-16, ! respectively. j 12.3.1.9 Vital Area Access i The following discussion details the post-accident access routes to the above vital areas. Vital area access ! routes are illustrated in Figures 12.3-9 through 12.3-16 by arrows. It should be noted that during normal I operating conditions, the Radiation Control Area (RCA) has a single point of access on elevation 91 +9. I However, during post-accident conditions, emergency access and egress are possible on all elevations to and from the RCA as needed.

Control Complex l

Areas in the contro! complex, such as the main control room and the Technical Support Center (TSC), I are generally accessible on all elevations from stairwells located at B-C,10-11 and B-C,23-24, as well as from elevators 100,200,300 and 400. The respective corridors' radiation zone designations are not impacted by post-accident conditions.

Hydrogen Recombiner Rooms Hydrogen recombiner room #1 is accessible by entering the RCA on elevation 70+0. The operator would travel along corridor B-Q,22-23 to the accessway into valve maintenance shop. The operator would then proceed to travel through the valve maintenance shop to the hydrogen recombiner room #1 accessway. Similarly, the operator would access the hydrogen recombiner room #2 via corridor B-R, 11-12 to the CVCS area rtorage. The operator would then travel through the CVCS area storage to the hydrogen recombiner room #2 accessway.

O 3 COL information item; see DCD Introduction Section 3.2. Approved Design Material- Radiation Protection Pops 12.3-12

Sv3t:m 80+ Design ControlDocument i 4

Process Sampling Panel and Primary Chemistry Lab Area The process sampling panels are accessible by entering the RCA on elevation 70+0. The operator would

, travel down corridor B-S,11-12 to the stairwelllocated at S-T,11-12. The operator would then proceed down the stairwell to elevation 50+0. Upon exiting the stairwell the operator would turn right and

proceed along the corridor T-U,11-16 to the process sample panel room. Once samples are taken at the

. process sample panels, the operator would proceed back down the corridor T-U,16-11 to the primary 2 chemistry lab area where the samples would be processed and analyzed. j

12.3.2 Shielding.

i .. l 12.3.2.1 Shielding Analysis i Calculations to determine the adequacy of the station shielding are based on Section 12.2.1 source j strengths and the methods outlined below. Dose points are selected inside and outside cubicles containing radioactive equipment. Cubicle ceilings and floors are generally the same thickness as the cubicle walls. Skyshine from the station is negligible because cubicles containing radioactive material are shielded overhead. 1 The only major source in the station is the reactor core at full power. The codes ANISN, DOT, l MORSE, and SABINE are used to verify the effectiveness of the primary shield. Sources of gamma radiation are distributed throughout the reactor 2ilding subsphere and nuclear annex. The codes SHIELD and KAP-IV are used to verify gamma source shielding. The following sequence typifies a gamma source shielding analysis:

1. Determine the concentration of each principal nuclide in the source medium.

2. Adjust the concentration to account for accumulation, dilution, decay, removal, etc.

3. Convert the resulting concentrations into gamma source strength.

4. Select an idealized model or combination of models to represent the physical shape of the source container and all shields present.

5. Assemble the necessary data on attenuation properties of the source and the shield materials,

6. Perform the calculation for the desired dose point and tabulate the results for comparison with design objective dose rates.

Steps 1 through 3 are done with the code N237 BURP and data from Sections 11.2 through 11.4. Step 4 is self-explanatory; tanks, demineralizers, filters, and pipes are modeled as a right circular cylinder, etc. Except for inputting material densities, Step 5 is code internal. Step 6 simply determines the adequacy of the shielding. All of the computer codes necessary to perform the above analysis meet all NRC and industry standard. Typical computer codes are described below: O ANISN performs shielding calculations by discrete ordinates solution of the Boltzman equation in one direction. Through use of transport theory with anisotropic scattering. ANISN is well suited to deep Approwd onion unterw - nemeuan prosecuan rage s2.3 12

Syst;m 80+ Design ControlDocument penetration problems. A 40 group coupled cross-section set is utilized to account for both neutron attenuation and secondary gamma radiation. Calculations are made in cylindrical geometry. SABINE solves neutron and gamma ray shielding problems with removal-diffusion methods. The neutron / gamma production is a specified fission distrib'.ition in the source region. The code calculates neutron attenuation through shields using nineteen removal energy groups that in turn feed twenty-six groups for the diffusion calculation. Secondary gamma production in each shield region is output as a polynomial curve fit. Gamma fluxes are also calculated. KAP IV employs the point kernal technique to dete"mine dose rates from complex sources whose geometries can be described by second order surface eq.tations. An exponential attenuation function with buildup is employed for gammas. Neutron attenuation functions are also available. SHIELD is a Duke Power Company code that calculates fluxes at receiver points with integrals over simple geometries. The gamma spectrum is divided into six energy groups. Input includes group specific source strength and average energy, source and shield geometries, and material densities. Whenever the spectrum of average energies changes, energy depend:nt parameters are recalculated. The code contains energy dependent data on tissue flux-to-dose conversic<n factors, mass attenuation coefficients for common source / shield materials and Taylor-form buildup factor coefficients. For combined shields the buildup factor is automatically based on the material with the greatest optical thickness in the lowest energy group. When calculations exceed code-internal data, appropriate warning statements are output. MORSE is a multi-purpose neutron and gamma ray Monte Carlo transport code. Through the use of multigroup cross-sections, either forward or adjoint solutions of neutron, ganuna ray, or coupled neutron-gamma ray problems may be obtained. Time dependence for both shielding and criticality problems is provided. Three dimensional as well as specialized geometry descriptions may be used. An albedo option is available at each material surface. Also available is isotropic or anisotropic scattering up to a Pi6 expansion of the angular distribution. DOT solves the Boltzman transport equation in two dimensional geometries by use of the particles moving along discrete directions in each cell of a two dimensional mesh. Anisotropic scattering is treated using a Legrendre expansion of arbitrary order. Both homogeneous and external source problems can be solved. Albedo boundary conditions are available. i i N237 BURP is a Duke Power Company code that calculates the accumulated activity on demineralizer i resins or filters and the resultant activity of the process stream. This is accomplished by solving a pair i of coupled, first order differential equatiorts. Required input is isotropic removal efficiencies and j operation time. Gamma source strengths are obtained from the calculated specific activities by ' considering gamma yield and losses due to conversion electrons. The nearly 300 individual gamma emissions of these isotopes are divided into six discrete energy groups. Group boundaries remain fixed, but the average group energy is calculated for each spectrum ofisotopes. This allows reasonably precise selections of energy dependent shield material properties for attenuation properties. 12.3.2.2 Shielding Design The plant shielding shall be designed to achieve the radiation zones designated in Tables 12.3-2 and 12.3-3 for normal operation and post-accident conditions respectively. Accessible areas that are potentially greater than 100 R/hr are listed in Table 12.3-6. Approved Design Material- Radiation Protection Page 12.314

                                                                     ,         _.             .. .                   ~    --

l Srtem 80+ Design controlDocument ( Transient sources of greater than 100 R/hr are considered in the System 80+ shielding design to ensure C adequate shielding is provided. One such source is a spent fuel assembly. During transfer of a spent fue assembly through the fuel transfer tube, adjacent corridors may experience elevated radiation levels. Streaming from this source up through thejoint between the Reactor Building and the Nuclear Annex has been a concern for the current generation of nuclear plants. The System 80+ design utilizes a connected building design to reduce the potential for streaming. In addition, a lead collar is provided around the fuel transfer tube, as well as several feet of additional concrete shielding to maintain adjacent corridors radiation levels ALARA. This permits personnel to perform maintenance and inspection activities in a l lower radiation area and reduces the potential for adverse radiation zones from impacting refueling l outage schedules. An inspection area is provided beneath the fuel transfer tube. Access control is ensured by l ' the provision of an alarm on the hatch to sound when the door is open to the spent fuel transfer tube i inspection area, as well as a lockable access door to the inspection area. Figure 12.3-17 provides a sectional view of the fuel transfer tube, the shielding provided to the adjacent areas, and the expected radiation zones in these areas during fuel transfer. ((The COL Applicant will perform a shielding analysis to verify the adequacy of the shielding i provided.))l Sufficient shielding will be provided to ensure that the areas adjacent to the spent fuel transfer tube inspection area are accessible and that the expected radiation zones are consistent with those illustrated in Figure 12.3-17 during transfer of a spent fuel assembly. Typically, infrequent access to pipe chases is required. The System 80+ design specifies that location of active components, such as valves, will be avoided whenever possible in pipe chases, to minimize the , frequency of access required by plant personnel and personnel exposure. When access is required, j radiation protection personnel would conduct a survey of the area to determine the high radiation areas (e)

   "      within the pipe chase. Temporary shielding will be utilized, as necessary, to minimize the personnel 1

exposure. For instance if access to the pipe chase (17-20, R-T) on elevation 70+0 through the adjacent pipe chase (13a-17, R-W) is required, temporary shielding will be used to reduce personnel exposure. The primary source of radiation in the adjacent pipe chase (17-20, R-T) is the spent resin or slurry transfer piping. Precautions will be taken by operating personnel to ensure that no spent resin transfer will be performed while personnel are in the pipe chase. The resin transfer lines are also provided with flushing capability to minimize the potential for hot spots in the piping. ((The COL Applicant will provide maintenance procedures that will include precautions regarding resin transfer during maintenance activities.))l The incore chase is potentially an extremely high radiation area (greater than 100 R/hr) during incore instrumentation withdrawal. Positive access control is provided to this area during movement of the incore instrumentation. A lockable access door is provided with a warning light. During withdrawal ofincore instrumentation, the warning light illuminates providing indication that the incore instrumentation is being withdrawn. An area radiation monitor is located in the incore chase to provide indication of radiation levels and to alarm when incore instrumentation is being withdrawn due to high radiation in the area.

       .An electrical interlock is provided between the area radiation monitor and the access door to prevent a'ccess into the incore chase during withdrawal of the incore instrumentation. Emergency egress from the area is also provided from the incore chase. Also, radiation protection personnel will post a high radiation sign just outside the entrance to the incore chase to warn personnel during incore instrumentation withdrawal.

V

        '           COL information item; see DCD Introduction Section 3.2.

A,;.,; A Design Material Radiation Protection (1/9h !*90 fa.3'15

l Sy:trm 80+ Design controlDocument 1 The plant shielding is designed not only to maintain personnel occupational exposure ALARA, but also to maintain the exposure to the general public ALARA. The plant shielding is designed so that the ' contribution from adjacent areas or cubicles to a radiation zone is a small fraction (i.e.,0.25) of the total j dose rate anticipated in that radiation zone. i In addition, as a result of normal operational occurrences, the contribution from direct and scattered l radiation to the maximally exposed member of the general public in the unrestricted area is less than or equal to a small fraction (i.e., 0.25) of the total ant!cipated exposure during normal operations. (( Detailed shielding calculations will be provided by the COL Applicant in accordance with the l methodology discussed in Sections 12.2 and 12.3.))1 i 12.3.3 Ventilation The spread of airborne contamination within the plant is minimized by the design of the plant ventilation l systems to provide airflow from areas of lower potential airborne contamination to areas of higher l potential airborne contamination. For building compartments with the potential for contamination, the l exhaust is designed for greater volumetric flow than is supplied to the area to minimize the amount of uncontrolled exfiltration from these areas. These design features ensure that the average concentration of radioactive material in the air for areas normally occupied is well within or a small fraction (i.e., 0.25) of 10 CFR 20, Appendix B. Table 1, Column 3 limits. Therefore, personnel exposure due to inhalation of and immersion in airborne contamination is maintained ALARA. Airborne radiation monitoring is provided to areas normally occupied with a significant potential to have airbome contamination. These monitors are designed to have the capability to detect the time integrated change of the most limiting pa:ticulate and iodine species equivalent to 10 CFR 20, Appendix B of Sections 20.1001 - 20.2402, inplant concentrations in each area within 10 DAC-hours. Airborne monitors are discussed in further detail in Section 11.5. The approximate location of the process effluent radiation monitors are shown in Figures 6.2.3-1, 9.4-2, 9.4-3, 9.4-5, 9.4-6, 9.4-8 and 9.4-9. The airborne radiation monitors are located upstream of the process filters. The ventilation systems are discussed in detail in Section 9.4. 12.3.4 Area Radiation and Airborne Radiation Monitoring Instrumentation Area and airborne radiation monitoring instrumentation are discussed in Section 11.5. Area radiation monitors are illustrated in Figures 12.3-9 through 12.3-16 and 12.3-18 through 12.3-21. (( Radiation monitors will be provided and located in accordance with Regulatory Guide 1.97 and NUREG-0737 recommendations.))l Table 12.3-5 lists the area radiation monitors and their approximate locations. Section 11.5 specifies each monitor's range of sensitivity, power supply, arv seismic category. Area monitors are located based on the expected frequency of access, occupancy time, and expected and potential radiation levels in plant work areas. O 3 COL information item: see DCD Introduction Section 3.2. Approved Design Material Radiation Protecthm Page 12.3-16

Sv? tem 80+ Deslan controlDocument

Areas which are typically high radiation areas, such as pipe chases, and have little or no accessibility will not be provided with an area radiation monitor.
Areas which typically have a high frequency of access and are normally low radiation areas but have a potential to be a high radiation area will be provided with an area radiation monitor.

In post-accident conditions these areas may become very high radiation areas, such as corridors outside the personnel and equipment hatches. Area monitors will not be located in areas which typically do not require access and are high radiation areas (e.g., pipe chases) or areas which are typically not radioactive and are accessed frequently (e.g., Turbine Building). Areas that require continuous occupancy and can be a radiation area will be provided with an area radiation monitor. In addition, areas that are frequently accessed that have a potential to become high radiation areas during normal or abnormal operational occurrences will be provided with area radiation monitors. Area monitors are also located in the new fuel vault storage area to provide , some degree of criticality monitoring capability. Area monitors are located to maximize area of coverage and minimize inadvertent shielding of the detector by stmetural materials. The containment high range monitors are located in accordance with guidance provided in NUREG-0737, Section II.F.3, which is specified as follows:

The two containment high range monitors are physically separated (i.e., widely separated) to provide independent measurements and ensure that a large fraction of the containment volume is viewed by the monitors.
The monitors will not be located in areas that are shielded by massive structural material.
The monitors will be located so that they are accessible for calibration and maintenance activities.

Process monitors for ventilation systems are located upstream of process filters. In general, process, effluent, and airborne monitors will be located in an area which is:

easily accessible; and
provided with sufficient shielding to ensure that the required sensitivity is achieved at the design background radiation level for the area.

((The COL Applicant will develop procedures for Radiation Protection to conduct smveys using ponable

 - radiation monitors as pan of the Radiation Protection program.))3 During post-accident conditions, plant personnel will be evacuated from the RCA. Radiation protection personnel would then conduct surveys using ponable radiation monitors to determine the optimal routes to vital areas through the RCA             at personnel exposure is minimized. Radiation protection personnel would typically escon mainte, nance personnel and operators into the RCA and would continue to monitor the radiation levels using ponable radiation monitors. The area monitors located in the plant would provide an audible and visual alarm if high radiation levels were detected.

a 3 COL information item; see DCD Introduction Section 3.2. Approved Design Atatorial- Rodneiert Protectiars Page 12.317

Syst~m 80 + Design ControlDocument Table 12.3-1 Normal Operation Accessibility Zone Designations Zone Designations Dose Rate (mrem /lir) 1 Less than 0.5 2 0.5 to 2.5 3 2.5 to 15 4 15 to 100 5 over 100

                  .                                                                           i Table 12.3-2 Normal Operation Radiation Zones 1

Peom Radiation OPS Zone S/D Reactor Building /Subsphere El 50+ 0, Figure 12.3-1 Emer. Feedwater Motor Driven Pump Room 2 2 Emer. Feedwater Turbine Driven Pump Room 2 2 l SCS Heat Exchanger Rooms 3 5 l CS Heat Exchanger Rooms 3 3 Maintenance Aisle 2 2 l CS Pump Rooms 3 5 ! SCS Pump Rooms 3 5 I CS h'iniflowlicat Exchanger Areas 3 5 l SCS Miniflow HX Areas 3 5 l SI Pump Rooms 3 3 Nuclear Annex El 50+0, Figure 12.3-1 Corridor to FD (B-H,11-12) 1 1 Corridor to FD (B-H,22-23) 1 1 i Corridor Past FD 2 2 ! Vital I&C Channel A, B, C, D 1 1 l Channel A, B, C, D Battery Room 1 1 Cable Chase A, B 1 1 Stainvell (D-E,18-19) 1 1 Elevators E100, E200 1 1 Elevators E300, E400 1 1 Stairwell (B-C,11-12) 1 1 Stairwell (B-C, 22-23) 1 1 Maintenance Work Area (E-H,9-11) 1 1 Mechanical Work Area (E-H,23-25) 1 1 Instrument Air Room (B-C,9-10) 1 I Instrumeu Air Room (B-C,23-25) 1 1 Sumps (H-J, 9-11) 2 2 Approved Design Afsterial- Radiation Protection Page 12.3-18

System 80+ Design controlDocument [3 , () Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Roota Radiation OPS Zone S/D Nuclear Annex El 50+0, Figure 12.3-1 (Cont'd.) Sumps (H.J. 22-23) 2 2 HVAC Chase (F-G,13-14) 1 1 Storage Area (B-C, 9-10) 1 1 Storage Area (B-C,24-25) 1 1 HVAC Chase (F-G,20-21) 1 1 CCW Pump Rooms (J-K,911) 2 2 CCW Pump Rooms (J-K,23-25) 2 2 Stairwell (0-P,11-12) 2 2 Stairwell (0-P 22-23) 2 2 Elevators E500, E600 2 2 HVAC Chase (P-Q,15-16) 2 2 HVAC Chase (P-Q,18-19) 2 2 HVAC Chase (M-N, Il 12) 2 2 HVAC Chase (N-0,22-23) 2 2 Pipe Chase (N-0,11-12) 2 2 CT Pipe Chase (M-N, 22-23) 2 2 NM Stairwell (S-T,11-12) 2 2 RD Pumps 5 5 EQ Drain Tank 5 5 Reactor Makeup Water Pumps 4 3 Charging Pump Rooms 5 5 CVCS Chem Add Pkg 5 5 Gas Stripper Control Panel 3 3 Charging Pump Mini Flow HX 5 5 Gas Stripper 5 5 Boric Acid Conc. 5 5 Back Control Panel 3 3 Equip Drain Sump 5 5 Resin Sluice Tanks 5 5 RST Valve Access 5 5 Resin Sluice Pumps 5 5 Sampling Panels 3 3 Sampling Panels Pipe Chase 5 5 ! Floor Drain Sump (T-U,16-17) 5 5

 ,,       Floor Drain Sump (S-T,18-19)                             5                  5

( ) CVCS Equip 'nt Drain Sump (T-U,11-12) 5 5 5 CVCS Equip::ent Drain Sump (T-U,21-22) 5

      ' L ..: Deekprs Atason\et

Meelween hoseceien Page 12.3-19

Sys:em 80+ Design ControlDocument Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) ! 1 Room Radiation OPS Zone S/D Nuclear Annex El 50+0, Figure 12.3-1 (Cont'd.) CVCS Area (T V,11-17) 3 3 Primary Chemistry Labs 2 2 Diesel Generators 2 2 Equip Access Areas 2 2 Unidentified Area (R-U,22-23) 3 3 Janitorial /HP Storage and Work Area 2 2 Cont Cooler Cor:1 Tank and Pump Rooms 4 4 Remetor Bldg /Subsphere El 70+0, Figure 12.3-2 Fuel Pool Cooling Pumps 3 3 Fuel Pool Heat Exchangers 3 3 Pipe Chase 4 4 Open to Elevation 50 Below 3 3 Inside Containment 5 5 Maint. Aisle / Valve Gallery 4 4 Nuclear Annex El 70+0, Figure 12.3-2 1 Corridors to RAD Control Point (C-I 22-23) 1 1 Corridors to RAD Control Point (C-111-12) 1 Storage Area (B-C,9-10) 1 1 Storage Area (B-C,24-25) 1 1 Elevators E100, E200 1 1 Stairwell (B-C,10-11) 1 1 Non-essential Chillers 1 1 Essential Chillers 1 I Remote Shutdown Panel 1 1 Division 1 Battery Room i 1 Division 2 Battery Room 1 1 Division 1 Channel Equipment (C-1,12-15) 1 1 Division 2 Channel Equipment (C-1, 20-22) 1 1 Cable Chases A, B, C, D 1 1 Stairwell (D-E,18-19) 1 1 HVAC Chase (F-G,20-21) 1 1 HVAC Chase (F-G,13-14) 1 I Elevators E300, E400 1 1 Stairwell (B-C, 21-23) 1 1 Reserved Areas for Cable (E-R 14-15) 1 1 Reserved Areas for Cable (E-R,19-20) 1 1 Reserved Areas for Cable (0-P,13-13a) 2 2 L . .::Desipre ateteriel Redoenlon Protection Page r2.3-20

Srten 80+ Design ControlDocument r {y J Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Room Radiation OPS Zone S/D Nuclear Annex El 70+0, Figure 12.3-2 (Cont'd.) Reserved Areas for Cable (0-P,20-21) 2 2 l EFW Tank Rooms 1 1 Equipment Access (S-T,12-13) 2 2 Cable Area Channel A & B to Subsphere 1 1 Equipment Access Shaft (Q-R,21-22) 2 2 FJevators E500. E600 2 2 Stairwell (0-P,11-12) 2 2 Stairwell (0-P 22-23) 2 2

      !!VAC Chase (M-N,11-12)                                  2                 2 liVAC Chase (N-0,22-23)                                  2                 2 Division 1 Channel Equipment (0-R,19-21a)                2                 2 Division 2 Channel Equipment (0-R,13-15)                 2                 2 Diesel Generators                                        2                 2 Fuel Pool Purif. IX's                                    5                 5 Purification IX's                                        5                 5 fN C

Deborating IX Preholdup IX 5 5 5 5 Boric Acid Conc IX 5 5 Boric Acid Makeup Pumps 5 5 Pool Purification Pumps 5 5 lioldup Pumps 5 5 Pipe Chase (R-W,13a-17) 5 5 Pipe Chase (R-T,17-20) 4 4 Pipe Chase (N-0, Il-12) 2 2 Pipe Chase (M-N,22 23) 2 2 Seal INJ llX 5 5 CVCS Area Storage 2 2 Post-Accident flydrogen Recombiner Areas 2 2 Valve Maintenance Shop 3 3 Unidentified Area (Q-U,20-23) 3 3 Nuclear Annex El 81 +0, Figure 12.3-3 Fuel Pool Purification IX's 5 5 Boric Acid Concentrator IX 5 5 Valve Room (V-W,13a-17) 4 4 , Purification IX's 5 5 (, j Deborating IX 5 5 Pre-holdup IX 5 5 Approved Design Acaterial . Radevon Protection Page 12.3-21

Design ControlDocument Q* tem 80 + l Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Room Radiation OPS Zone S/D Nuclear Annex El 81+0, Figure 12.3-3 (Cont'd.) Valve Room (U-V,13a-17) 4 4 Boric Acid Filter 5 5 RX Drain Filter 5 5 Pool Filters 5 5 Purif Filters 5 5 Valve Room (S-T,13a-17) 4 4 Valve Galleries 3 3 Seal Inj Filters 5 5 RMUW Filter 5 5 Valve Room (R-S,13a-17) 4 4 Pipe Chase (P-R 16-18) 4 4 Pipe Chase (R-S,16-17) 5 5 Pipe Chase (S-T,17-18) 4 4 Equip Access Shaft 2 2 Stairs (S-T,11-12) 2 2 Stairs (T-V,17-18) 3 3 IIVAC Chase (P-R,15-16) 2 2 , HVAC Chase (P-R,18-19) 2 2 Electrical Equipment (0-R.13-15) 2 2 Reactor Building El 91+9, Figure 12.3-4 Pipe Chase 4 4 Penetration Room A. B litl till Penetration Room C. D 2l!! 2131 Annulus Pipe Penetration Area 4 3 HVAC DistributionIIcader Areas 5 4 Reactor Drain Tank Room 5 5 RCP Motor Oil Drain Rooms 5 5 Elevator 5 5 incore Chase 5 5 IIoldup Volume 5 5 Reactor Vessel Cavity 5 5 Regenerative HX 5 5 , 1 Letdown llX 5 5 Nuclear Annex El 91+9, Figure 12.3-4 Corridor to Admin Bldg. I 1 0 lil Caution should be used in the placement of penetrations in these rooms.

4twowd Design Metenal . Radiation Protecten Page r2.3 22

    . . - _                 _        -. .          _ _       __...._.___._m    _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _

Sy' tem 80 + Desian controlDocument p. Table 12,3-2 Normal Operati<m Radiation Zones (Cont'd.) Room Radiation GPS Zone S/D Nuclear Annex El 91+9, Figure 12 3-4 (Cont'd.) ! Radiation Access Control Point 1 1 Stairwell (B-C,10-11) 1 1 Elevators E100 E200 1 1 CAS Room i 1 Stairwell (B-C, 23-34) 1 1-Elevators E300, E400 1 1 Corridor to Rad Control Points 1 1 Stairwell (D E,18-19) 1 1 HVAC Chases (F-G,12-14) 1 1 HVAC Chases (F-G,20-21) 1 1 Corridor Past Rad Control Point 2 2 EFW Rooms 1 1 Elevators E500 E600 2 2 Stairwell (0-P,11 12) 2 2 Stairwell (0-P, 22-23) 2 2 Hot Machine Shop 3 3 HVAC Chases (P-R,15-16) 2 2 HVAC Chases (P-R,18-19) 2 2 Truck Bays 2 2 ; Rail Car /Truch Washdown Areas 1 1 Equip Decon 3 3 EQ Access Shafts 2 2 Non-essential Elect. Equip N2 (B-I,12-17) 1 1 Non-essential Elect. Equip Ni (B-1,17-22) 1 1 Non-essential Elect. Equip N2 (N P,13-15) 2 2 . Non-essential Elect. Equip N1 (N-P,19-21) 2 2 IX and Filter Hatch Areas 3 3 Pipe Chase (R-T,19-20) 4 4 Pipe Chase (R S,16-17) 5 5 Storage (R-T,18-19) 2 2 13.8K VAC RCP Switchgear (F-G,12-13) 1 1 13.8K VAC RCP Switchgear (F-G,2122) 1 1 Security Equipment i 1 IIVAC Chase (M-N,11-12) 2 2 HVAC Chue (N-O,22-23) 2 2 Pipe Pre (N-O, Il-12) 2 2 { Pipe Chase (M.N. 22-23) 2 2 i

            , --- $~

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Srtem 80+ D sign ControlDocument Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) h ! Room Radiation OPS f Lone S/D Reactor Building El 115+6, Figure 12.3-5 Annulus 4 3 Personnel Air Lock 3 2 Area Outside Cranewall/Inside Coninment 4 2 Between Cranewall, RX Vessel & S/G Shield 5 5 Core Support Barrel Laydown Area 5 5 Upper Guide Structure Laydown Area 5 5 Reactor Vessel Cavity 5 5 Refueling Canal 5 5 Fuel Transfer Tube 5 5 Nuclear Annex El 115+6, Figure 12.3-5 Stairwell (B-C, Il-12) 1 1 Elevators E100, E200 1 1 Personnel Decon 2 2 Bmd Rwm i 1 Men's Restroom 1 1 Women's Restroom 1 1 Corridor to RAD Control Point (B-E,12-13) 1 1 Corridor to RAD Control Point (B-H,20-21) I 1 ! Control Room 1 1 l Tool Room i 1 Document /ISPD Room & Emerg. Supplies 1 1 Storage Room (E-G,18-20) 1 1 Stairwell (D-E,18-19) 1 1 Computer Room 1 I Reactor Operator's Office i 1 Control Room Supervisor's Office 1 1 Shift Supervisor's Office 1 1 Clerk's Office 1 1 Tag Out Area 1 1 Shift Assembly Area 1 1 Elevators E300, E400 1 1 Stairwell (B-C, 22-23) 1 1 Conference Rooms 1 1 11VAC Chase (F-G,13-14) 1 1 l IIVAC Chase (F-G. 20-21) 1 1 liot Machine Shop 3 3 Hot Tool Crib 3 3 Approved Design Atatorial Radiaten Protecten Page 12.3-24

Syhtem 80+ Design CcatrolDocument Table 12,3-2 -Normal Operation Radiation Zones (Cont'd.) l Room Radiation OPS Zone S/D l Nuclear Anux El 115+6, Mgure 12.3-5 (Cont'd.) l Hot Instrument Shop 3 3 l Main Steam Valve Houses 1 and 2 2 2 HVAC Chases (P-R,15-16) 2 2 HVAC Chases (P-R,18-19) 2 2 Refueling Canal 5 5 Spent Fuel Pool (Bottom) 5 5 Cask Laydown and Washdown Areas (Bottom) Slil Sl!1 Elevators E500, E600 2 2 Stairwell (0-P,11 12) 2 2 l Stairwell (0-P, 22-23) 2 2 i Equipment Access Shafts 2 2 Unidentified Area (R-W,1317) 3 3 Volume Control Tank 5 5 ; V',T Valve Access 3 3 Boric Acid Batch Tank 3 3 O O Pipe Chase (R-S,16-17) 4 2 4 2 Subsphere Exhaust l Pipe Chase (N-0, Il 12) 2 2 ; Pipe Chase (M N,22-23) 2 2 , Storage Area (Q-R,23-25) 2 2 , Nuclear Annex El 130+6, Figure 12.3-6 Stairwell (B-C,10-11) 1 1 Conference Room (B-D,19-21) 1 1 OSC's HVAC Room 1 1 TSC Area 1 1 OSC Room 1 1 TSC Ventilation Equipment Room 1 1 Control Room t 1 Controt Room HVAC Areas 1 1 Stairwell for Control Room (D-E,18-19) 1 1 Elevators E100, E200, E300, E400 1 1 Elevators E500. E600 - 2 2 . Stairwell (0-P, Il-12) 2 2 HVAC Chases (P-R,15-16) 2 2 HVAC Char.:s (P-R,18-19) 2 2

         - til        This area is zone 5 only when fuel is in the area.
           ? , . 2 Design Meterial- Re6neien hotecalan                                                         Pope 12.3-25

System 80 + De:Ign Control Document Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Room Radiation OPS Zone S/D Nuclear Annex El 130+6, Figure 12.3-6 (Cont'd.) EQ Access Areas 2 2 Stairwe!! (0-P, 22-23) 2 2 New Fuel UrJoading 2 2 New Fuel Storage 2 2 Spent Fuel Pool (Bottom) 5 5 Cask Laydown Area (Bottom) 5133 503 Cask Washdown Area 2 2 Nuclear Annex Supply 2 2 Nuclear Annex Edaust 2 2 Instrument Calibration Shop 2 2 Annulus Exhaust 2 2 Respirator Maintenance Shop 2 2 Main Steam Valve flouses 1 and 2 2 2 Volume Control Tank 5 5 IIVAC Chase (N-O,22-23) 2 2 IIVAC Chase (M-N,11-12) 2 2 Pipe Chase (N-O,11-12) 2 2 Pipe Chase (M-N,22-23) 2 2 Reactor Building El 146+0, Figure 12.3-7 Annulus 4 2 Area Between CV and Crane Wall 4 2 Area Between Cranewall & Primary Shield 5 2 Inside Primary Shield 5 5 Core Support Barrel Laydown Area 5 5 Upper Guide Structure Laydov n Area 5 5 Reactor Vessel Cavity 5 5 Personnel luk 3 2 Nuclear Annex El 146+0, Figure 12.3-7 Elevator Machine Room (B-C,11-12) 1 1 Elevator Machine Room (B-C,22-23) 1 1 Stairwell (B-C,10-11) 1 1 Stairwell (D-C, 23-24) 1 1 Laydown Work Space 2 2 Elevators E500, E600 2 2 Stairwell (0-P. Il 12) 2 2 H3 This area is zone 5 only when fuel is in the area. Approved Design hinterial Radatnon Protection Page 12.3-26

System (0 + Design ControlDocument Tab!e 12.3-2 Normal Operation Radiation Zones (Cont'd.) , Room Radiation OPS Zone S/D Nuclear Annex El 146+0, Figure 12.3-7 (Cont'd.) Stairwell (0-P, 22-23) 2 2 Hot Tool Cribs 3 3 Personnel Air lack 2 2 Hallway Outside Personnel lock 2 2 , Fuel Pool Laydov n & Equip Area 2 2 Fuel Pou! Arrace 2 2 HVAC Chases (N-0,22-23) 2 2 HVAC Chases (P-R,15-16) 2 2 IIVAC Chases (P-R,18-19) 2 2 Equipment Access Shafts 2 2 Pipe Chase (N-0,11-12) 2 2 Pipe Chase (M-N,22-23) 2 2 Personnel Decon Area 2 2

<      Reactor Building El 170+0, Figure 12.3-8 Annulus                                                  3                  2 Between Cont. Vessel & Cranewall                         3                  2 Inside Cranewall                                        3                   2 Steam Generator Rooms                                   5                   3 i       Nuclear Annex El 170+0, Figure 12.3-8 Elevator E500                                           2                   2 Stairwell (0-P, 22-23)                                  2                   2 Laydown/ Staging Space (S-W,13-15)                      2                   2

HVAC Equipment Area 2 2 HVAC Chases (P-R,15-16) 2 2 HVAC Chases (P-R,18-19) 2 2 HVAC Chases (V W,16-17) 2 2 HVAC Chases (N-0,22-23) 2 2 Fuel Pool HVAC Supply Area (V-W,15-16) 2 2 Fuel Pool Exhaust (Q-R,19-21 A) 2 2 Fuel Pool Exhaust (U W,16-17) 2 2 HI Purge Supply (T-V,15-16) 2 2 LO Purge Supply (R T,15-16) 2 2 Nuclear Annex Exhaust Area (N-P,19-22) 2 2 Pipe Chase (N-0,11-12) 2 2 Pipe Chase (M-N,22-23) 2 2

{. 2 ( Equipment Access Shaft 2 Hi Purge Exhaust Areas (R T,16-17) 2 2 Amroomt veew noww - nonieswr protecwwn rare r2.2-27

Sy~ tem 80+ Design ControlDocument Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Room Radiation OPS Zone S/D Nuclear Annex El 170+0, Figure 12.3-8 (Cont'd.) LO Purge Exhaust Areas (P R,16-17) 2 2 CCW Surge Tank (L-M,22-23) 2 2 CCW Surge Tank (N-0,11-12) 2 2 Turbine Building Blowdown Demineralizer Cubicles 3 3 Conderisate Polisher Cubicles 3 3 Flash Tank Room 3 3 301 3lti Pipe Chases Turbine Building General Area 1 1 Radwaste Building El 34+0 & 50+0, Figure 12.3-18 Chemical Waste Tank (CWT) Room 5 5 CWT Pump Rooms 5 5 Chemical Sample Tank (CST) Room 5 5 CST Pump Rooms 5 5 Detergent Sample Tank (DST) Room 4 4 DST Pump Room 5 5 Laundry & ilot Shower Tank (LliST) Room 4 4 LiiST Pump Room 5 5 Laundry Filter Area 4 4 Waste Monitor Tank (WMT) Rooms 5 5 WMT Pump Rooms 5 5 Demineralizer Rooms 5 5 Equipment Waste Tank (EWT) Rooms 5 5 EWT Pump Room 5 5 EWT Filter Area 5 5 Floor Drain Tank (FDT) Rooms 5 5 FDT Pump Room 5 5 Oil Drum Storage Area 3 3 Waste Filter Area 5 5 Empty filC Storage Area 5 5 HIC Dewatering /Washdown/ Inspection / Labeling and Transfer 5 5 Area Remote Operation Room 2 2 General Area 2 2 Low Activity Spent Resin Tank (LASRT) Room , 5 5 9 UI Radiation zone designation is based on resin slurry transfer piping. Approved Orsion Maternal- Radietmn Protection Page 12.3 28

System 80+ Design ControlDocument n Q Table 12.3-2 Normal Operation Radiation Zones (Cont'd.) Room Radiation OPS Zone S/D Radwaste Building El 34+0 & 50+0, Mgure 12.3-18 (Cont'd.) High Activity Spent Resin Tank (HASRT) Rooms 5 5 Pipe Chase (F-G, 7-8) 5 5 Radwaste Building El 70+0, Hgure 12.3-19 CWT Room 5 5 CST Room 5 5 General Area 2 2 WMT Rooms 5 5 EWT Rooms 5 5 FDT Rooms 5 5 Oil Separator Rooms 3 3 Chemical Lab 2 2 4 LASRT Room 5 5 liASRT Rooms 5 5 Pipe Chase (F-G,7-8) Sill Stil Radwaste Building El 91+9, Hgure 12.3-20 Solid Waste Storage Area 3 3 3 Empty Box Area 2 2 Full Cask Storage Area 5 5 HIC Storage Area 5 5 Corridor (E-F. 8-11) 2 2 Decontamination Area 3 3 , Low and Ifigh Activity Spent Resin Surge Tank Room 4 4 , Radwaste Building Access Control Area 2 2 Counting Room 2 2 Radwaste Building El 115+6, Mgure 12.3-21 HVAC Equipment Room 2 2 Control Room 1 1 Electrical Equipment and Battery l!ocm 1 ! Corridors (G-li, 4-10) (G-1, 3-4) 2 2 I Filter Rooms 3 3 O-GI l'I Radiation zone designation is based on resin slurry transfer piping. l AcaprownniDesign Masone! Redie61on Protection Page 12.3-29

System 80+ Design ControlDocument Table 12.3-3 Post-accident Acessibility Zone Designations Zone Designations Dose Rate (rem /hr) I less than 0.0025 2 0.0025 to 1.0 3 1.0 to 10.0 4 10.0 to 100.0 5 over 100.0 Table 12.3-4 Post-accident Radiation Zones DBA LOCA Room Zone No. Comments General Arrangement at El 50+0, Figure 12.3-9 Emer. Feedwater Motor Driven Pump Room 2 Emer. Feedwater Turbine Driven Pump Room 2 SCS Ileat Exchanger Rooms 5 CS IIcat Exchanger Rooms 5 Maintenance Aisle 5 Note 1 CS Pump Rooms 5 SCS Pump Rooms 5 CS Miniflow IIcat Exchanger Areas 5 SCS MiniflowliX Areas 5 St Pump Rooms 5 Nuclear Annex El 50+0, Figure 12.3 0 Corridor to FD (E-ll,11-12) 2 Corridor to FD (E-II,22-23) 2 Corridor Past FD 5,4,3 Note 2 VitalI&C Channel A, B 1 Vital I&C Channel C, D 2 Channel A, B Battery Room 1 Channel C, D Battery Room 2 Cable Chase A, B 2 Stairwell (D-E,1819) N/A Elevators E100 E200 N/A Elevators E300. E400 N/A Stairwell (B-C,11 12) N/A Stairwell (B-C, 22-23) N/A Maintenance Work Area (E-il,9-11) N/A Maintenance Work Area (E-ll,23-25) N/A Instrument Air Rooms N/A Approved Design Material Rodoation ,hotec%on l' age 12.3-30

t Sv tem 80+ Deskn ControlDocument Table 12.3-4 Post-accident Radiation 24nes (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 50+0, Figure 12.3-9 (Cont'd.) Sumps (H-J, 9-11) 5,4 Sumps (H J 22-23) 5,4 HVAC Chase (F-G,1314) 2 Storage Area (B-C,9-10) N/A Storage Area (B-C,24-25) N/A HVAC Chase (F-G,20-21) 2 CCW Pump Rooms (J-K,9-11) 2 CCW Pump Rooms (J-K,23-25) 2 Stairwell (0-P,11-12) 5,4 Note 2 Stairwell (0-P, 22-23) 5,4 Note 2 Elevator E500,600 5,4 Note 2 HVAC Chase (P-R,15-16) 3 Note 2 HVAC Chase (P-Q,18-19) 3 Note 2 HVAC Chase (M-N,11-12) 3 Note 2 HVAC Chase (N-0,22-23) 5,4 Note 2 O Pipe Chase (N-0, Il-12) 5,6 Note 2 Note 2 Pipe Chase (M-N,22-23) 3 Stairwell (S-T,11-12) N/A RD Pumps N/A EQ Drain Tank N/A Reactor Makeup Water Pumps N/A Charging Pump Rooms N/A CVCS Chem Add Pkg N/A Gas Stripper Control Panel N/A Charging Pump Mini Flow HX N/A Gas Stripper N/A Boric Acid Conc. N/A BAC Control Panel N/A Equip Drain Sump N/A Resin Sluice Tanks N/A RST Valve Access N/A Resin Sluice Pumps N/A Sampling Panels 5 Note 3 Sampling Panels Pipe Chase 5 Note 3 Floor Drain Sump (T-U,16-17) N/A Floor Drain Sump (S-T,18-19) N/A Anwoveet Deepn Mesuriet . Redenen Proteceion Page 12.3-31

System 80+ Design Control Document Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 50+0, Figure 12.3-9 (Cont'd.) CVCS Equipment Drain Sump (T-U,11 12) N/A CVCS Equipment Drain Sump (T-U,21-22) N/A CVCS Area (T-V,11-17) N/A Primary Chemistry Labs N/A Unidentified Area (R-U,22-23) N/A Diesel Generators N/A Equip Access Areas 3,4 Note 2 Janitorial /HP Storage and Work Area N/A Cont Cooler Cond Tanks N/A Reac. Bldg /Subsphere El 70+0, Figure 12.3-10 Fuel Pool Cooling Pumps 3 Note 2 Fuel Pool Heat Exchangers 3 Note 2 Pipe Chase 4,5 Note 4 Open to Elevation 50 Below 5 Inside Containment 5 Maint. Aisle / Valve Gallery 3 Note 2 Nuclear Annex El 70+0, Figure 12.3-10 Corridors to RAD Control Point (C-122-23) N/A Corridors t , RAD Control Point (C-I 11-12) N/A Storage Area (B-C,9-10) N/A Storage Area (B-C,24-25) N/A Elevators E100. E200 N/A Stairwell (B-C,10-1I) N/A Non-essential Chillers N/A Essential Chillers N/A Remote Shutdown Panel N/A Division 1 Battery Room N/A Division 2 Battery Room N/A Division 1 Channel Equipment (C-1,12-15) N/A Division 2 Channel Equipment (C-1, 20-22) N/A Cable Chases A, B, C, D N/A Stairwell (D E,18-19) N/A IIVAC Chase (F-G,20-21) N/A IIVAC Chase (F-G,13-14) N/A Elevators E300, E400 N/A Stairwell (B-C, 21-23) N/A Apyweved Desigrr Motorial Rodneters Protectroer Page 12.3-32

System 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 70+0, Figure 12.3-10 (Cont'd.) Reserved Areas for Cable (E-I,12-15) N/A Reserved Areas for Cable (E-I,17-22) N/A Reserved Areas for Cable (0-P,13-13a) N/A Reserved Areas for Cable (0-P,20-21) N/A EFW TarA Rooms N/A Entipuv:nt Access Shaft (S-T,1213) N/A Cable Area Channel A & B to Subsphere N/A Equipment Access Shaft (Q-R,21-22) N/A Elevators E500. E600 N/A Stairwell (0-P,11 12) N/A Stairwell (0-P, 22-23) N/A HVAC Chase (M N,11-12) N/A HVAC Chase (N-0,22-23) N/A Division 1 Channel Equipment (0-R,19-21 A) N/A Division 2 Channel Equipment (0-R,13-15) N/A O- Diesel Generators N/A - Fuel Pool Purif. IX's N/A Purification IX's N/A , Deborating IX N/A 3 Preholdup IX N/A l Boric Acid Conc IX N/A i Boric Acid Makeup Pumps N/A Pool Purification Pumps N/A l Holdup Pumps N/A l Pipe Chase (R-U,16-17) 5 Pipe Chase (R-T,17-20) N/A l Pipe Chase (N-0,11-12) N/A l Pipe Chase (M-N, 22-23) N/A Seal laj HX N/A CVCS Area Storage M/A Post-accident Hydrogen Recombiner Areas 4 Note 12 Valve Maintenance Shop N/A Unidentified Area (Q-U,20-23) N/A g Nuclear Annex El 81+0, Figure 12.3-11 h Fuel Pool Purification IX's Boric Acid Concentrator IX N/A N/A AppumpventCne$ atenend Medinaion Proteceion Pope 12.3-33 I I

System 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd,) DBA LOCA Room Zone No. Comments Nuclear Annex El 81+0, Figure 12.3-11 (Cont'd.) Valve Room (V-W,13A-17) N/A Purification IX's N/A Deborating IX N/A Pre-holdup IX N/A Valve Room (U-V,13 A-17) N/A Boric Acid Filter N/A RX Drain Filter N/A Pool Filters N/A Purif Filters N/A Valve Room (S-T,13A-17) N/A Valve Galleries N/A Seal Inj Filters N/A RMUW Filter N/A Valve Room (R S,13A-17) N/A Pipe Chase (P-R,16-18) 5 Note 5 Pipe Chase (R-S,16-17) 5 Note 5 Pipe Chase (S-T,17-18) N/A Equip Access Shaft N/A Stairs (S-T,11-12) N/A Stairs (T-V,1718) N/A HVAC Chase (P-R,15-16) N/A HVAC Chase (P R,18-19) N/A Electrical Equipment (0-R,13-15) N/A Reactor Building El 91+9, Figure 12.3-12 Pipe Chase 3 Note 2 Penetration Room A, B 3 Note 6 Penetration Room C, D 3 Note 6 Annulus Pipe Penetration Area 5 HVAC Distribution Header Areas 5 Reactor Drain Tank Room 5 RCP Motor Oil Drain Rooms 5 Elevator 5 Incore Chase 5 lioldup Volume 5 Reactor Vessel Cavity 5 Regenerative HX 5 Approwd DesJgn Materie! Radiositm Protection Page 12.344

f System 80+ oestan controlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) ~ DBA LOCA Room Zone No. Comments Reactor Building El 91+9 Figure 12.3-12 (Cont'd.) ; letdown HX 5 Nuclear Annex El 91+9, Figure 12.3-12 ! Corridor to Admin Bldg N/A Radiation Access Control Point N/A Stairwell (B-C,10-11) N/A Elevators E100, E200 N/A CAS Room N/A Stairwell (BC, 23 34) N/A Elevators E300, E400 N/A Corridor to RAD Control Points N/A Stairwell (D-E,18-19) N/A HVAC Chases (F-G,12-14) N/A IIVAC Chas (F-G,20-21) N/A Corridor Past RAD Control Point 1,2 Note 2 EFW Rooms N/A I Elevators E500, E600 N/A Stairwell (0-P,11 12) N/A Stairwell (0-P, 22-23) N/A Hot Machine Shop N/A ilVAC Chases (P-R,15-16) N/A j llVAC Chases (P R.18-19) N/A l Truck Bays N/A l Rail Car / Truck Washdown Areas N/A ] Equip Decon N/A l Eq Access Shafts N/A f Non<ssential Elect. Equip N2 (B-1,12-17) N/A Note 6 l Non<ssential Elect. Equip N1 (B-1,17 22) N/A Note 6 Non-essential Elect. Equip N2 (N-P,13-15) N/A Note 6 I N/A Note 6 l Non-essential Elect. Equip N1 (N-P,19-21; IX and Filter Hatch Areas N/A Pipe Chase (R T,19-20) N/A Pipe Chase (R-S,16-17) 5 Note 2 l Storage (R-T,18-19) N/A l 13.8K VAC RCP Switchgear (F-G,12-13) N/A l 13.8K VAC RCP Switchgear (F-G,21-22) N/A l Security Equipment N/A l 1 l

   . Approved Dennon neenwas! Renneaion hotechen                                        Page 12.3-35      }

System 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 91+9, Figure 12.3-12 (Cont'd.) liVAC Chase (M-N,11-12) N/A liVAC Chase (N-O,22-23) N/A Pipe Chase (N-O,11-12) N/A Pipe Chase (M-N, 22-23) N/A Reactor Building El 115+6, Figure 12.313 Annulus 5 Personnel Air Lock 5 Note 7 Area Outside Cranewall/Inside Containment 5 Between Cranewall, RX Vessel & S/G Shield 5 Core Support Barrel Laydown Area 5 Upper Guide Sttucture Laydown Area 5 Reactor Vessel Cavity 5 Refueling Canal 5 Fuel Transfer Tube 5 Nuclear Annex El 115+6, Figure 12.3-13 Stairwell (B-C,11-12) N/A Elevators E100, E200 N/A Personnel Decon 5,4,3,2 Note 8 Break Room 1 Note 8 Men's Restroom 1 Note 8 Women's Restrootn i Note 8 Corridor to RAD Control Point (B-E,12-13) 1 Corridor To RAD Control Point (B-E,20-21) 1 Corridor Past RAD Control Point (E-I,12-13) 3,4,5 Note 8 Corridor Past RAD Control Point (E-1,20-21) 3,4,5 Note 8 Control Room 1 Tool Room N/A Document /ISPD Room & Emerg Supplies N/A Storage Room (E-G,18-20) N/A Stairwell (D-E,18-19) N/A Computer Room N/A Reactor Operator's Office N/A Control Rm Supervisor's Office N/A Shift Supervisor's Office N/A Clerk's Office N/A Approved Desiger Material . Radiation Protec ten Page 12.3-36

'l

Sy~ tem 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments

Nuclear Annex El 115+6, Figure 12.313 (Cont'd.) Tag Out Area N/A Shift Assembly Area N/A Elevators E300, E400 N/A Stairwell (B{. 22-23) N/A Conference Rooms N/A - IIVAC Chase (F-G,13-14) 4 Note 8 HVAC Chase (F-G, 20-21) N/A Ilot Machine Shop N/A Hot Tool Crib 1 Note 8 Hot Instrument Shop N/A Main Steam Valve llouses 1 and 2 N/A HVAC Chases (P-R,15-16) 5 Note 9 HVAC Chases (P-R,18-19) 5 Note 9 Refuelig Canal N/A N/A

]/

Spent Fuel Pool (Bottom) Cask Laydown Area (Bottom) N/A Cask Washdown Area N/A Elevator E500 3 Note 9 Elevator E600 5 Note 9 Stairwell (0-P,11-12) 5 Note 9 Stairwell (0-P, 22-23) 5 Note 9 Equipment Access Shaft 5 Note 9 Unidentified Area (R-W,13-17) N/A l Volume Control Tank N/A l VCT Valve Access N/A Boric Acid Batch Tank N/A l Pipe Chase (R.S,16-17) 5 Note 2 Subsphere Exhaust 5 Note 9 l Pipe Chase (N-0,11-12) 5 Pipe Chase (M-N,22-23) 5 Storage Area (Q-R,23-24) N/A Nuclear Annex El 130+6, Figure 12.3-14 Stairwell (B-C,10-1!) N/A , Conference Room (B-D,19-21) N/A l OSC's HVAC Room N/A TSC Area N/A' Agnpwwwed Deny Meterial Madeeien hotecnion Page 12.3-37

System 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 130+6, Figure 12.3-14 (Cont'd.) OSC Room N/A TSC Ventilation Equipment Room N/A Control Room N/A Control Room }{VAC Areas N/A Stairwell for Control Room (D-E,18-19) N/A Elevators E100. E200, E300 E400 N/A Elevator E500 $ Note 10 Elevator E600 5 Note 10 Stairwell (0-P, Il 12) 5 Note 10 11VAC Chases (P-R,15-16) 5 Note 10 IIVAC Chases (P-R,18-19) 5 Note 10 EQ Access Areas 4 Note 10 Stairwell (0-P, 22-23) 5 Note 10 New Fuel Unloading N/A New Fuel Storage N/A Cask Laydown Area N/A Cask Washdown Area N/A Nuclear Annex Supply 5 Note 10 Nuclear Annex Exhaust 5 Note 10 Instrument Calibration Shop 2, N/A Annulus Exhaust 5 Note 10 Respirator Maintenance Shop N/A Main Steam Valve flouses 1 and 2 N/A Volume Control Tank N/A HVAC Chase (N-0,22-23) 5 Note 10 llVAC Chase (P-Q,11 12) 5 Note 10 Pipe Chase (N-0,11 12) 5 Note 10 hpe Chase (M-N,22-23) 5 Note 10 Reactor Building El 146+0, Figure 12.3-15 Annulus 5 Area Between CV and Crane Wall 5 Area Between Crane Wall & Primary Shield 5 Inside Primary Shield 5 Core Support Barrel Laydown Area 5 Upper Guide Structure Laydown Area 5 Personnel lock 5 Note 7 Approved Desigrs Mat %I- Re&aten Protection Page 12.3-38

Sy tem 80 + Desion controlDocument ( Table 12.3-4 Post-accident Rsdiation Zones (Cont'd.) DBA LOCA Room Zoac No. Comments Nuclear Annex El 146+0, Figure 12.3-15 Elevator Machine Room (B-C,11-12) N/A Elevator Machine Room (B-C, 22-23) N/A Stairwell (B-C,10-11) N/A Stairwell (B-C, 23-24) N/A Laydown Work Space N/A Elevator E500 4 Note 7 Elevator E600 N/A Stairwc!! (0-P, Il-12) N/A Stairwell (0-P, 22-23) 4 Note 7 Hot Tool Cribs 4. N/A Note 7 Personnel Air Lock 5 Hallway Outside Personnel Air lock 5,4 Fuel Pool Laydown & Equip Area 3 Fuel Pool Surface N/A

  /   IIVAC Chases (N-O, 22-23)                                 5 5   HVAC Chases (P-R,15-16)                                 N/A HVAC Chases (P-R,18-19)                                 N/A Equipment Access Shafts                                4, N/A Pipe Chase (N-O,11 12)                                  N/A Pipe Chase (M-N,22-23)                                    5 Personnel Decon Area                                    N/A Reactor Building El 170+0, Figure 12.3-16 Annulus                                                  5,4 Between Cont. Vessel & Cranewall                          5 inside Cranewall                                          5 Steam Generator Rooms                                     5 Nuclear Annex El 170+0, Figure 12.316 Elevator E500                                             5 Stairwell (0-P, 22-23)                                    5 Laydown/ Staging Space (S-W,1315)                       N/A HVAC Equipment Area                                     N/A IIVAC Chases (P-R,15-16)                                N/A HVAC Chases (P-Q,18-19)                                   5 HVAC Chases (Q-R,18-19)                                 N/A N/A

() HVAC Chases (V W 16-17) HVAC Chases (N-0,22-23) 5

     %.a nwen neonard nasosan notecaion                                         rose 12.3-29

System 80+ Design ControlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Room Zone No. Comments Nuclear Annex El 170+0, Figure 12.3-16 (Cont'd.) Fuel Pool HVAC Supply Area (V-W,15-16) N/A Fuel Pool Exhaust (Q-R,19-21 A) N/A ; Fuel Pool Exhaust (U-V,16-17) N/A 111 Purge Supply (T-V,15-16) N/A Note 11 ID Purge Supply (R-T.15-16) N/A Nuclear Annex Exhaust Areas (N-P,19-22) 5 Pipe Chase (M-N,22-23) 5 Pipe Chase (N-0,11-12) N/A Equipment Access Shaft N/A 111 Purge Exhaust Areas (R-T,16-17) N/A Note 11 LO Purge Exhaust Areas (P-R,16-17) N/A Note 11 CCW Surge Tank Room (L-M,22-23) 4 CCW Surge Tank Room (N-0,11-12) N/A Corridor (P-Q,17-23) 4 Turbine Building , Blowdown Demineralizer Cubicles N/A Condensate Demineralizer Cubicles N/A Flash Tank Room N/A Pipe ChasesD3 N/A Turbine Building General Area N/A Radwaste Building El 34+0 & 50+0, Figure 12.3-18 Chemical Waste Tank (CWT) Room N/A CWT Pump Rooms N/A Chemical Sample Tank (CST) Room N/A CST Pump Rooms N/A Detergent Sample Tank (DST) Room N/A DST Pump Room N/A Laundry & Hot Shower Tank (LilST) Room N/A LilST Pump Room N/A Laundry Filter Area N/A Waste Monitor Tank (WMT) Rooms N/A WMT Pump Rooms N/A Demineralizer Rooms N/A Equipment Waste Tank (EWT) Rooms N/A ! I'3 Radiation zone designation is based on resin slurry transfer piping. AMvoved Design Matadel- Radiathws Protection Page 12.340 l

System 80+ onion controlDocument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) D A LOCA Room Zone No. Comments Radwaste Building El 34+0 & 50+0, Figure 12.3-18 (Cont'd.) EWT Pump Room N/A EWT Filter Area N/A Floor Drain Tank (FDT) Rooms N/A ; FDT Pump Room N/A Oil Drum Storage Area N/A N/A Waste Filter Area Einpty HIC Storage Area N/A HIC Dewatering /Washdown/ Inspection / labeling and Transfer Area N/A Remote Operation Room NlA General Area N/A law Activity Spent Resin Tank (LASRT) Room N/A ! High Activity Spent Resin Tank (HASRT) Rooms N/A Pipe Chase (F-G,7-8) N/A Radweste Building El 70+0, Figure 12.3-19 t CWT Room N/A CST Room N/A General Area N/A WMT Rooms N/A EWT Rooms N/A FDT Rooms N/A Oil Separator Rooms N/A Chemical Lab N/A LASRT Room N/A f HASRT Rooms N/A Pipe Chase (F-G,7 8)l4 N/A Radwaste Building El 91+9, Figure 12.3-20 , Solid Waste Storage Area N/A N/A Empty Box Area N/A N/A ) Full Cask Storage Area N/A N/A j HIC Storage Area N/A N/A I I Corridor (E-F, 8-11) N/A N//. 1 Decontamination Area N/A N/A Low and High Activity Spent Resin Surge Tank Room N/A N/A Radwaste Building Access Control Area N/A N/A , J IU Radiation zone designation is based on resin slurry transfer piping. L . .;:Donen aneserier . nonesien nereceron rope 12.3-4 r .

-1

System 80+ Design Contro/ D'cument Table 12.3-4 Post-accident Radiation Zones (Cont'd.) DBA LOCA Zone No. Comments Room Radwaste Building El 91+9, Figure 12.3-20 (Cont'd.) Counting Room N/A N/A Radwaste Building El 115+6, Figure 12.3-21 IIVAC Equipment Room N/A Control Room N/A Electrical Equipment and Battery Room N/A Corridors (G-11, 4-10) (G-1, 3-4) N/A Filter Rooms N/A Notes / Comments: General Assumptions: Zone ids were assigned to areas based on the present general arrangement drawings. Post-LOCA source terms are described in Table 12.2-20. Post-accident zones reflect the contribution from post-accident sources only. Normal operation sources have not been added. Therefore, areas may show lower zoned designations for post-accident conditions when in reality the normal operation sources would indicate higher zones. i

1. Assumes post-accident piping for the CS and SCS equipment is in the area.

2. Assumes scatter and streaming due to post-accident piping in the area.

3. Assumes post. accident sampling is conducted in this area.

4. Assumes post-accident piping is in the area.

5. Assumes post-accident sampling piping is in the area.

6. Assumes minimal or no unshielded penetrations in the area.

7. Personnel hatch provides a major streaming path out of the reactor building.

8. Post accident streaming and scatter as a result of the personnel hatch nearby.

9. Assumes subsphere ventilation filters are in this area and are operating post LOCA.

10. Assumes the annulus and nuclear annex ventilation filters are located in this area and operate post-accida

11. Assumes the purge ventilation systems does not operate post LOCA. ,

i

12. Post accident hydrogen recombiners are portable and will be installed within 72 hours after the initiation of the accident.

N/A. Radiation zone not impacted in post-accident condition (i.e., radiation zone remains same as in normal operation conditions). i O Approved Dasbprs Motorial Ra&sticos Protectiors Page 12.3 42 l l i

t l Syztem 80+ Design ControlDocument Table 12.3-5 Area Monitor Locations Location l Number of Monitors Nuclear Annex El 50+0 Normal Operation Floor Drain Sump Room (U,17)(T,18) 2 Sample Panel Room (S,16) 1 ; CVCS Area (U,14) ! . Primary Chemistry Lab Area (S,10) 1 l I Personnel Corridors (0,12) (P,21a) 2 Personnel Corridors (11-1, 22) (11 1,12) 2 , Post Accident ! Sample Panel Room (S.16) 1 Personnel Corridors (0,12) (P 21a) 2 Personnel Corridors (H-1, 22) (H-1,12) 2 Subsphere El 50+0 Nonnat and Post-Accident : Maintenance Isle (45 *,135', 225', 315') 4 Nuclear Annex El 70+0 Q Nonnal and Post-Accident CVCS Storage Area (R,10) i Valve Maintenance Shop (R,24) 1 l Control Complex El 70+0 ' l Area Adjacent to Remote Shutdown Panel (E.17) 1 , Subsphere El 70+0 l Nonnal and Post Accident l Subsphere Entrance (230',130') 2 l

Subsphere Maintenance Aisle (45',315') 2 l Nuclear Annes El 81+0 l Nonnat and Post-Accident l Valve Gallery Entrances (V,14) (S.16-17) 2 Purification Filter Cubicles (T,15-16)(T 16)Iil 2 Reactor Building El 91+9 incore Chase (1-J, 90') I til Used primarily as an indication of crud burst from the Reactor Coolant System and to prevent activity on the filter from exceeding Class C waste concentrations. It also provides guidance for ALARA filter changeouts.

4pemt oman uneraer- ne seem herecem rare 12.343

Syster.: 80 + Design ControlDocument Table 12.3-5 Area Monitor Locations (Cont'd.) Location l Number of Monitors Nuclear Annex El 91+9 Normal Operation Equipment Decontamination Room (U,12) I liot Machine Shop (S 9) 1 Resin Storage Area (V,15)(S.15) 2 Fuel Pool Storage Area (R,24) 1 Personnel Corridors (P.12) (P. 21a) 2 Post-Accident ~ Personnel Corridors Outside Subsphere Entrance (90*,270') l 2 Subsphere El 91+9 Normal and Post Accident Penetration Rooms A, B, C, D l 4 Nuclear Annex El 115+0 Normal llot Instrument Shop (U,12) 1 Spent Fuel Pool Refueling Bridge (S,18) 1 Cask Laydown and Washdown (S-T,21) 1 Personnel Corridor Adjacent to Spent Fuel Pool (U 19) 1 Area Adjacent to VCT Room (Va,16) 1 Post-Accident Personnel Corridors (P, 21a) (P,12) 2 Personnel Corridor Outside Personnel Lock (F.12-13) 1 Main Steam Valve llouses 1 & 2 (K, 24) (K.10) 2 Subsphere El 115+0 Nonnal Operation Personnel Corridor Adjacent to Spent Fuel Transfer Tube inspection Area (180') l 1 Reactor Building El 115+0 Post Accident Primary Coolant Loop Monitors inside Crane Wall (90*,270*) l 2 Control Complex El 115+0 Normal and Post-Accident Control Room (E,16-17) l 1 Nuclear Annex El 130+6) Normal Openition New Fuel Storage Area (T,24)(T-S,24) 2 Personnel Corridors Adjacent to Annulus Exhaust Rooms (Q,13)(P-Q,22-23) 2 Personnel Corridor Adjacent to Nuclear Annex Exhaust Rooms (S,14) 1 Control Complex El 130+6 Normal and Post-Accident TSC Area (B,17) l 1 Appresed Des &n Material Radebon Protection Page 12.344

V System 80+ Design ControlDocument / (J Table 12.3-5 Area Monitor Locations (Cont'd.) Location Number of Monitors ,. OSC Area (D,15) I ! Reactor Building El 140+0 Normal Operation Refueling Bridge Crane l 1 Nuclear Annex El146+0 Normal and Post-Accident Laydown/ Work Space (S.13a) 1 Personnel Corridor Outside Personnel leck (P, 22) I j Entrance to Spent Fuel Pool Area (R,20-21) 1 Personnel Decontamination Area (T,17) 1 - Nuclear Annex El 170+0 Normal and Post-Accident ; lew-Purge Exhaust Room (Q,17) 1 Personnel Corridor (P. 22) 1 Reactor Building El 200+0 t Post Accident High Range Containment Monitors (90',270*) (Outside the Crane Wall) l 2 Radwaste Buildingl 'l Normal Operation El 34+0 and El 50+0 HIC Inspection /Washdown Area and Empty filC Storage Area 1 General Access Area at El 50*+0 1 El 70+0 Chemical Lab 1 ; General Access Area at El 70*+0 1 General Access Area in Future Permanent Solid Waste Area 1 El 91 +0 Truck Bay 1 Dry Waste Compacting Area 1 Liquid Waste System Mobile demin. area 1 Counting Room 1 El 115 +0 Control Room i O I'l For funher information on location, see Figures 12.3-18 through 12.3 21. Appreewd Design Meterial. Rodina~en Protection Page 12.345

System 80+ Design ControlDocument Table 12.3-6 List of Accessible Areas Potentially >100 R/HRlll h Area Coordinates Nuclear Annex, El 50+0, Figure 12.3-1 R-S,13-14 Equipment Drain Tank Room Va-W,14-17 Resin Sluice Tarl Rooms t2 Va-W,13a-14 & 16-17 Resin Sluice Tank Valve Access ) Ua-Va,14-17 Resin Sluice Pump Roomst21 __ Nuclear Annex, El 70+0, Figure 12.3-2 Purification lon Exchanger cubicles T-U,15-16 R-S,16-17 Pipe Chaset21 Reactor Building, El 70+ 0, Figure 12.3-4 I-1, O' Incore Chase Reactor Vessel Cavity Reactor Building,115+6, Figure 12.3-5 Between Cranewall, Rx Vessel & S/G Shield Core Support Barrel Laydawn L-M,180' Area El 96+9 Reactor Vessel Cavity M-P,180* Fuel Transfer Tubet31 Area Between Cranewall and Primary Shield Inside Primary Shield Upper Guide Structure Laydown I-J, O' Area El 104+6 Steam Generator Rooms J L, 90', 270' Nuclear Annex, El 115+6, Figure 12.3-5 Refueling Canal I33 P-V,17-18 Fuel Transfer Tube!33 P-Q,17-18 Spent Fuel Pool (Bottom - El 104 +0) R-U,18-20 Cask Laydown and Washdown Areasf33 (Bottom -- El 101 +0,119 +0) R-U, 20-21 IU During normal operating conditions and anticipated operational occurrences. 121 Only durmg spent resin transfer. til Only when fuel is in area. Approwd Design Material- Radiation Protectieur Page 12.346

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