ML21210A136

2 to Updated Final Safety Analysis Report, Chapter 11, Radioactive Wastes and Radiation Protection (EPID L-2020-LRO-0076) - Redacted

ML21210A136
Person / Time
Site:	Surry
Issue date:	09/30/2020
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
Thomas V
Shared Package
ML21208A006	List: ML21208A367 ML21208A409 ML21208A418 ML21208A424 ML21208A429 ML21210A125 ML21210A126 ML21210A127 ML21210A128 ML21210A129 ML21210A133 ML21210A135 ML21210A136 ML21210A137 ML21210A138 ML21210A139 ML21210A140
References
20-325, EPID L-2020-LRO-0076
Download: ML21210A136 (116)

Text

Surry Power Station Updated Final Safety Analysis Report Chapter 11

Intentionally Blank Revision 52Updated Online 09/30/20 SPS UFSAR 11-i Chapter 11: Radioactive Wastes and Radiation Protection Table of Contents Section Title Page 11.1 GENERAL DESCRIPTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-2 11.2 RADIOACTIVE WASTE SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.2 System Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.3 Liquid Waste Disposal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-3 11.2.3.1 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-4 11.2.3.2 Processing Steam Generator Blowdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-8 11.2.4 Solid Waste Disposal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-9 11.2.4.1 Solid Waste Handling Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-9 11.2.4.2 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-11 11.2.5 Gaseous Waste Disposal System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-13 11.2.5.1 Process Vent Subsystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-13 11.2.5.2 Ventilation Vent Subsystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-16 11.2.5.3 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-16 11.2.6 Tests and Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-18 11.2.6.1 Construction and Fabrication. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-18 11.2.6.2 Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-19 11.2 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-20 11.3 RADIATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.1.1 Leak Reduction Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.2 Shielding Design and Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-2 11.3.2.1 Primary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-2 11.3.2.2 Secondary Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-3 11.3.2.3 Reactor Coolant Loop Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-4 11.3.2.4 Reactor Containment Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-4 11.3.2.5 Fuel Handling Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-4 11.3.2.6 Auxiliary Equipment Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.3.2.7 Waste Storage Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.3.2.8 Accident Shielding. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-6 11.3.2.9 Postaccident Shielding Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-6 11.3.2.10 Steam Generator Storage Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-8

Revision 52Updated Online 09/30/20 SPS UFSAR 11-ii Chapter 11: Radioactive Wastes and Radiation Protection Table of Contents (continued)

Section Title Page 11.3.3 Process Radiation Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-9 11.3.3.1 Process Vent Particulate Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-11 11.3.3.2 Process Vent Gas Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-11 11.3.3.3 Ventilation Vent No. 1 Gas Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-11 11.3.3.4 RF Vent Particulate and Gas Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-11 11.3.3.5 Component Cooling Water Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-11 11.3.3.6 Component Cooling Heat Exchanger Service Water Monitor . . . . . . . . . . . . 11.3-12 11.3.3.7 Liquid Waste Disposal System Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-12 11.3.3.8 Condenser Air Ejector Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-12 11.3.3.9 Steam Generator Blowdown Sample Monitors . . . . . . . . . . . . . . . . . . . . . . . . 11.3-12 11.3.3.10 Recirculation Spray Cooler Service Water Outlet Monitors . . . . . . . . . . . . . . 11.3-13 11.3.3.11 Reactor Coolant Letdown Gross Activity Monitors . . . . . . . . . . . . . . . . . . . . 11.3-13 11.3.3.12 Circulating Water Discharge Tunnel Monitors . . . . . . . . . . . . . . . . . . . . . . . . 11.3-13 11.3.3.13 Ventilation Vent No. 2 Particulate and Gas Monitors. . . . . . . . . . . . . . . . . . . 11.3-14 11.3.3.14 High-Range Postaccident Radiation Monitors . . . . . . . . . . . . . . . . . . . . . . . . 11.3-14 11.3.3.15 Storm Drain Radiation Sample System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-15 11.3.4 Area Radiation Monitoring System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-15 11.3.4.1 General. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-15 11.3.4.2 Containment Particulate Monitors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-16 11.3.4.3 Containment Gas Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-17 11.3.4.4 Other Area Radiation Monitoring Equipment . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-17 11.3.4.5 High-Range Postaccident Containment Monitors . . . . . . . . . . . . . . . . . . . . . . 11.3-18 11.3.4.6 Technical Support Center Area Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-18 11.3.5 Environmental Survey Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-18 11.3.5.1 Air Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-20 11.3.5.2 Milk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-21 11.3.5.3 Shellfish, Crabs, and Fish . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-21 11.3.5.4 Silt and Sediment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-21 11.3.5.5 Water Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-21 11.3.5.6 Food Products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-22 11.3.5.7 Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-22 11.3.5.8 Environmental Dosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-22 11.3.5.9 Interlaboratory Comparison Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-22 11.3.6 Control Room . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-23 11.3 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-23

Revision 52Updated Online 09/30/20 SPS UFSAR 11-iii Chapter 11: Radioactive Wastes and Radiation Protection Table of Contents (continued)

Section Title Page 11.3 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-25 Appendix 11A Radiation Exposure Evaluation for Expected Radioactive Effluents . . . 11A-i 11A.1 ANALYTICAL BASIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-1 11A.2 INPUT INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-1 11A.2.1 General Plant Information. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-1 11A.2.2 Meteorological Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-2 11A.2.3 Dispersion Factors and Ground Deposition Factors . . . . . . . . . . . . . . . . . . . . . . 11A-2 11A.3 PLANT AND ANIMAL CENSUS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-3 11A.4 DOSE CALCULATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-3 11A.4.1 Doses From Liquid Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-3 11A.4.2 Doses From Gaseous Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-4 11A References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-5

Revision 52Updated Online 09/30/20 SPS UFSAR 11-iv Chapter 11: Radioactive Wastes and Radiation Protection List of Tables Table Title Page Table 11.2-1 Waste Disposal Systems Design Data . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-21 Table 11.2-2 Estimated Waste Gas Release With Recombiner Not Operating . . . . . 11.2-33 Table 11.2-3 Estimated Waste Gas Release With Recombiner Operating . . . . . . . . . 11.2-34 Table 11.2-4 Liquid Waste Disposal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-35 Table 11.3-1 Allowable Dose Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-26 Table 11.3-2 Containment Shielding Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-27 Table 11.3-3 N-16 and Activated Corrosion Product Activity Used in the Original Plant Shielding Design Analysis . . . . . . . . . . . . . 11.3-28 Table 11.3-4 Postaccident Mitigation Phase Source Terms . . . . . . . . . . . . . . . . . . . . 11.3-28 Table 11.3-5 Process Radiation Monitoring Systema . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-29 Table 11.3-6 Process Radiation Monitoring System Counting Rates of Limiting Isotopes (cpm/mCi/cm3). . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-31 Table 11.3-7 Area Radiation Monitoring Locations, Number, and Ranges . . . . . . . . 11.3-32 Table 11.3-8 High-Range Post-Accident Radiation Monitors . . . . . . . . . . . . . . . . . . 11.3-33 Table 11A-1 General Plant Information* . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-7 Table 11A-2 Monthly t150 ft - 35 ft Stability Distribution (%) . . . . . . . . . . . . . . . . . . 11A-8 Table 11A-3 /Q And D/Q Values at Special Distances and Release Modes for a 2-Year Data Perioda. . . . . . . . . . . . . . . . . . . . 11A-9 Table 11A-4 Annual Average /Q (sec/m3) Values Based on a Ground-Level Release for a 2-Year Data Perioda . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-10 Table 11A-5 Annual Average D/Q (M-2) Values Based on a Ground-Level Release for a 2-Year Data Perioda . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-11 Table 11A-6 Annual Average /Q (Sec/m3) Values Based on a Mixed-Mode Release for a 2-Year Data Perioda . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-12 Table 11A-7 Annual Average D/Q (M-2) Values Based on a Mixed-Mode Release for a 2-Year Data Perioda . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-13 Table 11A-8 0- to 5-Mile Highpoints by Mile and 5- to 10-Mile Highpoints for 16 Cardinal Points From Surry Power Stationa . . . . . . . . . . . . . . . . . . . 11A-14 Table 11A-9 Liquid Source Terms From Surry Units 1 & 2 (Per Unit Basis)

Without Steam Generator Blowdown Processing . . . . . . . . . . . . . . . . . 11A-15 Table 11A-10 Liquid Source Terms From Surry Units 1 & 2 (Per Unit Basis)

With Steam Generator Blowdown Processinga . . . . . . . . . . . . . . . . . . . 11A-19

Revision 52Updated Online 09/30/20 SPS UFSAR 11-v Chapter 11: Radioactive Wastes and Radiation Protection List of Tables (continued)

Table Title Page Table 11A-11 Ladtap Input Data and Results - Maximally Exposed Individual Dose Calculations for Surry Units 1 and 2 (Demineralizer Radwaste System Without Steam Generator Blowdown Treatment) . . . . . . . . . . . . . . . . . 11A-23 Table 11A-12 Ladtap Input Data and Results - Maximally Exposed Individual Dose Calculations for Surry Units 1 and 2 (Modified Liquid Radwaste System With Steam Generator Blowdown Treatment)a . . . . . . . . . . . . . . . . . . . 11A-24 Table 11A-13 Gaseous Source Terms From Surry Units 1 and 2 (Per Unit) . . . . . . . . 11A-25 Table 11A-14 Maximum Doses to an Individual Resulting From Gaseous Effluents From Surry Units 1 And 2 With Steam Generator Flash Tank (mrem/yr per Unit) . 11A-27 Table 11A-15 Maximum Doses to an Individual Resulting From Gaseous Effluents From Surry Units 1 and 2 With Steam Generator Blowdown Cooled (mrem/yr per Unit) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11A-28

Revision 52Updated Online 09/30/20 SPS UFSAR 11-vi Chapter 11: Radioactive Wastes and Radiation Protection List of Figures Figure Title Page Figure 11.2-1 Liquid Waste Disposal System; Process Flow Diagram, Sheet 1 . . . . 11.2-36 Figure 11.2-2 Liquid Waste Disposal System; Process Flow Diagram, Sheet 2 . . . . 11.2-37 Figure 11.2-3 Liquid Waste System Process Flow Diagram . . . . . . . . . . . . . . . . . . . 11.2-38 Figure 11.2-4 Laundry Waste System Process Flow Diagram . . . . . . . . . . . . . . . . . . 11.2-39 Figure 11.2-5 Bitumen Solidification System Process Flow Diagram. . . . . . . . . . . . 11.2-40 Figure 11.3-1 Shield Arrangement, Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-34 Figure 11.3-2 Shield Arrangement, Elevation, Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . 11.3-35 Figure 11.3-3 Shield Arrangement, Elevation, Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . 11.3-36

Revision 52Updated Online 09/30/20 SPS UFSAR 11.1-1 CHAPTER 11 RADIOACTIVE WASTES AND RADIATION PROTECTION 11.1 GENERAL DESCRIPTION Note: As required by the Renewed Operating Licenses for Surry Units 1 and 2, issued March 20, 2003, various systems, structures, and components discussed within this chapter are subject to aging management. The programs and activities necessary to manage the aging of these systems, structures, and components are discussed in Chapter 18.

Waste disposal systems are provided to separate, treat, and dispose of radioactive liquid, gaseous, and solid waste materials. The liquid, solid, and gaseous waste disposal systems are common to both reactor units and designed to serve both units simultaneously. These systems incorporate one or more of the following basic processes:

1. Filtration, to remove particulate matter.

2. Evaporation, to concentrate and remove contaminants.

3. Demineralization, to remove dissolved material.

4. Compaction, to reduce the volume of compressible wastes.

5. Natural decay of radioactive isotopes.

6. Dilution, to reduce concentration.

Liquid, gaseous, and solid waste materials originate in the reactor coolant system, the auxiliary and emergency systems, the waste disposal system, and as a result of operation and maintenance procedures. Waste materials enter the waste disposal system directly from their source or via the vent and drain system (Section 9.7).

Adequate sampling, analysis, and monitoring of the waste disposal system are provided to comply with the design criteria. Process radiation monitors and flow-measuring equipment are provided for the surveillance of various station and radwaste effluents and process streams to ensure compliance with applicable regulations and to provide early indications of possible malfunctions and hazardous conditions.

Sufficient shielding is provided to reduce radiation to acceptable levels for normal operation and incident conditions. Allowable dose rates are based on applicable regulations, expected frequency, and the duration of exposure to radiation.

Area radiation monitoring equipment, health physics facilities, environmental programs, and administrative controls are provided for the surveillance and control of radiation exposure levels. These ensure radiation protection for plant personnel and the general public in accordance with applicable criteria.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.1-2 Radiological and chemical respiratory protection equipment approved by the National Institute for Occupational Safety and Health/Mine Safety and Health Administration (NIOSH/MSHA) is provided. Equipment not tested and certified by NIOSH/MSHA requires specific authorization by the NRC and an approved exemption from 10 CFR 20.1703(a)(1),

10 CFR 20.1703(c), and certain parts of 10 CFR Part 20, Appendix A, Protection Factors for Respirators, Footnote d.2.(d) before use. Authorization has been received and appropriate exemption granted for the use of MSA Model Firehawk M7 SCBA, ultralite respirator, and 4500 psi tank charged with 35% oxygen and 65% nitrogen. All units are to be equipped with rubber face-pieces. Regulator use is not to be initiated at temperatures >135°F. Units may be used in areas where temperatures exceed 135°F if regulator use is initiated prior to entry into those areas. Breathing gas quality and composition, including hydrocarbon exclusion, are insured by strict controls and maintained in accordance with the latest revision of the United Stated Pharacopeia (USP) - The National Formulary (NF).

Prior to Unit 1 operation, a radiological study of the environs was performed (Section 11.3.5). It included an investigation of the background radiology relating to various forms of the aquatic and terrestrial environment. The nature and extent of the postoperational environmental survey were determined from the results of the preoperational study.

11.1 References

1. Letter from Karen Cotten, USNRC, to David A. Heacock, Virginia Electric & Power, May 28, 2010, North Anna Power Station, Unit Nos. 1 and 2 And Surry Power Station, Units 1 and 2, Exemption From Certain Requirements of 10 CFR 20 (TAC Nos ME2835, ME2836, ME2828 and ME2829), Serial No.10-363.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-1 11.2 RADIOACTIVE WASTE SYSTEMS 11.2.1 Design Bases It is Vepcos waste management policy to maintain radioactive waste effluent from the Surry Power Station at the lowest practical level. In keeping with this policy, the Radioactive waste disposal system is designed, to the extent possible in accordance with maintenance practices, to maintain releases of radioactive material and radiation exposures to unrestricted areas as far below the limits of 10 CFR 201 as is practical. Normally, no radioactive waste stream will be discharged from the station without having first been processed through the waste disposal system.

The liquid, solid, and gaseous waste disposal systems are common to both reactor units.

Each waste disposal system is designed to accommodate radioactive wastes produced during simultaneous operation of the two units. Both units are assumed to be operating on a daily load follow cycle using boric acid between 100% and 50% power.

The systems are also designed to accommodate the corrosion products originating in the reactor coolant system, and not removed in other systems.

11.2.2 System Design The waste disposal system and radiation monitoring system are designed to satisfy the applicable sections of the general design criteria of Section 1.4. In addition, these systems are designed to limit the discharge of radioactive materials from the station so as not to exceed the limits of 10 CFR 20 or the suggested criteria of 10 CFR 100, and so as not to endanger the health of station operating personnel. The transportation of radioactive materials from the station is carried out in such a manner as to conform with applicable Federal, state, and local ordinances.

Design data are given in Table 11.2-1. An evaluation of the waste disposal systems in accordance with the requirements of 10 CFR 50, Appendix I, is provided in Appendix 11A.

The liquid waste disposal system is described in detail in Section 11.2.3. This system has been designed to ensure that the release of radioactivity to the environment will be kept at the lowest practical level.

All normally radioactive waste gases from the gaseous waste disposal system, the gas stripper in the boron recovery system, the vent and drain system, various pressure relief valves, and the containment vacuum system are regulated before discharge by the process vent subsystem, as described in Section 11.2.5.1. All these sources of gaseous effluent are, before discharge collected; diluted; filtered through charcoal filters; monitored for flow rate, pressure, temperature, and particulates and gaseous activity; and then released through the process vents.

1. Virginia Power implemented the revised 10 CFR 20 on January 1, 1994. However, as allowed by the NRC, the calculational methodology used for the design analyses is based on the revision of 10 CFR 20 to which the plant was originally licensed.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-2 Gas stripped from liquids entering the boron recovery system is stored in decay tanks as discussed in Section 11.2.5. The process vents and the process vent blowers are sized such that the minimum exit velocity is approximately 100 fps, which prevents any significant downdrafting of the effluent.

Radioactive waste gases may also be present in the Radwaste Facility (RF) ventilation system. Minor amounts of noble gases may be entrained in the liquid waste being processed by the RF. If evaporation of the liquid waste is the process method used, these gases will be released to the RF ventilation system. The RF ventilation system is designed to process and monitor these gases and any other airborne activity produced in the facility process areas. All tanks and process equipment vents which have radioactive contents are connected to the tank vent system. This system has a demister, a charcoal filter, and high-efficiency particulate air (HEPA) filter bank.

The general area ventilation for the radiological controlled areas passes through a HEPA filter bank prior to discharge. Both the general area and the tank vent ventilation systems discharge through the RF vent stack and are continuously monitored. The exit velocity of the RF vent stack is in excess of 3200 fpm.

Other gaseous effluents will normally not be radioactive during operation; however, ventilation exhaust from some primary plant areas is subject to comparatively slight radioactive contamination from such limited sources as pump gland or pipe weepage. Features are incorporated in these exhaust systems to protect the environment from these relatively remote contamination possibilities. Three filter banks (two common and one stand alone) are capable of handling the largest possible exhaust ventilation flow rate that can be aligned through the individual filter bank(s). Each filter bank consists of roughing, HEPA, and charcoal filters.

Exhaust bypass arrangements for selective filtration of any exhaust add to the flexibility of the system. All bypass and filter dampers are remote manually operated from the control room as required. Ventilation exhaust trains for the safeguards areas and charging pump cubicles automatically realign on a safety injection signal as discussed in Section 9.13.4.1.

The process vent, ventilation vents, and the RF vent are continuously monitored so that the effluent activity release rates result in concentrations considerably less than those limits provided by 10 CFR 20 at the site boundary. The gaseous waste disposal system is designed to provide adequate radioactive decay storage time for the waste gases prior to discharge through the process vent and, in addition, provides sufficient capacity to allow adequate holdup of these gases even when high-flow letdown is required.

The estimated releases from the gaseous waste disposal system are based on assumptions, discussed in Section 11.2.5, regarding the operation of the system. These assumptions were developed during the original plant design in order to generate estimates of gaseous effluent releases. These estimates were used to demonstrate compliance with effluent release regulations as part of the original licensing basis. Adherence to the gaseous waste effluent requirements is monitored by procedures in accordance with the Offsite Dose Calculation Manual (ODCM).

Monitoring gaseous effluents in accordance with the ODCM ensures that the composite results of

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-3 the variations in gaseous waste inputs and processing on the actual releases are within the accepted current licensing basis for the gaseous waste disposal system as specified in the acceptance criteria of the ODCM.

An analysis of the estimated curies of each radionuclide released from the station gaseous waste disposal system via the waste gas decay tanks has been made to demonstrate that 10 CFR 20 will be met. This analysis is presented in Section 11.2.5, and, as can be seen from Tables 11.2-2 and 11.2-3, the yearly dose at the site boundary with the recombiner operating is about 0.007 rem and is about 0.019 rem without the recombiner. These values are 1.4% and 3.8%

(with and without the recombiner, respectively) of the unrestricted area dose of 0.5 rem specified in the revision of 10 CFR 20 to which the plant was originally licensed.

11.2.3 Liquid Waste Disposal System The liquid waste disposal system for Units 1 and 2 is shared, except for the primary drain transfer tanks and the gaseous drain system in each containment. Two systems currently exist for treating liquid wastes. These are the boron recovery system and the liquid waste disposal system.

The boron recovery system, which is described in detail in Section 9.2, treats effluents collected in the primary drain tank from the vents and drains system, as well as letdown from the primary coolant that is diverted from the chemical and volume control system (CVCS). The liquid waste disposal system treats the liquid wastes originating from containment, auxiliary building, fuel building, safeguards, radwaste facility and decontamination building sumps, and from laboratory drains. Steam generator blowdown may be transferred via the component cooling heat exchanger pit sump to the liquid waste disposal system as discussed in Section 11.2.3.2.2. Liquid waste originating from the containment, auxiliary building, fuel building, safeguards, component cooling water heat exchanger, and decontamination sumps and from the laboratory drains are collected in either the low-level waste drain tank or the high-level waste drain tank depending on the valve lineup in the primary vent and drain system. Liquid wastes are normally transferred to one of the two 30,000-gallon RF liquid waste collection tanks for processing through the RF evaporator system or are processed in the RF liquid waste reverse osmosis and demineralizer system. Liquid waste originating in the radwaste facility itself is collected in sumps and pumped directly to the RF liquid waste collection tanks or through filters to the laundry drain monitor tanks. An additional 60,000 gallons of transfer/storage capacity is available via the RF liquid waste surge tanks. The RF evaporator system is available for use during high liquid waste generation periods or as a backup to the RF liquid waste reverse osmosis and demineralizer system. Processed liquid waste is sent to the liquid waste monitor tanks. An inclined plate suspended solids/oil separator is in-line for either the evaporation or reverse osmosis and demineralization process options.

Laundry waste and personnel decontamination showers and sink wastes are collected in the contaminated waste drain tanks. These tanks are pumped to the RF for processing. Some large particulate liquid wastes originating in the radwaste facility are collected in the sumps and pumped to the laundry waste system for processing. Waste is processed with a laundry pre-filter

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-4 and the laundry waste filter, then collected in the laundry drains monitor tank. Processed laundry waste and RF waste can be sampled and released or mixed with other station liquid waste.

Liquid and laundry wastes discharged to the circulating water system via the RF are monitored. The liquid effluent radiation monitor is an on-line monitor with automatic isolation of the effluent discharge when a high radiation alarm is received.

The Steam Generator Storage Facility sump is periodically pumped out and processed via the Laundry Drain System if contaminated.

Table 11.2-4 presents information regarding the originally licensed Surry liquid waste treatment system. This information provides parameters used as input into the calculation of radiation exposure to the public presented in Appendix 11A. With respect to the present Radwaste Facility, the information in Table 11.2-4 is conservative when compared to the RF radwaste volumes, DFs, and hold-up capacity. Therefore, the original evaluation of radiation exposure to the public presented in Appendix 11A bounds the design of the RF. Reference Drawings 1 through 3 and Figures 11.2-1, 11.2-2, 11.2-3, and 11.2-4 depict the liquid and laundry waste systems in the RF and their tie-ins to the station collection points.

11.2.3.1 Components 11.2.3.1.1 High-Level Waste Drain Tanks Two high-level waste drain tanks are provided. Each tank has a usable capacity of approximately 2000 gallons. Level indicators are provided. These are stainless steel tanks designed according to Section III.C of the ASME Boiler and Pressure Vessel Code.

11.2.3.1.2 Low-Level Waste Drain Tanks Two low-level waste drain tanks are provided. Each tank has a usable capacity of approximately 1785 gallons. Level indicators are provided. These are stainless steel tanks designed according to Section III.C of the ASME Code.

11.2.3.1.3 Contaminated Drain Tanks Two contaminated drain tanks are provided. Each tank has a usable capacity of approximately 1045 gallons. Level indications are provided. These are stainless steel tanks designed to Section VIII of the ASME Boiler and Pressure Vessel Code.

11.2.3.1.4 Waste Disposal Evaporator and Auxiliaries (Installed But No Longer Used)

One forced-circulation evaporator with a feed capacity of 6 gpm is provided. The evaporator shell is fabricated from a high-nickel alloy in accordance with Section III.C of the ASME Code. Internals are fabricated from an austenitic stainless steel not susceptible to stress cracking.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-5 The external heat source is a shell and tube steam reboiler fabricated on the tube side from a high-nickel alloy and on the shell side from carbon steel. Distillate is condensed in a water-cooled shell and tube condenser fabricated from austenitic stainless steel. The reboiler, shell, and tube condenser are all fabricated in accordance with Section III.C of the ASME Code, and TEMA Standards. (The external heat source steam lines have been cut and capped to preclude steam and/or water leakage.)

The condensed distillate is held in the distillate accumulator. This tank is fabricated from austenitic stainless steel in accordance with Section III.C of the ASME Code.

A distillate cooler is provided to further cool the distillate. The tube side of the distillate cooler is fabricated from austenitic stainless steel and the shell side from carbon steel, in accordance with Section III.C of the ASME Code.

11.2.3.1.5 Waste Disposal Evaporator Test Tanks (Installed but no longer used)

Two waste disposal evaporator test tanks, each of 3000-gallon capacity, with level indicators, are provided. These tanks are stainless steel and designed according to Section VIII of the ASME Code.

11.2.3.1.6 Pumps Centrifugal frame-mounted pumps with single or double mechanical seals are provided.

The waste disposal evaporator bottoms pump is a canned pump. One pump is provided for each tank with cross ties where appropriate, such as on high-level waste drain tank pumps. External cooling and seal water is supplied to radioactive pump seals as required.

11.2.3.1.7 RF Liquid Waste Evaporator System The evaporator system consists of a 30-gpm forced circulation evaporator system. The evaporator is designed to concentrate waste up to a boron concentration of 24,500 +/-5% ppm or the total solids concentration of 25 weight percent.

Normal feed to the evaporator is from a liquid waste transfer pump after the SPI oil/SS remover. Clean liquid effluent from the evaporator is transferred to the evaporator distillate demineralizer for further processing prior to transfer to the liquid waste monitoring tanks. The evaporator concentrates are forwarded to the bitumen solidification system for volume reduction, solidification, and packaging, or stored for later shipment in liquid form.

The evaporator is a forced circulation system utilizing a mechanical vapor recompression (MVR) system. The MVR evaporator operates on a heat pump principle. The process vapors are compressed to a higher pressure so they will condense at a higher temperature. The hot compressed vapors condense in the heater. The liberated heat causes boiling in the evaporator vapor body. Desuperheating water is added to the vapors to recover the superheat as sensible heat.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-6 The liquor entering the vapor body flash boils to release heat in the form of water vapor. As the water is driven from the system in to vapor phase, the liquor contained in the vapor body is further concentrated. When the boron concentration reaches 24,500 +/-5% ppm or the total solid concentration of 25 weight percent, a portion is removed by gravity to the Evaporator Bottoms Tank. The amount of concentrates removed is replaced by an increase in the feed rate, thus the solids concentration in the evaporator is decreased again.

The vapor leaving the vapor body passes through an entrainment separator. In the separator, the vapor flows upward through percolated trays and mesh pads. These remove droplets entrained in the vapor to protect the compressor wheel from erosion and to insure clean condensate.

The vapor from the separator is compressed by a high speed centrifugal compressor. The compressed vapors exit the compressor with a large amount of superheat, which is desuperheated to bring the vapor temperature close to saturation temperature and recover the superheat as sensible heat.

The vapor from the compressor is condensed on the shell side of the heater. The distillate flows by gravity to the distillate flash tank, where it is first flashed to the entrainment separator to recover heat. Then it is pumped through a distillate subcooler to the distillate demineralizer.

Concentrated waste from the evaporator is periodically discharged to the evaporator bottoms tank. The bottoms tank vent is connected to the tank vent system after passing through a vent cooler.

The major components of this subsystem are the vapor body, the heater, the recirculation pump, the entrainment separator, the vent gas cooler, a motor driven vapor compressor, a bottoms tank and bottoms tank pump. Design information on these components are given in Table 11.2-1.

11.2.3.1.8 RF Liquid Waste Reverse Osmosis and Demineralizer System The RF liquid waste reverse osmosis (RO) and demineralizer system is designed to remove radioactivity and dissolved solids from the liquid waste process prior to collection in the liquid waste monitor tanks where liquids are sampled and discharged or reused. The RF liquid waste reverse osmosis and demineralizer system is normally in service to process liquid waste streams.

The RF liquid waste reverse osmosis and demineralizer system is designed to remove total suspended solids to < 25 ppm, and oil and grease to < 15 ppm prior to entering the liquid waste monitor tanks.

The system consists of demineralizer vessels and a Thermex RO unit. The RO concentrates are recirculated while the permeate is directed to the liquid waste monitoring tanks. The content of the process feed tank is directed to a collection tank when concentrate limits have been met.

The demineralizer vessels are designed and constructed per ASME VIII. The reverse osmosis skid was designed per ANSI B31.1.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-7 11.2.3.1.9 Postaccident Radiation Waste Connection The capability for processing highly radioactive postaccident liquid waste has been incorporated into the liquid waste disposal system. A flanged connection (Reference Drawing 1) is located in the vicinity of the boron recovery tanks for the purpose of discharging radioactive liquids to an external process system without requiring personnel to enter high radiation areas.

The external process system would be brought onsite, if needed, following an accident. The postaccident radiation waste connection also has an isolation valve that can be operated by reach rod to further minimize personnel exposure.

11.2.3.1.10 RF Laundry Waste System The RF laundry waste system receives waste from the contaminated drain tanks and from the RF building drain system sump pumps via a cross-connect line. The RF laundry systems consist of laundry drain prefilters and the main laundry drain filter.

The two laundry drain prefilters are installed in parallel with one operational to remove large solid matter such as cotton fibers to extend the service cycle of the downstream laundry drain filter. The filters are designed to operate at 50 gpm. The filtration media element is a bag type constructed of artificial fiber cloth. The filter media will be periodically removed manually and transferred to the dry activated waste (DAW) area for drying and volume reduction at the Radwaste Facility or packaged for offsite processing. The filter housing is constructed of 304 stainless steel.

The laundry drain filter is designed to remove particulate matter from the incoming laundry drain stream. The filter is designed to operate at a rate of 50 gpm. It is designed to remove suspended solids with a filtration media of polyethylene fiber balls. The expended filter media will be removed manually and transferred to the DAW compaction area for drying and volume reduction at the Radwaste Facility or packaged for offsite processing.

The filter vessel is designed and constructed in accordance with ASME VIII. The principal material of construction is 304 stainless steel.

Laundry waste and RF building drain system waste processed by this system is sent to the laundry waste monitor tanks for sampling prior to monitored discharge.

11.2.3.1.11 RF Monitor Tanks The RF liquid waste evaporator, reverse osmosis, and/or demineralization process systems send their processed waste to one of two liquid waste monitor tanks. These tanks are 15,000 gallons, vertical tanks. Each tank has level indication and recirculation capability. They are constructed of 304 stainless steel and have an atmospheric pressure design. The tanks are designed and constructed to ASME Section III.

Laundry waste processed through the RF is sent to one of two laundry waste monitor tanks.

These tanks are 7500 gallons, vertical tanks. Each tank has level indication and recirculation

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-8 capability. They are constructed of 304 stainless steel and have an atmospheric design. The tanks are designed and constructed to ASME Section III.

11.2.3.2 Processing Steam Generator Blowdown The steam generator blowdown system is described in Section 10.3.1.2. A review of the effects of the power uprate to a core power of 2546 MWt was conducted and the steam generator blowdown system was found to be adequate.

A steam generator blowdown treatment system, located in the Condensate Polisher Building, was utilized to remove impurities from the blowdown stream. The system contained prefilters, demineralizers, and postfilters. The blowdown treatment system was designed to be used during normal operation and following steam generator tube leakage. Subsequently, the system was determined to be incompatible with changes made in secondary water chemistry. As a result, the blowdown treatment system is no longer used for blowdown treatment and blowdown is untreated, except as described below.

11.2.3.2.1 Normal Operation During startup and power operations, blowdown is either released to the discharge canal or returned to the condenser hotwell as described in Section 10.3.1.2. During outages, the steam generators may be gravity drained through the blowdown lines to a waste neutralization sump located in the condensate polishing building. Water in the waste neutralization sump can be treated, recirculated, sampled and discharged to either the settling pond or the circulating water discharge. During discharge, the water may be directed through a filter or the filter can be bypassed.

11.2.3.2.2 Operation Following Steam Generator Tube Leak If a steam generator tube leak occurs and shutdown is desired, station procedures provide guidance on evaluating contamination potential and determining appropriate actions for processing blowdown. Depending on activity levels, blowdown may be directed to the condenser hotwell or may need to be processed through the Surry Radwase Facility (SRF). If it is decided to process the blowdown through the SRF, procedures direct that the inventory from the affected steam generator be transferred to the component cooling (CC) heat exchanger (HX) pit sump using blowdown hose connections. Steam generator pressure will provide the motive force for transfer from the affected steam generator to the sump, but gravity transfer is possible. The CC HX pit sump pump transfers water in the sump to the combined containment and safeguard area sump pump discharge header where it can be processed by the liquid waste disposal system (Section 11.2.3). The flow rate into the sump from the affected steam generator will be limited to less than 25 gpm so that it will not exceed the pumping capacity of the CC HX pit sump pump.

The flow rate can be controlled from the main control room by use of an HCV in the blowdown line.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-9 Controls are in place to minimize the airborne activity levels in the vicinity of the CC HX pit. Hose connections will be monitored for leakage. Water entering the sump from the blowdown lines will be cooled by the steam generator blowdown coolers to a subcooled condition to preclude flashing. Flow rates into the CC HX pit sump will be limited to below the capacity of the CC HX pit sump pump so that the water level remains in the sump and does not enter the pit. This minimizes the liquid surface area and would allow for an exhaust hood to be installed over the sump if needed to reduce airborne activity levels. Health Physics will monitor the radiation levels in the areas of the routed hose and the radiation and airborne levels around the CC HX pit.

11.2.4 Solid Waste Disposal System The solid waste disposal system provides logging, packaging, and storage facilities for scheduled shipment off the site and ultimate disposal of radioactive waste material. Materials handled as solid waste include concentrated liquid sludge, water, spent resin, spent filter cartridges, solid noncompactible and compactible trash, and other miscellaneous materials resulting from station and RF operation and maintenance. The operation of the system is described below.

11.2.4.1 Solid Waste Handling Operations 11.2.4.1.1 Expended Filter-Cartridge Handling Operations Radioactive liquid service filters are removed from the system when the pressure drop across the filters becomes excessive or when the radiation level exceeds a predetermined maximum. The filter housing is surveyed prior to any opening or removal. After this is completed, the filter cover is remotely opened and removed by personnel using appropriate tools and protected by a filter removal shield, when required. A lead cask is placed over the filter housing and the filter is then moved upward into the lead cask. A drip pan is then secured to the bottom of the cask and the entire assembly is transported to the Surry Radioactive Waste Facility (SRF). The filter is placed in an approved container. The filter and container remain in the SRF until shipment to the burial site.

Surveys are conducted on the filter when the filter housing is opened, in the shielded casks, and when the container is transported to the Radwaste Facility or removed for shipment. The transport vehicle undergoes a complete radiological survey before leaving the site.

11.2.4.1.2 Spent Resin Handling Operations A spent resin catch tank and spent resin blend tank are provided to receive spent resin from the stations ion exchangers located in the Auxiliary Building. A transfer pump is associated with each tank. Spent resin is transferred from the blend tank to a high-integrity container for shipment to a burial facility. A shipping container may be sent to the Radwaste Facility for staging prior to shipment offsite.

Primary plant resins are directed to the spent resin catch tank. Resins from the catch tank are slurried to the blend tank to produce a mixture of resin whose contents may be shipped in a high

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-10 integrity container. As an option, resin from the spent resin catch tank can be sluiced to a mobil resin transfer vessel (MRTV) for shielded transport to the Radwaste Facility. Subsequent to the processing operations, the lines are flushed with primary grade water.

Spent low-activity resins from the condensate polishing system are typically dewatered to acceptable strong tight containers and sent offsite for disposal. If the activity in these resins becomes high enough that the disposal site would not accept them in drums, the resins would be slurried to high integrity containers (HICs) and then dewatered prior to shipment to an offsite radwaste burial site. There is also an option to transport the spent condensate polishing resins to the Radwaste Facility. From the Radwaste Facility, condensate polishing resins can be sent to the bitumen solidification system, sent to a high integrity container filling and dewatering station, or proportionately blended with other resins for the purpose of lowering the dose rate of some higher activity resins.

11.2.4.1.3 Evaporator Concentrate Operations Solids that are concentrated in the evaporator are discharged to the evaporator bottoms tank.

These concentrates are at 25% by weight solids or at 24,500 +/- 5% ppm boron. Additionally, sludge from the suspended solids separator is periodically pumped to the bottoms tank.

The concentrates and sludges in the bottoms tank may be pumped to the waste batch tanks through heat traced lines where the concentrate waste is pretreated for processing by the solidification system.

In order to preclude plugging of the concentrates transfer piping, redundant heat tracing circuits are installed. Clean, hot water flushing connections are included to clean each line following concentrates transfer.

11.2.4.1.4 Solidification Operations (Installed but no longer used)

The bitumen solidification system incorporates a chemical and physical process for reducing the volume of radwaste and for incorporating the radwaste into a solidified bitumen matrix. The process uses a LUWA thin-film evaporator operating at a waste product outlet temperature of approximately 320°F. This results in the evaporation of free water from waste effluents and the remaining solids are incorporated in a bitumen matrix. Solidification of the end product occurs upon the natural cooling of the binder.

The system is capable of processing waste which includes evaporator concentrates and spent bead resin. Waste to be processed is collected in one of two waste batch tanks. The waste is sampled and chemically pretreated to prepare it for processing.

The conditioned waste is fed at a controlled rate to a thin film evaporator. Molten bitumen is simultaneously fed into the evaporator through a second feed nozzle. The evaporator is heated by means of hot thermal fluid flowing through an external jacket. As the waste flows downward through the evaporator, the water is evaporated and the water vapor flows counter-currently

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-11 upward and out of the evaporator. The waste solids are mixed with molten bitumen and exit the bottom of the evaporator, flowing into a waste container. Upon cooling, the waste/bitumen mixture solidifies into a freestanding, monolithic solid with free liquids less than 0.5 percent by volume of the waste form.

The water vapor leaving the thin film evaporator is condensed in a shell and tube condenser.

The condensate flows into the distillate oil separator. When this tank is filled, the distillate is pumped to a liquid waste collection tank.

The bituminized waste product flows from the discharge valve of the thin film evaporator into a 55-gallon steel drum. Once filled and cooled, drums are inspected for free liquid, capped, smeared and surveyed. The drums are then transferred to the RF storage area to await shipment to a licensed disposal contractor.

Figure 11.2-5 depicts the solidification process flow.

11.2.4.1.5 Ultimate Disposal Operations All packages containing radioactive nonfissionable material, and the procedures used to prepare these for offsite shipment, are in accordance with U. S. Department of Transportation regulations. The Radwaste Facility, Low-Level Waste Storage Facility and Sea Van Storage Pad are facilities used for the storage of radioactive material. All waste material is transferred either to a licensed disposal or processing contractor or to common carrier for delivery to a licensed disposal or processing contractor. Radwaste shipments fall under the purview of Vepcos procedures and quality assurance program.

11.2.4.2 Components All components listed below except the spent resin catch tank, the spent resin blend tank, and their associated pumps are located within the Radwaste Facility.

11.2.4.2.1 Spent Resin Catch Tank Spent Resin Blend Tank One of each tank is provided. Each tank is installed in a separate cubicle on Elevation 6 ft.

10 in. in the Decontamination Building. The normal operating volume (high level to low level) is approximately 214 ft3. Total usable volume is approximately 245 ft3. Vessels are designed to ASME Section VIII.

11.2.4.2.2 Spent Resin Catch Tank Transfer Pump Spent Resin Blend Tank Transfer Pump One of each pump is provided. Pumps are of the progressive cavity design. Each pump is designed to deliver 27.7 gal/100 rpm at 0 psi.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-12 11.2.4.2.3 Evaporator Bottoms Tank The evaporator bottoms tank is a 5000-gallon, 11-foot diameter, vertical tank with a dished bottom and a flat top. The tank is constructed of Inconel 625 and is equipped with a mixing eductor and a demineralizer water flush header, level indication, and heat tracing. The heat tracing prevents concentrates from solidifying in the tank.

11.2.4.2.4 Spent Resin Collection Tanks There are four spent resin collection tanks in the Radwaste Facility. These tanks are 1020-ft3 capacity, vertical tanks with a 10-foot diameter. Each tank is constructed of 304 stainless steel and is designed for atmospheric pressure. Each tank is equipped with mixing lines, flush lines, decant lines and an overflow. Each of the two mixing lines is attached to internal mixing eductors.

11.2.4.2.5 Spent Resin Collection Tank Pumps There are two spent resin collection tank pumps each of which is capable of pumping from any of the four spent resin collection tanks. Each pump is rated at 440 gpm at a total dynamic head of 86 psi. Parts of the pumps in contact with the radioactive resin/water slurry are made of stainless steel.

11.2.4.2.6 Waste Batch Tanks There are two 1000-gallon waste batch tanks. Each tank is constructed of 316L stainless steel and is equipped with external heating elements to prevent the solidification of evaporator concentrates. Each tank has mechanical agitators for mixing.

11.2.4.2.7 Bitumen Storage Tank The bitumen storage tank is a horizontal 6000-gallon carbon steel tank. The tank has an internal electric heater and is heavily insulated. The tank is equipped with instrumentation for tank level and temperature.

11.2.4.2.8 Bitumen Metering and Transfer Bitumen is metered by a gear type metering pump capable of accurate control of the bitumen feed between 0.04-0.92 gpm. The piping from the storage tank to the thin film evaporator are not heat traced. Instead, transfer lines are jacketed pipes using the heating oil from the thin film evaporator heating system to maintain flow of the bitumen feed.

11.2.4.2.9 Thin Film Evaporator The thin film evaporator is a LUWA design capable of a 52-gal/hr evaporation rate. The body of the evaporator is made of 316L stainless steel. The internal paddle assembly in the evaporator continuously spreads the feed material into a thin film along the vessel walls to assist in the evaporation. The constant action of the paddle assembly also assures adequate and uniform

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-13 mixing of the waste and the bitumen binder. The paddle assembly is turned by a 15 hp electric motor.

11.2.4.2.10 Distillate Oil Separator Tank Water vapor from the thin film evaporator is condensed and collected in the distillate oil separation tank. Due to the use of bitumen in this process, small amounts of light weight oils are volatilized during the evaporator process. These oils are separated in the distillate oil separation tank and are skimmed off for separate treatment. The condensed water in the tank sent to the liquid waste collection tanks via the distillate transfer pump.

11.2.5 Gaseous Waste Disposal System The process vent subsystem regulates the discharge of potentially high-activity waste gases to the atmosphere. The ventilation vent subsystem described in Section 9.13 and the RF vent described in 11.2.2 regulate the discharge of potentially low-activity air streams to the atmosphere. Radioactive waste discharges from these subsystems are monitored by particulate and gas monitors that are part of the process radiation monitoring system described in Section 11.3.3. Limitations on gaseous releases, and associated reporting requirements, are included in the Technical Specifications and the Offsite Dose Calculation Manual.

Waste gases, primarily hydrogen, nitrogen, and minor amounts of fission product gases, such as xenon and krypton, are removed from reactor coolant letdown by the stripper in the boron recovery system. The stripped gases are processed in the gaseous waste disposal system.

The gaseous waste disposal system is designed to provide adequate radioactive decay storage time for the waste gases and, in addition, provide long-term holdup of these gases when high-flow letdown is required.

Gases pass from the stripper to the stripper surge tank, where they are compressed. From the surge tank, the gases are bled off to the waste gas surge drum. At a pressure of approximately 1 atm, the waste gas diaphragm compressor transfers the gases to one of two waste gas decay tanks.

When released, effluent from the waste gas decay tanks is mixed with dilution air, effluent from the containment vacuum system, and the aerated vents from the vent and drain system. The combined gaseous waste is filtered through charcoal and high-efficiency particulate air (HEPA) filters before being released to the atmosphere. The process vent blowers maintain a small vacuum in the charcoal filters to prevent outleakage from the filter assembly. The decay tank contents are sampled before any release to the process vent.

11.2.5.1 Process Vent Subsystem Gaseous wastes enter the process vent subsystem from the gaseous waste disposal system, the stripper in the boron recovery system, the vent and drain system, various pressure relief valves, and the containment vacuum system, as shown in Reference Drawings 4 and 5.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-14 A catalytic recombiner system is installed (but not used) as part of the gaseous waste disposal system. A summary description of the catalytic recombiner is provided in Section 11.2.5.3.1.

Two double-walled waste gas decay tanks are provided. Each tank is buried beneath the Waste Gas & Boron Recovery Pump House and receives tornado protection from this building within its footprint. Where the tanks extend outside the footprint of this building, tornado protection is provided by the depth of burial in soil. The inner tank is fabricated from austenitic stainless steel in accordance with Section III.C of the ASME Code, and the outer tank from carbon steel, in accordance with Section VIII of the ASME Code. Sampling connections are provided for the tank contents and for leakoff in the annular intercept space between the tanks.

The decay tanks have piping connections for parallel operation with alternate feed and bleed.

Overpressure relief protection is provided at the waste gas decay tanks in accordance with Section III.C of the ASME Code. The protective devices consist of bellows-sealed pressure relief valves followed by rupture disk assemblies. The use of bellow seals and rupture disks precludes the leakage of the waste gas to the environment during normal operation of the gaseous waste disposal system. The piping downstream of the protective devices relieves to the process vent through the radiation monitor station.

Effluent from the waste gas decay tanks is mixed with dilution air, effluent from the containment vacuum system, and the aerated vents from the vent and drain system. The combined gaseous waste is filtered through charcoal filters before being released to the atmosphere. The process vent blowers maintain a small vacuum in the charcoal filters to prevent outleakage from the filter assembly. The decay tank contents are sampled before any release to the process vent.

The entire discharge stream of radioactive letdown gas and dilution air is monitored for flow rate, pressure, temperature, and particulate and gaseous activity before release through the process vents. The total flow is regulated by a flow control valve on the process vent blower. The ratio of dilution air to waste gas letdown flow is such that the mixed streams never enter the flammability region of the air-steam-hydrogen phase diagram.

The process vent and the process vent blowers are sized such that the minimum exit velocity is approximately 100 fps. This exit velocity prevents any significant downdrafting of the effluent. The process vent terminates at an elevation approximately 22 feet above the top of one of the containment structures.

The process vent monitors are set such that the effluent activity release rate results in concentrations less than those limits provided in the revision of 10 CFR 20 to which the plant was originally licensed. In the event that the activity of the effluent stream exceeds the setting of the monitors, the process vent control station automatically terminates the release of waste effluents from the waste gas decay tanks and isolates the containment vacuum system from the process vent sub system. The monitor also activates an alarm in the control room before valve closure if the activity approaches a preset value. Subsequent restart of the system is manual, in accordance

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-15 with procedures. The discharge of gases from the waste gas decay tanks is initiated and controlled separately.

The gaseous waste disposal system is designed to provide adequate radioactive decay storage time for the waste gases and, in addition, to provide long-term holdup of these gases when high-flow letdown is required.

The combined volume of the two waste gas decay tanks is sized to process the gas stripped from the estimated annual average letdown flow of 17 gpm, based on simultaneous operation of two reactor units.

The average gas stripping rate is a function of the average letdown flow rate, and this flow rate is dependent on the assumed plan of operation as described in Section 9.2.

On this basis, the total annual letdown volume for two units is 8.94 x 106 gal, and the average annual letdown flow rate for two units is 17 gpm.

If the hydrogen volume is assumed to be 35 cm3/kg and 90% of the total gas volume, then the hydrogen stripping rate at an annual average of 17-gpm letdown is 0.0792 scfm and the total gas stripping rate is 0.088 scfm.

The gas decay tanks are sized so that a 17-gpm letdown rate with the recombiner not operating gives an average holdup time equivalent to approximately 5 half-lives of Xe-133 (30 days).

Assuming 1% failed fuel, the estimated curies of each radionuclide released from the station via the gaseous waste disposal system for the original plant design are listed in Tables 11.2-2 and 11.2-3. Table 11.2-2 is for a waste gas cycle with the recombiner not operating, which is the design basis for the system, and Table 11.2-3 is for a waste gas cycle with the recombiner operating. With the recombiner not operating, the gas cycle is 30 days of feed/20 days of decay/10 days of bleed; most of the gas is hydrogen. With the recombiner operating, the feed portion of the waste gas cycle can vary between 30 days and approximately 300 days, the time needed to reach maximum design pressure in the tank. To be conservative, 300 days of feed was chosen as the basis for Table 11.2-3, and the waste gas cycle considered was 300 days of feed/20 days of decay/10 days of bleed.

In each case, it is assumed that all of the gases and 0.1% of the iodines are removed at the gas stripper and sent to the waste gas decay tanks except for hydrogen in the case with the recombiner operating. The system is operated so that one tank is on the feed portion of the cycle while the other is on the decay and bleed portion.

The equilibrium reactor coolant activity is a function of the waste gas removal rate by the gas stripper. Using the parameters listed in Table 9.1-5 and a 17-gpm letdown rate to the gas stripper, the equilibrium coolant activity for each radionuclide was calculated. These are also listed in Tables 11.2-2 and 11.2-3.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-16 As can be seen from these tables, the yearly dose at the site boundary is about 0.019 rem for continuous operation with a 30 days of feed/20 days of decay/10 days of bleed cycle, and about 0.007 rem for the 300 days of feed/20 days of decay/10 days of bleed cycle. Both of these values are well below the member of the public dose limit of 0.5 rem/yr set forth in the revision of 10 CFR 20 to which the plant was originally licensed.

11.2.5.2 Ventilation Vent Subsystem The ventilation vent subsystem is considered to be a portion of the gaseous waste disposal system only for purposes of radiological surveillance, and it is designed on this basis. However, since it handles air streams of very low activity levels, and since the gases to be handled are predominantly of nonradioactive origin, this subsystem has been considered as an auxiliary system for the purpose of this report. A full description of this subsystem is included in Section 9.13.

11.2.5.3 Components The major components of the gaseous waste disposal system are described below.

11.2.5.3.1 Catalytic Recombiner (Installed But Not Usable)

One skid-mounted catalytic recombiner system is provided. The system includes duplicate full-capacity recycle compressors, duplicate full-capacity electrical preheaters, duplicate full-capacity catalytic recombiners, one aftercooler condenser, one moisture separator, one electrical reheater, duplicate hydrogen analyzers of the thermal conductivity type for the recombiner influent and effluent, duplicate oxygen analyzers of the paramagnetic type on the recombiner effluent, a single oxygen analyzer on the recombiner influent, and one bleed stream cooler. The recombiner system operates at approximately 22 psia and has a feed capacity of approximately 1.14 scfm. The diluent is nitrogen. The catalytic recombiner system is designed according to Section III.C of the ASME Code.

The recycle compressors are rotary positive blowers designed to circulate 40 cfm at 8 psig discharge pressure. They are of gas-tight construction, with three mechanical shaft seals in series between the circulating gas and the outside atmosphere. The end bell of the compressor is pressurized with nitrogen as a further precaution against outward leakage.

The preheaters are stainless steel pipes with external electrical heating elements and are used to raise the temperature of the recycle stream to 300°F before the recycle stream enters the catalyst bed.

The catalytic recombiners are all-metal, low-halogen catalysts. Each bed contains miles of crimped-nickel alloy ribbon coated with catalytically activated precious metals, mainly palladium and platinum.

The aftercooler condenser is a pipe-within-a-pipe heat exchanger, with the recycle gas flowing through the inner pipe and component cooling water flowing through the outer pipe. The

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-17 aftercooler condenser condenses the water vapor by cooling the recycle gas stream and lowering the water vapor in it to a dewpoint of 75°F.

The moisture separator is a centrifugal separator with an automatic drain operated by a level controller and has high- and low-level alarms. Moisture from the moisture separator drains to a high-level waste drain tank.

The reheater is similar to the preheaters, and raises the temperature of the recycle stream to 120°F.

The hydrogen analyzers are of the thermal conductivity type; the oxygen analyzers are of the paramagnetic type.

11.2.5.3.2 Waste Gas Surge Tank One waste gas surge tank with a 15.7 ft3 capacity is provided. This tank is operated at a pressure of approximately 10 to 20 psia. The tank is fabricated from austenitic stainless steel in accordance with Section III.C of the ASME Code.

11.2.5.3.3 Waste Gas Compressor Two waste gas compressors of the diaphragm type are provided. Each has a rated capacity of 1.5 scfm at a discharge pressure of 120 psig. The compressor heads are leak tested to ensure that the leakage does not exceed a predetermined amount.

11.2.5.3.4 Waste Gas Decay Tank Two buried waste gas decay tanks are provided. These tanks have double-wall construction with feed and bleed lines; sample, nitrogen purge, drain, and relief valve lines to the inner tank; and sample, nitrogen purge, drain, and relief valve lines from the outer tank. An access opening is provided to the inner tank. In addition, adequate grounding and corrosion protection are provided.

The inner tank is fabricated from stainless steel in accordance with Section III.C of the ASME Code, and the outer tank from carbon steel in accordance with Section VIII of the ASME Code.

To ensure an explosive gas mixture does not develop in the tanks, samples are taken via the oxygen analyzer. Compressors have been installed as sample pumps but are normally isolated and bypassed. The differential pressure between the waste gas decay tank and the waste gas surge drum is used to induce flow through the oxygen analyzer.

11.2.5.3.5 Process Vent Blowers Two full-capacity dilution air blowers of 300 cfm capacity at 2 psia are provided. The blowers are of a centrifugal type, located in a field fabricated box with the blower suction from the boxs interior. Some inleakage is tolerated.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-18 11.2.5.3.6 Charcoal Filters Two charcoal filter beds are provided to service approximately 300-scfm radioactive gas.

The filters are maintained at a subatmospheric pressure.

11.2.5.3.7 Postaccident Radiation Waste Connection The capability for processing highly radioactive postaccident gaseous waste has been incorporated into the gaseous waste disposal system. A flanged connection (Reference Drawings 4 & 5) is located in the vicinity of the boron recovery tanks for the purpose of discharging radioactive gases to an external process system without requiring personnel to enter high radiation areas. The external process system would be brought onsite, if needed, following an accident. The postaccident radiation waste connection also has an isolation valve that can be operated by reach rod to further minimize personnel exposure.

11.2.6 Tests and Inspections 11.2.6.1 Construction and Fabrication During the manufacturing period, Vepcos inspectors inspected all equipment periodically, as required, to ensure that all equipment had been provided in strict accordance with specifications. Shop hydrostatic and performance tests of principal equipment were witnessed by Vepcos inspectors. Certified code inspection data sheets were provided by manufacturers for all equipment covered by ASME or other applicable codes.

During the construction period, all pressure systems were subjected to field hydrostatic or pneumatic tests to verify the integrity of welded connections and to ensure that the system as a whole functioned as intended.

During the preliminary operation period, all equipment in the waste disposal system was tested to verify conformance with specification performance requirements. All control systems and interlocks were tested and operated to ensure satisfactory functional performance and reliability.

The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

On February 9 and 10, 1971, a performance test of the waste disposal system catalytic recombiner was conducted to demonstrate the actual catalyst performance over a range of operating conditions. Tests were rerun at three hydrogen inlet concentrations: test 1, 2.0%

hydrogen; test 2, 0.5% hydrogen; and test 3, 3.2% hydrogen.

The specified test requirements at each test condition were to maintain the outlet oxygen concentration at no more than 1% and to have less than 100 ppm of hydrogen in the outlet.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-19 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

The catalyst efficiency was greater than 99.5% over the range of hydrogen flow rates and concentrations tested. The outlet hydrogen and oxygen concentrations were well within the stated limits. The tests confirmed the capability of the recombiner to operate over the range of operating conditions expected during normal operations.

11.2.6.2 Operation The basic function of the waste disposal system is to release controlled amounts of radioactivity to the environment with no undue effects on the health and safety of the general public. This is accomplished by ensuring that all releases from the station are at less than the maximum levels of radioactivity set by applicable regulatory agencies, as given in Section 11.2.1.

The following is a list of the types and areas monitored to ensure the proper functioning of the waste disposal system:

1. Continuous Process Monitoring: As described in Section 11.3.3, process radiation monitors continuously monitor certain key systems where radioactive material may exist. These monitors give an indication of the waste-processing requirements of certain systems.

2. Batch Sample Process Monitoring: As described in Section 9.6, batch samples, obtained from certain subsystems, provide information on the effectiveness of ion exchangers, filters, and evaporators. This gives an indication of the effectiveness of the various waste processing subsystems. The monitoring of the gas in the waste gas holdup tanks avoids the storage of excessive activity.

3. Continuous Monitoring of Discharge Effluents: As described in Section 11.3.3, radiation monitors continuously monitor discharges from the process and ventilation vent systems and RF systems and the liquid waste disposal and service water systems. These monitors give an indication of liquid and gaseous radiation discharges to the environment and provide alarms with automatic valve closure when radiation levels exceed a preset level, thus terminating discharge.

4. Radiation Survey of Radioactive Waste/Material Containers: Radiation surveys and smear samples are taken of shipping casks, drums, etc., that contain radioactive waste/material to ensure that such waste/material is properly contained and meets transportation regulations.

5. Environmental Monitoring: As described in Section 11.3.5, environmental samples are taken to indicate the effect of liquid and gas discharges on the environment and the compliance of these discharges with applicable regulations.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-20 To ensure that the performance of the waste disposal systems is meeting design criteria, the following checks are made:

1. Standardized laboratory radiochemical analytical procedures are used to verify decontamination factors.

2. Radiation monitors are periodically checked with remotely operated check sources. The Local Processing Units (LPUs) of the MGP Instruments (MGPI) monitors perform various self checks automatically. Their electrical self check introduces a known and fixed level of pulses into the electronics excluding the detector and verifies that the response is correct, otherwise a fault is generated. Additionally, the electronics continuously monitor the detector for a minimum countrate, otherwise a fault alarm is generated. In addition, samples are withdrawn from the process streams being monitored and analyzed to ensure compliance with regulatory limits. Those monitors that actuate control valves by a high radiation signal are periodically tested to ensure the valves activate on alarm signal.

3. Portable survey instruments and laboratory analytical instruments are periodically calibrated with known radiation sources in accordance with station health physics procedures.

4. Radiation levels on the outside of components, pumps, valves, and piping in the waste disposal systems are monitored periodically to avoid inadvertent discharge of activity that may accumulate with time.

11.2 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11448-FM-083C Flow/Valve Operating Numbers Diagram: Vent and Drain System, Unit 1

2. 11448-FM-30B Flow Diagram: Waste Disposal System

3. 11448-FM-30C Flow Diagram: Waste Disposal System

4. 11448-FM-090A Flow/Valve Operating Numbers Diagram: Gaseous Waste Disposal System, Unit 1

5. 11448-FM-090B Flow/Valve Operating Numbers Diagram: Gaseous Waste Disposal System, Unit 1

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-21 Table 11.2-1 WASTE DISPOSAL SYSTEMS DESIGN DATA Spent Resin Catch Tank (1-LW-TK-14)

Number 1 Capacity 1955 gal Operating capacity 214 ft3 Design pressure 150 psig Design temperature 200°F Operating pressure Atmospheric Operating temperature 60-120°F Material SS, SA-240, 316L Design code ASME VIII, D.V.

Spent Resin Blend Tank (1-LW-TK-15)

Number 1 Capacity 1955 gal Operating capacity 214 ft3 Design pressure 150 psig Design temperature 200°F Operating pressure Atmospheric Operating temperature 60-120°F Material SS, SA-240, 316L Design code ASME VIII, D.V.

High-level Waste Drain Tank Number 2 Capacity (each) 2390 gal Operating capacity (each) 2000 gal Design pressure 25 psig Design temperature 200°F Operating pressure Atmospheric Operating temperature 120°F Material SS Type 316 Design code ASME III.C Low-Level Waste Drain Tank Number 2 Capacity (each) 2874 gal Operating capacity (each) 1785 gal Design pressure 25 psig Design temperature 200°F Operating pressure Atmospheric Operating temperature 120°F

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-22 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Low-Level Waste Drain Tank (continued)

Material SS Type 316 Design code ASME III.C Contaminated Drains Collection Tanks Number 2 Capacity (each) 1230 gal Operating Capacity (each) 1045 gal Design pressure 25 psig Design temperature 200°F Operating pressure Atmospheric Operating temperature 120°F Material SS Type 304 Design code ASME VIII Liquid Waste Test Tanks (installed but no longer used)

Number 2 Capacity (each) 3000 (usable) gal Design pressure 25 psig Design temperature 212°F Operating pressure Atmospheric Operating temperature 140°F Material SS Type 304 Design code ASME VIII Waste Gas Catalytic Recombiner (installed but no longer used)

Number 1 Capacity, feed 1.31 scfm Design feed pressure 22 psia Design feed temperature 70-120°F Feed composition, scfm Max. Avg. Min.

H2 1.14 0.0805 0 H20 0.04 0.026 0 Xe Trace Kr Trace N2 0.130 0.00922 0 Design bleed pressure 14.0 psia Design H2 bleed concentration 0.1% max.

Design bleed volume 10% of feed Design code ASME III.C

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-23 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Waste Gas Decay Tanks Number 2 Capacity (each) 434 ft3 Design pressure Outer Tank Inner Tank From 30 in. Hg From 30 in. Hg vacuum to 150 psig vacuum to 150 psig Design temperature 200°F 200°F Operating pressure Atmospheric 115 psig Operating temperature 120°F 140°F Material Carbon SS Type 304L Design code ASME VIII ASME III.C Earthquake design Complies with Class I requirements Waste Gas Surge Tank Number 1 Capacity 15.7 ft3 Design pressure From 30 in. Hg vacuum to 30 psig Design temperature 300°F Operating pressure 15 psig Operating temperature 120°F Material SS Type 304 Design code ASME III.C Low-level Waste Drain Pumps Number 2 (1 required)

Type Horizontal centrifugal Motor horsepower 7.5 hp Seal type Double mechanical Capacity (each) 120 gpm Head at rated capacity 94 ft Design pressure 150 psig Materials Casing SS Type 316 Shaft A 5564, Type 630 Impeller SS Type 316 High-level Waste Drain Pumps Number 2 (1 required)

Type Horizontal centrifugal Motor horsepower 7.5 hp Seal type Double mechanical Capacity (each) 120 gpm

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-24 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA High-level Waste Drain Pumps (continued)

Head at rated capacity 86 ft Design pressure 150 psig Materials Pump casing SS Type 316 Shaft A564, Type 630 Impeller SS Type 316 Contaminated Drains Transfer Pumps Number 2 (1 required)

Type Horizontal centrifugal Motor horsepower 10 hp Seal type Mechanical Capacity (each) 75 gpm Head at rated capacity 166 ft Design pressure 150 psig Materials Pump casing SS Type 316 Shaft A564, Type 630 Impeller SS Type 316 Spent Resin Catch Tank Transfer Pump (1-LW-P-12)

Number 1 Type Progressive cavity Capacity 27.7 gal/100 rpm @ 0 psi Design pressure 100 psig Design temperature 150°F Material SS Type 316 Design Code None Spent Resin Blend Tank Transfer Pump (1-LW-P-13)

Number 1 Type Progressive cavity Capacity 27.7 gal/100 rpm @ 0 psi Design pressure 100 psig Design temperature 150°F Material SS Type 316 Design Code None

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-25 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Process Vent Blower Number 2 (1 required)

Type Multistage centrifugal Motor horsepower 7.5 hp Capacity (each) 300 scfm Differential pressure 2 psi Suction pressure 14.0 psia Design pressure 15 psig Materials Casing Cast iron Impeller Aluminum Shaft SS Type 304 Waste Gas Compressor Number 2 (1 required)

Type Diaphragm Motor horsepower 1.5 hp Capacity (each) 1.5 scfm Discharge pressure at rated capacity 120 psig Design pressure 220 psig Materials Cylinder Carbon steel Piston rod Forged steel Piston Nodular iron Diaphragms and parts contacting waste gas SS Types 304/316 Low-Level Waste Drain Filter (installed but no longer used)

Number 1 Retention size, microns 5 Filter element material Fibre Capacity normal 50 gpm Capacity maximum 75 gpm Material SS Type 304 Design pressure 150 psig Design temperature 250°F Design code ASME III.C High-level Waste Drain Filter (installed but no longer used)

Number 1 Retention size, microns 5 Filter element material Fibre Capacity normal 50 gpm

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-26 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA High-level Waste Drain Filter (installed but no longer used) (continued)

Capacity maximum 75 gpm Material SS Type 304 Design pressure 150 psig Design temperature 250°F Design code ASME III.C Contaminated Drains Filters (installed but no longer used)

Number 2 (1 required)

Filter element material Porous stone media Capacity normal 50 gpm Material SS Type 304 Design pressure 150 psig Design temperature 120°F Design code ASME VIII Liquid Waste Collection Tank Number 2 Capacity 30,000 gal Design pressure Atmospheric Plus Content Design temperature 150°F Operating pressure Full of water Operating temperature 104°F Material SS 316L Design code ASME III Liquid Waste Surge Tanks Number 2 Capacity 30,000 gal Design pressure Atmospheric Plus Content Design temperature 150°F Operating pressure Full of water Operating temperature 104°F Material SS 316L Design code ASME III Liquid Waste Collection Tank Pumps Number 2 Type Centrifugal Motor horsepower 15 hp Seal type Double Mechanical Capacity (each) 300 gpm

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-27 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Liquid Waste Collection Tank Pumps (continued)

Head at rated capacity 74 ft Design pressure 142.2 psig Materials Casing SS 316 Shaft SS 316 Impeller SS 316 Liquid Waste Surge Tank Pumps Number 2 Type Centrifugal Motor horsepower 15 hp Seal type Double Mechanical Capacity (each) 300 gpm Head at rated capacity 74 ft Design pressure 142.2 psig Materials Casing SS 316 Shaft SS 316 Impeller SS 316 SPI Suspended Solids/Oil Separator Number 2 Design capacity 30 gpm (each)

Design pressure Atmospheric Design temperature 150°F Material SS 316L Separation area, equiv. 8 ft2 (each)

Design code ASME III Liquid Waste Transfer Pumps Number 2 Type Centrifugal Motor horsepower 10 hp Seal type Double Mechanical Capacity (each) 40 gpm Heat at rated capacity 200 ft Design pressure 142.2 psig Materials Casing SS 316 Shaft SS 316 Impeller SS 316

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-28 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Liquid Waste Filter/liquid Waste Oil Filter Number 1 Retention size, microns N/A Filter element material Charcoal Capacity normal 60 gpm Capacity maximum 60 gpm Material SS 316L Design pressure 200 psig Design temperature 150°F Design code ASME VIII Evaporator Recirculation Pump Number 1 Type Axial Flow Motor horsepower 75 hp Seal type Double Mechanical Capacity (each) 9000 gpm Head at rated capacity 15 ft Design pressure 60 psig Materials Casing Alloy 20 Shaft Alloy 20 Impeller Alloy 20 Oil Drain Tank Number 1 Capacity 1070 gal Design pressure Atmospheric Plus Water Full Design temperature 150°F Operating pressure Full of Content Operating temperature 104°F Material SS 316L Design Code ASME III Vapor Recompressor Number 1 Type Centrifugal Motor horsepower 600 hp Seal type Labyrinth Capacity (each) 7904 ACFM Compression Ratio 1.97 Design pressure 150 psig

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-29 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Vapor Recompressor (continued)

Materials Casing SS 316L Shaft SS 316L Impeller SS 316L Distillate Flash Tank Number 1 Capacity 150 gal Design pressure 30 psig Design temperature 300°F Operating pressure 0.5 psig Operating temperature 212°F Material SS 316L Design Code ASME VIII Distillate Subcooler Number 1 Total duty 1,589,994 Btu/hr Shell Tube Design pressure 150 psig 75 psig Design temperature 250°F 300°F Operating pressure 60 psig 52 psig Operating temperature, in/out 80/92.5°F 212/120°F Material SS 304/Tube Side Carbon Steel/Sheet Side Fluid Cooling Water Distillate Design code ASME VIII Evaporator Heater Number 1 Total duty 17,787,245 Btu/hr Shell Tube Design pressure 50 psig 45 psig Design temperature 300°F 300°F Operating pressure 15 psig 5 psig Operating temperature, in/out 218.3/222.

247.8/247.8°F 3°F Material SS 316L/Alloy 20 Inconel 625 Shell Tube Fluid Steam Liquid Waste Design code ASME VIII

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-30 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Liquid Waste Monitor Tanks Number 2 Capacity 15,000 gal Design pressure Atmospheric Plus Contents Design temperature 150°F Operating pressure Full of Water Operating temperature 104°F Material SS 304 Design Code ASME III Liquid Waste Monitor Tank Pumps Number 2 Type Centrifugal Motor horsepower 60 hp Seal type Single Mechanical Capacity (each) 330 gpm Head at rated capacity 326 ft Design pressure 275 psig Materials Casing 316 SS Shaft Steel Impeller 316 SS Laundry Drain Pre-filters Number 2 Rentention size 50 microns Filter element material Bag Capacity, normal 50 gpm Capacity maximum 50 gpm Material SS 304 Design pressure 150 psig Design temperature 150°F Design Code ASME VIII Laundry Drain Filter Number 1 Rentention size, microns N/A Filter element material Polyester Capacity, normal 50 gpm Capacity maximum 50 gpm Material SS 304 Design pressure 150 psig

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-31 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Laundry Drain Filter (continued)

Design temperature 150°F Design Code ASME VIII Laundry Drain Monitor Tank Number 2 Capacity 7500 gal Design pressure Atmospheric Plus Content Design temperature 150°F Operating pressure Full of Water Operating temperature 104°F Material SS 304 Design Code ASME III Laundry Drain Monitor Pump Number 2 Type Centrifugal Motor horsepower 40 hp Seal type Single Mechanical Capacity (each) 140 gpm Head at rated capacity 310 ft Design pressure 225 psig Materials Casing Ductile Iron Shaft Steel Impeller Cast Iron Evaporator Bottoms Tank Number 1 Capacity 5000 gal Design pressure Atmospheric Plus Contents Design temperature 275°F Operating pressure Full of Water Operating temperature 212°F Material Inconel 625 Design Code ASME III Evaporator Bottoms Tank Pump Number 1 Type Centrifugal Motor horsepower 7.5 hp Seal type Double Mechanical Capacity (each) 165 gpm

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-32 Table 11.2-1 (CONTINUED)

WASTE DISPOSAL SYSTEMS DESIGN DATA Evaporator Bottoms Tank Pump (continued)

Head at rated capacity 60 ft Design pressure 142.2 psig Materials Casing SS 316 Shaft SS 316 Impeller SS 316 Evaporator Bottoms Tank Number 1 Capacity 5000 gal Design pressure Atmospheric Plus Contents Design temperature 275°F Operating pressure Full of Water Operating temperature 180°F Material Inconel 625 Design code ASME III Liquid Waste Demineralizer Vessels Number 3 Capacity 29 ft3 Design flow 30 gpm Design pressure 150 psig Material SS 304 Design code ASME VIII RF Liquid Waste Reverse Osmosis Unit Number 1 Capacity 25 gpm Materials SS, PVC Pressure rating 150 psig (low pressure portion) 500 psig (high pressure portion)

Design code ANSI B31.1 Distillate Demineralizers Number 1 Capacity 50 ft3 Design flow 30 gpm Distillate Demineralizers (continued)

Design pressure 150 psig Material SS 304 Design code ASME VIII

Revision 52Updated Online 09/30/20 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 11.2-2 ESTIMATED WASTE GAS RELEASE WITH RECOMBINER NOT OPERATING Based on gaseous activity from:

a. Waste gas cycle: 30 days of feed/20 days of decay/10 days of bleed.

b. Two units, 2546 MWt each.

c. 1% failed fuel.

d. 17-gpm total letdown flow rate to gas stripper from both units.

e. 100% of gases and 0.1% of iodines removed in gas stripper.

f. X/Q - 7.5 x 10-6 sec/ m3 for calculating dose at site boundary.

Equilibriu Curies In Curies In Discharge Rate (Ci/sec) m Coolant Tank At Tank After Curies Total Curies Dose At Site Activity End Of 20 Days Discharged Discharged Boundary Nuclide (Ci/cm3) Feed Cycle Decay Initial Final In One Cycle Per Year (rem/yr)

Kr-85m 1.22 3.63 x 101 2.05 x 10-35 2.14 x 10-41 1.61 x 10 -59 4.44 x 10-37 5.40 x 10-36 6.4 x 10-42 Kr-85 3.25 x 10-1 1.25 x 103 1.25 x 103 1.30 x 10-3 1.30 x 10-3 1.12 x 103 1.36 x 104 5.4 x 10-3 Kr-87 8.27 x 10-1 7.96 0 0 0 0 0 0 Kr-88 2.36 4.92 x 101 0 0 0 0 0 0 Xe-131m 4.62 x 10-2 1.95 x 102 6.16 x 101 6.43 x 10-5 3.61 x 10-5 4.22 x 101 5.13 x 102 1.6 x 10-4 SPS UFSAR Xe-133m 1.33 5.41 x 102 1.30 1.36 x 10-6 6.66 x 10-8 3.70 x 10-1 4.50 1.8 x 10-6 Xe-133 80.9 7.52 x 104 5.47 x 103 5.71 x 10-3 1.54 x 10-3 2.75 x 10-3 3.35 x 104 1.3 x 10-2 Xe-135m 7.13 x 10-1 1.40 5.70 x 10-25 5.95 x 10-31 1.02 x 10-41 2.07 x 10-26 2.52 x 10-25 3.0 x 10-31 Xe-135 3.39 2.27 x 102 2.80 x 10-14 2.92 x 10-20 3.24 x 10-28 1.38 x 10-15 1.68 x 10-14 1.9 x 10-20 Xe-138 4.50 x 10-1 9.59 x 10-1 0 0 0 0 0 0 1-131 1.72 2.29 x 10-1 4.11 x 10-2 4.28 x 10-8 1.81 x 10-8 2.48 x 10-2 3.01 x 10-1 1.2 x 10-3 1-132 5.85 x 10-1 1.00 x 10-3 0 0 0 0 0 0 1-133 2.65 4.15 x 10-2 5.44 x 10-9 5.68 x 10-15 2.06 x 10-18 6.19 x 10-10 7.53 x 10-9 7.9 x 10-12 1-134 3.61 x 10-1 2.35 x 10-4 0 0 0 0 0 0 1-135 1.40 7.01 x 10-3 2.03 x 10-24 2.12 x 10-30 3.60 x 10-41 7.38 x 10-26 8.98 x 10-25 2.9 x 10-25 Total 7.76 x 104 6.78 x 103 7.07 x 10-3 2.87 x 10-3 3.91 x 103 4.76 x 104 1.9 x 10-2 11.2-33

Revision 52Updated Online 09/30/20 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 11.2-3 ESTIMATED WASTE GAS RELEASE WITH RECOMBINER OPERATING Based on gaseous activity from:

a. Waste gas cycle: 300 days of feed/20 days of decay/10 days of bleed.

b. Two units, 2546 MWt each.

c. 1% failed fuel.

d. 17-gpm total letdown flow rate to gas stripper from both units.

e. 100% of gases and 0.1% of iodines removed in gas stripper.

f. X/Q - 7.5 x 10-6 sec/ m3 for calculating dose at site boundary.

Equilibriu Curies In Curies In Discharge Rate (Ci/sec) m Coolant Tank At Tank After Curies Total Curies Dose At Site Activity End Of 20 Days Discharged Discharged Boundary Nuclide (Ci/cm3) Feed Cycle Decay Initial Final In One Cycle Per Year (rem/yr)

Kr-85m 1.22 3.63 x 101 2.05 x 10-35 2.14 x 10-4 1.61 x 10 -59 4.44 x 10-37 4.44 x 10-37 5.3 x 10-43 Kr-85 3.25 x 10-1 1.63 x 104 1.63 x 104 1.70 x 10-3 1.69 x 10-2 1.46 x 104 1.46 x 104 5.8 x 10-3 Kr-87 8.27 x 10-1 7.96 0 0 0 0 0 0 Kr-88 2.36 4.92 x 101 0 0 0 0 0 0 Xe-131m 4.62 x 10-2 1.95 x 102 6.16 x 101 6.43 x 10-5 3.61 x 10-5 4.22 x 101 4.22 x 101 1.3 x 10-5 SPS UFSAR Xe-133m 1.33 5.42 x 102 1.30 1.36 x 10-6 6.67 x 10-8 3.71 x 10-1 3.71 x 10-1 1.5 x 10-7 Xe-133 80.9 7.68 x 104 5.58 x 103 5.82 x 10-3 1.57 x 10-3 2.80 x 10-3 2.80 x 103 1.1 x 10-3 Xe-135m 7.13 x 10-1 1.40 5.75 x 10-25 6.00 x 10-31 1.02 x 10-41 2.09 x 10-26 2.09 x 10-26 2.5 x 10-32 Xe-135 3.39 2.32 x 102 8.52 x 10-14 2.98 x 10-20 3.30 x 10-28 1.40 x 10-15 1.40 x 10-15 1.7 x 10-21 Xe-138 4.50 x 10-1 9.59 x 10-1 0 0 0 0 0 0 1-131 1.72 2.48 x 10-1 4.44 x 10-1 4.64 x 10-8 1.96 x 10-8 2.69 x 10-2 2.69 x 10-2 1.0 x 10-4 1-132 5.85 x 10-1 1.00 x 10-3 0 0 0 0 0 0 1-133 2.65 4.23 x 10-2 5.50 x 10-9 5.79 x 10-15 2.10 x 10-18 6.31 x 10-10 6.31 x 10-10 6.6 x 10-13 1-134 3.61 x 10-1 2.35 x 10-4 0 0 0 0 0 0 1-135 1.40 7.07 x 10-3 2.05 x 10-24 2.14 x 10-30 3.63 x 10-41 7.44 x 10-26 7.44 x 10-26 2.4 x 10-29 Total 9.41 x 104 2.19 x 104 2.28 x 10-2 1.85 x 10-2 1.75 x 104 1.75 x 104 7.0 x 10-3 11.2-34

Revision 52Updated Online 09/30/20 SPS UFSAR 11.2-35 Table 11.2-4 LIQUID WASTE DISPOSAL SYSTEM Clean and Equipment Dirty Steam Generator Detergent Shim Bleed Drains Wastes Blowdown Wastes Reactor coolant Primary Waste drain Laundry Sources letdown drain tanks tanks wastes Flow rate, gpd 2500 1150 10,150 90,600 1250 Activity, FPCA 1.0 1.0 0.36 Collection tank volume, gal 120,000 120,000 5804 1230 Collection rate, gpda 7,300b 7,300b 20,300 2500 Collection time days 13.2 13.2 0.23 0 0.39 Processing rate, gpd 31,700 31,700 34,560 72,000 Processing time, days 3.03 3.03 0.14 0 0.01 Discharge tank volume, gal 30,000 30,000 3548 1, 230 Discharge rate, gpd 7, 300 7, 300 72,000 72,000 Discharge time, days 3.28 3.28 0.04 0 0.01 Fraction of processed steam released 1.0 1.0 1.0 1.0 1.0 Anion Mixed-Bed Demin.

Ion Same as Mixed-Bed Exch. Evap. shim bleed Demin Case 1 Case 2 DFs Not 102(10) None I 102 102 102(10) treated 10 (10)

Cs, Rb 1 103 2 (10) 102 Others 1 103 102 (10) (10)

Regenerant time, Not Not days Not regenerated regenerated regenerated Not regenerated NA See See Tables See Tables See Tables Tables 11A-9 11A-9 11A-9 See Tables 11A-9 Source terms and 11A-10 and 11A-10 and 11A-10 11A-9 and 11A-10 and 11A-10

a. Reflects shared system.

b. Sum of equipment drains and shim bleed for both units.

Revision 52Updated Online 09/30/20 Figure 11.2-1 LIQUID WASTE DISPOSAL SYSTEM; PROCESS FLOW DIAGRAM, SHEET 1 SPS UFSAR 11.2-36

Revision 52Updated Online 09/30/20 Figure 11.2-2 LIQUID WASTE DISPOSAL SYSTEM; PROCESS FLOW DIAGRAM, SHEET 2 SPS UFSAR 11.2-37

Revision 52Updated Online 09/30/20 Figure 11.2-3 LIQUID WASTE SYSTEM PROCESS FLOW DIAGRAM SPS UFSAR 11.2-38

Revision 52Updated Online 09/30/20 Figure 11.2-4 LAUNDRY WASTE SYSTEM PROCESS FLOW DIAGRAM SPS UFSAR 11.2-39

Revision 52Updated Online 09/30/20 Figure 11.2-5 BITUMEN SOLIDIFICATION SYSTEM PROCESS FLOW DIAGRAM SPS UFSAR 11.2-40

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-1 11.3 RADIATION PROTECTION 11.3.1 Design Bases Radiation protection, including radiation shielding, was designed to ensure that the criteria specified in 10 CFR 201 are met during normal operation and that the guidelines in 10 CFR 50.67 are met in the event of the design-basis loss of coolant accident (Section 14.5.5). Maximum dose limits for design-basis accident conditions are given in Table 14.5-11.

Allowable dose rates are based on the expected frequency and duration of occupancy.

Occupancy time and dose rates are such that no personnel shall receive in excess of those doses recommended in 10 CFR 20. All dose rate calculations are based on 1% failed fuel elements.

Allowable dose rates for typical locations are given in Table 11.3-1.

11.3.1.1 Leak Reduction Program A leak reduction program has been established to minimize leakage from systems outside containment that would or could contain highly radioactive fluids during transients or accidents.

The following systems, which are described in detail elsewhere in the FSAR, are at least partially located outside containment and are expected to contain potentially radioactive fluids immediately following an accident:

1. Safety injection system.

2. Containment and recirculation spray systems.

3. CVCS (those portions associated with safety injection).

4. Boron recovery system.

5. Resin waste disposal system.

6. Sampling system.

7. Containment vacuum system.

8. Containment purge system.

Several plant systems have been excluded from the leak reduction program. Their exclusion is justified because their unavailability would not eliminate any of the options for cooling the reactor core, nor would it prevent the use of any safety system. The excluded systems are the following:

1. CVCS (those portions not associated with safety injection).

2. Purification system.

1. Virginia Power implemented the revised 10 CFR 20 on January 1, 1994. However, as allowed by the NRC, the calculational methodology used for the design analyses is based on the revision of 10 CFR 20 to which the plant was originally licensed.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-2

3. Gas stripper system.

4. Vent and drain systems.

5. Liquid waste disposal system.

6. Reactor cavity purification system.

7. Spent-fuel pool cooling and purification systems.

The letdown portion of the CVCS is normally used for reactor coolant system inventory control, reactor coolant pump seal injection, and reactor coolant system purification. After an accident, the safety injection system using the refueling water storage tank or containment sump would provide the necessary inventory control and seal injection functions. Coolant purification would be deferred until some time after the accident and would probably be performed using a temporary system. Since CVCS letdown will not be used after an accident, the gas stripper system and the vent and drain systems are not needed to support the associated vent. The containment sump would be used for liquid waste storage in place of the liquid waste storage system. The reactor cavity purification system is used during refueling outages. The spent-fuel pool cooling and purification system is not connected to the reactor coolant system or to containment.

A preventive maintenance program, including periodic leak tests, has been established for the systems in the leak reduction program. This program uses administrative controls and procedures as outlined by the plant quality assurance program.

11.3.2 Shielding Design and Evaluation 11.3.2.1 Primary Shielding Primary shielding is provided to limit radiation emanating from the reactor vessel. The radiation consists of neutrons diffusing from the core, prompt fission gammas, fission product gammas, and gammas resulting from the slowing down and capture of neutrons.

The primary shielding is designed to:

1. Attenuate neutron flux to prevent excessive activation of unit components and structures.

2. Reduce the contribution of radiation from the reactor to obtain a reasonable division of the shielding function between primary and secondary shields.

3. Reduce residual radiation from the core to a level that does not limit access to the region between the primary and secondary shields at a reasonable time after shutdown.

4. Postaccident shielding considerations are discussed in Section 11.3.2.9.

The primary shield consists of a water-filled neutron shield tank with a radial dimension of approximately 3 feet, surrounded by 4.5 feet of reinforced concrete. The neutron shield tank is designed to prevent overheating and dehydration of the concrete primary shield wall and to

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-3 prevent activation of the plant components within the reactor containment. A thermosiphon cooling system is provided for cooling the water in the shield tank.

A 2-inch-thick cylindrical lead shield that is approximately 15-foot-high is located beneath the neutron shield tank to protect station personnel servicing the neutron detectors during reactor shutdown.

A 3-1/2-inch thick stainless steel radiation shield is provided at the 12-inch diameter incore sump room drain to protect station personnel during normal power operation and refueling outage. The drain is designed to convey the held up water, in excess of the invert elevation of the Incore Sump Room drain, from the Incore Sump Room to the containment sump strainer. This additional water facilitates submergence of the containment sump strainer for RS and LHSI pumps for post-LOCA operation.

To maintain the integrity of the primary shield, streaming shields fabricated from both masonite Benelex 70 and steel are provided in the annular gap between reactor vessel flange and the primary shield concrete. In addition, masonite Benelex 401 and steel streaming shields located outside the primary concrete shield are provided around all of the reactor coolant pipe penetrations.

The primary shield arrangement is shown in Figures 11.3-1 and 11.3-2. The shield materials and thicknesses are listed in Table 11.3-2.

11.3.2.2 Secondary Shielding Secondary shielding consists of reactor coolant loop shielding, reactor containment shielding, fuel handling shielding, auxiliary equipment shielding, and waste storage shielding.

Nitrogen-16 is the major source of radioactivity in the reactor coolant during normal operation and establishes the combined thickness of the crane and containment walls. Activated corrosion and fission products in the reactor coolant system establish the shutdown radiation levels in the reactor coolant loop areas. Tables 9.1-4 and 11.3-3 list the activities that were used in designing the containment secondary shielding. Table 9.1-4 lists the fission product activities in the reactor coolant system with 1% failed fuel. Table 11.3-3 lists the activated corrosion product activities and the N-16 activity at the reactor vessel outlet nozzle.

Activated corrosion and fission products from the reactor coolant system are the radioactive sources for which shielding is required in the auxiliary and waste disposal systems.

Auxiliary steam used for space heating and other purposes throughout the station may become contaminated due to primary-to-secondary leakage. The estimated dose rate at the surface of a 6-inch, 15-psig auxiliary steam supply header is 2 x 10-6 mrem/hr. The estimated dose rate at 3 feet is 1.25 x 10-7 mrem/hr. The dose rate estimates are based on 1% failed fuel activities (Table 9.1-4), 25% steam flow, 1000 cm3/hr total steam generator leakage, and zero decay time.

All noble gases and 1% of the halogens are assumed to leak into the steam generator. Based on

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-4 these assumptions, the dose received by an individual in the vicinity of the auxiliary steam piping is insignificant.

11.3.2.3 Reactor Coolant Loop Shielding Interior shield walls separate reactor coolant loop, pressurizer, incore instrumentation, and containment access sectors. This shielding allows access to the incore instrument sector during normal operation and facilitates maintenance in all sectors during shutdown. The crane support wall provides limited access protection in the annulus between the crane wall and the reactor containment wall and provides part of the exterior shielding required during power operation.

Shield walls are provided around each steam generator above the charging floor to a height required for personnel protection. The shielding arrangement is shown in Figures 11.3-1 and 11.3-2. The shield materials and thicknesses are listed in Table 11.3-2.

11.3.2.4 Reactor Containment Shielding The containment shielding consists of the steel-lined, steel-reinforced concrete cylinder and hemispherical dome, as further described in Chapter 5. This shielding, together with the crane support wall, attenuates radiation during full-power operation at the outside surface of the containment to less than 0.75 mrem/hr. In addition, it attenuates the dose rate from the design-basis accident to design levels.

11.3.2.5 Fuel Handling Shielding Fuel handling shielding is designed to facilitate the removal and transfer of spent fuel assemblies from the reactor vessel to the spent-fuel pool. It is designed to protect personnel against the radiation emitted from the spent fuel and control rod assemblies.

The refueling cavity above the reactor vessel is flooded to Elevation +45-1/3 ft. to provide a temporary water shield above the components being withdrawn from the reactor vessel. The water height is thus approximately 27 feet above the reactor vessel flange. This height ensures that more than 84 inches of water is present above a withdrawn fuel assembly at its highest point of travel.

Under these conditions, the dose rate is less than 50 mrem/hr at the water surface.

The fuel is removed from the reactor vessel to the spent-fuel pool by the fuel transfer mechanism via the refueling canal.

The spent-fuel pool in the fuel storage building is permanently flooded to provide more than 84 inches of water above a fuel assembly when it is being withdrawn from the fuel assembly transfer basket. Water height above stored fuel assemblies is at least 24 feet. The sides of the spent-fuel pool, three of which also form part of the fuel storage building exterior walls, are 6-foot-thick concrete to ensure a dose rate of no more than 2.5 mrem/hr outside the building.

Sixteen feet of earth shielding is provided above the fuel transfer tube between the reactor containment and the fuel storage pool wall.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-5 11.3.2.6 Auxiliary Equipment Shielding The auxiliary components exhibit varying degrees of radioactive contamination due to the handling of various fluids. The function of the auxiliary shielding is to protect personnel working near the various auxiliary system components, such as those in the CVCS, the boron recovery system, the waste disposal system, and the sampling system. Controlled access to the auxiliary building is allowed during reactor operation. Each equipment compartment is individually shielded so that compartments may be entered without having to shut down and, possibly, decontaminate the entire system. Ilmenite concrete is used in certain areas where substantial shielding is required and space is at a premium, such as the primary drain tank compartment and the mixed-bed demineralizer compartments.

Ion exchangers and the most highly contaminated filters are located in the ion-exchange structure along the north wall of the auxiliary building. Each ion exchanger or filter is enclosed in a separate, shielded compartment. The concrete thicknesses provided around the shielded compartments are sufficient to reduce the surrounding area dose rate to less than 2.5 mrem/hr, and the dose rate of any adjacent cubicle to less than 100 mrem/hr. The shielding thicknesses around the mixed-bed demineralizers are based on a saturation activity that gives a contact radiation level of nearly 11,000 rem/hr.

Ion exchangers and potentially contaminated filters are also associated with the steam generator blowdown treatment system (no longer used) located in the condensate polisher building.

In many areas, tornado missile protection in the form of thick concrete affords more shielding than that required for radiation protection.

11.3.2.7 Waste Storage Shielding The waste storage and processing facilities in the auxiliary building and decontamination building, the RF, and the waste storage tanks are shielded to provide protection of operating personnel in accordance with the radiation protection design bases set forth in Section 11.3.1.

Periodic surveys by health physics personnel using portable radiation detectors ensure that radiation levels outside the shield walls meet design specifications, and establish access limitations within the shielded cubicles. In addition, continuous surveillance is provided at the Radwaste Facility drumming area and control area and the RF compactor area by area radiation monitors (Section 11.3.4).

Area and process monitoring also ensure that any accidental radioactivity release would be detected within a reasonable period of time. The largest accidental radioactivity release from the waste disposal system would be the rupture of one of the waste gas decay tanks. An analysis of this accident is provided in Section 14.4.2. Furthermore, periodic samples of the gas in the waste gas decay tanks are analyzed by health physics personnel to ensure that the activity level in these tanks is never above the design level used in the accident analysis.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-6 11.3.2.8 Accident Shielding Accident shielding is provided by the reactor containment, which is a reinforced concrete structure lined with steel. For structural reasons, the thicknesses of the cylindrical walls and dome are 54 and 30 inches, respectively. These thicknesses are more than adequate to meet the requirements of 10 CFR 100 at the exclusion boundary.

Additional shielding is provided for the control room. This, together with the shielding afforded by its physical separation from the containment structure, ensures that an operator would be able to remain in the control room for 30 days after an accident and not receive an integrated whole-body dose in excess of 5 rem. The calculational methods and radiation sources used in designing the control room shielding are discussed in Section 11.3.6.

In addition, the control room will serve as a fallout shelter with a protection factor of better than 500, as defined by the Office of Civil Defense.

11.3.2.9 Postaccident Shielding Review A postaccident shielding review was conducted during the 1979 to 1980 time period using the Stone & Webster GAMTRAN1 computer code with inputs developed from the ACTIVITY-2 and RADIOISOTOPE computer codes. NRC-specified source terms were used. The review assumed that the postaccident period was divided into two phases: the mitigation phase and the recovery phase.

The mitigation phase, which was assumed to last 6 months, experiences radiation levels resulting from the operation of the recirculation portion of the safety injection and recirculation spray systems and the postaccident sampling system. This phase also experiences radiation levels from the auxiliary building sump and from the drain lines from the discharge of the auxiliary building and safeguard building sump pumps to the low-level liquid waste tank, as well as to the containment.

The recovery phase has been identified as the period beginning 6 months after the accident, when cleanup and plant recovery is performed. The recovery phase is a controlled evolution that will be planned and carried out to meet the specific recovery requirements of the particular accident.

All essential system piping and equipment that are required to mitigate the effects of a loss-of-coolant accident (LOCA), and that contain or could contain highly radioactive fluids, were considered as sources in the shielding review. These systems included the high-head safety injection system, the low-head safety injection system, the recirculation spray system, the sample system, and the containment atmosphere cleanup (hydrogen recombiner) system. In addition, other systems that are not required to mitigate a LOCA, but that could contain significant radioactivity, were considered, such as drain lines and standing water in sumps and waste tanks.

All branch connections to and from these systems were considered as sources to the first isolation valve. Other sources, such as the shine from the containment dome, shine through containment

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-7 penetrations, and shine through the personnel hatch, were considered. The location of field run pipe, which is part of the systems listed above, was considered in this analysis. The routing and location of radioactive piping is such that the piping is in shielded areas. The exact routing of field run pipe is not critical in the production of radiation zone maps. The highest activity level in each zone is calculated, and that level is considered for the entire zone. For instance, the highest activity may be 12 inches from a pipe, regardless of its exact location within the zone.

Indirect radiation was not considered as a source. Buildup factors in shield walls are considered, but scatter over walls or through labyrinth doorways was not considered. Airborne activity was not considered as a source in the shielding review.

All vital areas were also considered during the review. Vital areas for personnel exposure are defined as those areas that require continuous or frequent occupancy in order to control, monitor, and evaluate the accident. These areas include the control room, technical support center, the counting lab/health physics area, the operational support center, and the security control center. In addition, any area to which access is required to perform manual operation of equipment in systems that are used to mitigate the accident was considered. Vital areas for equipment qualification include all areas in which mitigating equipment is located. Nonvital areas include the entire auxiliary building, main steam valve house, quench spray pump house, safeguards building, service building, and selected areas in the yard.

11.3.2.9.1 Mitigation Phase The integrated radiation dose calculated for this phase is comprised of the original license period of 40-year normal dose and a 6-month mitigation phase dose. The safety equipment required to operate during the mitigation phase is the same as that equipment tabulated in Chapter 7 for NRC I&E Bulletin 79-01 (Reference 1). The source term developed to calculate the 40-year normal operating dose is based on the assumptions in Chapter 11. The source terms assumed to calculate the 6-month mitigation phase dose are based on TID-14844 (Reference 2) and Regulatory Guide 1.4 (Reference 3), and are listed in Table 11.3-4. The impact of increased integrated dose associated with an additional 20 years of normal operation is small and is accounted for in the environmental qualification of equipment applied to the mitigation phase.

The exact course of an accident is unpredictable. It is impractical to determine the dose rate and shielding requirements for every possible location and time duration associated with each possible failure, or incident, that requires personnel access. Therefore, radiation zone maps have been developed for use as postaccident administrative guidelines. These maps show estimated worst-case gamma rates in various areas of the plant as a function of time. The zone maps are used to help plant operators plan access and egress routes, and to help evaluate the relative benefits of delaying certain actions to allow for radioactive decay. The gamma dose rates on the zone maps are based on worst-case source terms, but do not consider an airborne source term. Depending on the severity of the situation, actual dose rates may be smaller. The zone maps are used only as guidelines, and actual radiation levels will be determined through actual

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-8 postaccident surveys. In addition, the dose rates on the zone maps are based on the highest, or one of the highest, dose rates in that area. The dose rates at other locations within that area may be lower.

11.3.2.9.2 Recovery Phase The design basis for certain systems and their associated shielding did not consider postaccident recovery operations, that is, postaccident cleanup of highly radioactive fluids. These are the waste disposal system, the boron recovery system, the containment purge system, and the letdown and charging portions of the CVCS. As a result, these systems will not be used for postaccident cleanup operations.

The activity levels (based on Regulatory Guide 1.4 and TID-14844) of the influent to the liquid waste disposal system or to the boron recovery system are approximately 2 x 103 Ci/cm3 after 6 months of radioactive decay. The area radiation dose rates from concentrated waste and from waste storage tanks would severely limit access to parts of the auxiliary building and would hinder the operation of both units. Since the radioactive waste disposal systems are common to both Units 1 and 2, the use of these systems for the cleanup of waste in the accident-affected unit would preclude the normal use of the radioactive systems for the non-accident-affected unit.

There is extensive piping for the above-listed systems throughout the auxiliary building.

The resulting dose rate if all these systems operated simultaneously would severely limit freedom of access for required operations. Shielding for the piping and components would be very difficult, and in some cases impossible, to install, because of the physical arrangement of the piping and components.

11.3.2.9.3 Postaccident Sampling Capability The Surry Power Station has the capability to sample the reactor coolant, containment sump, and the containment atmosphere. The reactor coolant sample can be taken and analyzed within approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the decision to sample. A containment atmosphere sample can also be taken with the high radiation sample system. Provisions are included for personnel exposure control. A detailed discussion of the postaccident sampling system is contained in Chapter 9.

11.3.2.10 Steam Generator Storage Facility The Steam Generator Storage Facility is a reinforced concrete structure located within the station but outside the protected area. Major components housed in the facility include the original Steam Generator lower assemblies and the original Reactor Vessel Closure Heads. The facility has no interface with any permanent plant structures and is a restricted area. It is included in the Radiological Survey Program.

The radionuclide inventory of a steam generator as it is removed from the containment has been estimated about 1400 curies. This will decay to about 2-20 curies in 30 years and 0.0027 to

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-9 0.027 curies in 80 years. At the end of station life, it can be conservatively assumed that the total inventory of 12-120 curies of Cobalt-60 for the 6 steam generators will represent less than 0.1%

of the total activity on site, exclusive of fuel and control rods. The lower shells could be handled easily as part of station decommissioning activities.

The facility is an above ground concrete structure on a poured structural slab. The exterior walls are as thick as necessary to meet the 40 CFR 190 requirement of 25 millirem/year. The facility is designed as an 80 year structure. It is not necessary that the facility be a seismic category I structure. The facility is designed with a dose rate criteria of radiation Zone I for the contact dose.

The original steam generator lower shells are tracked for decay and the radiation safeguards of the storage building is maintained. The exterior walls of the facility are periodically measured for low-level radiation dose.

The Steam Generator Storage Facility sump is periodically pumped out and taken to the Laundry Drain System for processing if contaminated. Table 11.3-6 list the Laundry Facility Continuous Effluent Particulate and Iodine.

11.3.3 Process Radiation Monitoring System The process radiation monitoring system continuously monitors selected lines containing, or possibly containing, radioactive effluents. Lines through which waste liquids and gases are discharged to the environment are also monitored. The function of this monitoring system is to warn personnel of increasing radiation levels that could result in a radiation health hazard and to give early warning of a system malfunction. An audible alarm in the control room has been incorporated to indicate the loss of power to the monitor cabinet. The process radiation monitoring system serving both units is comprised of the channels listed in Table 11.3-5.

Each radiation monitoring channel is designed to provide continuous information about the process or effluent stream being monitored. Continuous, as used to describe the operation of the process and effluent radiation monitoring systems, means that a monitor provides the required information at all times with the following exceptions: (1) the system is not required to be in operation because of specified plant conditions per the Technical Specifications, or (2) the system is out of service for testing or maintenance and approved alternate monitoring, sampling, or recording methods are in place.

With the exception of the monitoring channels from the Laundry and Radwaste Facilities, each channel has a readout in the control room, and selected channels, as indicated in Table 11.3-5, have a readout at the detector location. The Laundry Facility monitors all readout locally and the Radwaste Facility monitors readout in the Radwaste Facility control room as well as locally. In addition, each channel has an audible and visual alarm for radiation levels in excess of preset values, as well as a visual alarm for detector malfunction. The output from all channels is recorded on recorders that produce a continuous record of radiation levels and radioactive

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-10 discharges from the station. Each channel has its own power supply and check source thus making it completely independent of any other channel. Each channel check source is remotely operated from the main control room except the process monitors in the Radwaste Facility, the Laundry Facility, and the High-Range Effluent monitors. The normal/high range Process Vent and Vent Stack No. 2 (MGPI) effluent monitors do not have a check source. The High-Range Effluent monitors for the main steam line and the exhaust of the turbine AFW pump have a built-in radioactive source that provides a life-zero signal for testing purposes. The Radwaste Facility and the Laundry Facility check sources are remotely operated from local control panels. The MGPI equipment is continually self-checking that, should the detector output signal drop below some prescribed value, effectively indicating that there is no background signal, the detector will be considered faulty. This monitoring, in conjunction with continual testing of the electronic signal processing circuits by generation and confirmed receipt of test signals, provides assurance that the circuits remain in good health. The adjustment of alarm setpoints, voltage, power, and other variables is made from the control room for all radiation monitors except for the following:

1) monitors in the Radwaste, and 2) monitors in the Laundry Facilities. Adjustments to the Radwaste and Laundry Facility monitors are performed locally. The normal/high range Process Vent and Vent Stack No. 2 MGPI effluent monitors can be adjusted either at the local display unit (LDU) or at the remote display unit (RDU). The High-Range Effluent monitors can be adjusted at the local processing and display unit (LPDU). The entire system is designed with emphasis on system reliability and availability. Certain channels, as indicated in the following text, actuate control valves on a high-activity alarm signal. In the event of a loss of power to these detectors, the system is designed to provide an alarm of the failed condition. Any control functions associated with a high radiation alarm are also initiated.

The expected concentrations of radionuclides in the process streams monitored by the ventilation vent monitors, component cooling water monitors, component cooling heat exchanger service water monitors, condenser air ejector monitors, steam generator blowdown monitors, and recirculation spray cooler heat exchanger service water monitors are natural background radioactivity. The sensitivity of these detectors ensures that abnormal plant conditions will be detected before they cause a hazard to the operators or to the general public.

The use of a single detector is justified in lines used for normal releases from the plant. The surveillance requirements for each of the liquid effluent monitors and gaseous effluent monitors are given in the Offsite Dose Calculation Manual. The liquid and gas waste tanks are sampled and analyzed before and during discharges. Effluent source terms are discussed in Appendix 11A.

Channels monitoring Unit 1 are supplied from the emergency bus for Unit 1. Channels monitoring Unit 2 are supplied from the emergency bus for Unit 2. Channels monitoring systems or areas common to both units can be supplied from the emergency bus for either Unit 1 or Unit 2.

The type of detector, sensitivity, range, background radiation, and other information for each channel are listed in Table 11.3-5. Counting rates are given in Table 11.3-6. A description of each channel is included in the following text.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-11 11.3.3.1 Process Vent Particulate Monitor This channel continuously withdraws a sample from the process vent and passes the sample through a moving filter paper with a collection efficiency of 99% for particle sizes greater than 1.0. The amount of deposited activity is continuously scanned by a silicon diode type. A high-activity alarm automatically initiates the closure of the process vent discharge line valves.

A separate isokinetic sampling nozzle used for Health Physics accountability is provided for each unit, as shown in Reference Drawing 2. The nozzles sample the process vent fluid to ensure that a representative sample is taken. Isokinetic sampling is achieved by locating the nozzles in a straight unobstructed piping run of at least five pipe diameters (30 inches). The sampling systems include a pump and a mass flow meter, and supply 1-cfm flow to a sample filter.

11.3.3.2 Process Vent Gas Monitor This channel takes the continuous process vent sample, after it has passed through the particulate filter paper, and draws it through a sealed system to the process vent gas monitor assembly, which is a fixed lead-shielded sampler containing a silicon diode type detector. The sample activity is measured, and is then returned to the process vent. A high-activity alarm automatically initiates the closure of the process vent discharge line valves. A purge system is integral with the gas monitoring system for flushing the sampler with clean air for purposes of calibration.

11.3.3.3 Ventilation Vent No. 1 Gas Monitor This channel withdraws a sample from ventilation vent no. 1 and passes the sample through a particulate and iodine filter assembly. The sample then enters a gas monitor assembly, which is a fixed lead-shielded sampler enclosing a beta scintillation detector. The sample activity is measured, and is then returned to the ventilation vent. A purge system is integral with the gas monitoring system for flushing the sampler with clean air for purposes of calibration. A multi-probe isokinetic sampling nozzle is provided to obtain a representative sample in the duct.

11.3.3.4 RF Vent Particulate and Gas Monitors These two channels continuously sample the RF ventilation stack particulate and gas. The monitors process a sample (approximately 2 scfm) through a replacable fixed 0.3-micron filter.

Noble gases are then monitored in a pressure compensated chamber. Sample flows are supplied by an isokinetic probe. This monitor has provisions for a removable silver zeolite iodine sampler cartridge and for a grab sample for tritium analysis.

11.3.3.5 Component Cooling Water Monitors These two channels continuously monitor the component cooling water by means of a gamma scintillation detector enclosed in lead shielding and mounted on the component cooling piping. The complex piping arrangement of this system dictates that two detectors are required to

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-12 ensure that the system is properly monitored. Activity is indicative of a leak into the component cooling system (Section 9.4) from one of the radioactive systems that exchange heat to the component cooling system.

11.3.3.6 Component Cooling Heat Exchanger Service Water Monitor These four channels continuously monitor the service water effluent from the four component cooling water heat exchangers. Each channel consists of an inline gamma scintillation detector located in a pipe well on the discharge side of the heat exchanger.

11.3.3.7 Liquid Waste Disposal System Monitor The RF has a liquid waste discharge monitor. This channel continuously monitors processed liquid and laundry waste leaving the RF by means of a gamma scintillation detector mounted on a 3-inch discharge pipe. The unit is shielded to ensure the required detector sensitivity. A high-activity alarm automatically initiates closure of a valve that terminates discharges from the RF.

11.3.3.8 Condenser Air Ejector Monitors There are two identical radiation detection channels (1-SV-RM-111 and 2-SV-RM-211 for Units 1 and 2, respectively) for continuously monitoring the normal gaseous effluent from the condenser air ejectors to the atmosphere. The detectors are gamma scintilators mounted in an in-line sampler surrounded by lead shielding. Activity is indicative of a primary-to-secondary system leak. On a high-activity alarm, the flow is automatically diverted to the containment.

If a condition exists such that the normal radiation monitor alarms, but the containment is under Phase 1 isolation (Section 5.2.2), isolation valves 1-SV-TV-102, 102A, and 103 (Unit 1) and 2-SV-TV-202, 202A, and 203 (Unit 2) would shut (Reference Drawing 1) to stop all flow from the air ejector. However, the operator can maintain condenser vacuum by manually establishing air ejector flow through a discharge line to a point upstream of the ventilation vent no. 2 high-range noble gas effluent radiation monitor (Section 11.3.3.14). This monitor provides both normal and high-range effluent radiation detection. If ultimately required, the air ejector effluent can be isolated remotely from the control room.

11.3.3.9 Steam Generator Blowdown Sample Monitors Each of these channels (two channels per unit) monitors the liquid phase of the steam generators for radioactivity indicative of a primary-to-secondary system leak. The three steam generator blowdowns are combined and continuously monitored by the detectors. Upon indication of radioactivity, a valving arrangement enables the steam generators to be individually sampled, in turn, to determine the source of the activity. Once it has been established which steam generator is leaking, one of the detectors monitors only the blowdown from that steam generator, while the other detector monitors the combined blowdown from the other two steam generators.

The detectors are gamma scintillation detectors, mounted in liquid samplers surrounded by lead shielding.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-13 11.3.3.10 Recirculation Spray Cooler Service Water Outlet Monitors The recirculation spray coolers, as part of the recirculation spray system (Section 6.3.1),

operate only when containment pressure increases to the Hi Hi Consequence Limiting Safeguards (CLS) setpoint.

There are four recirculation spray coolers per unit, and each service water outlet line from the coolers is monitored, thus giving a total of eight channels. Each of these channels is identical.

If the recirculation spray system is placed in service, a 5-gpm to 10-gpm sample is drawn out of each service water outlet line by a small pump with a 2-hp motor and passed through an offline liquid sampler, where it is monitored for activity indicative of a leak in the respective recirculation spray cooler. After passing through the liquid sampler, which is located outside the containment, the sample is returned to the service water line. Each monitor consists of a gamma scintillation detector mounted in a standard offline sampler surrounded by 3 inches of lead.

To ensure low background radiation in the event of an accident, the Unit 1 monitors are located in the Unit 2 safeguards building and the Unit 2 monitors are located in the Unit 1 safeguards building.

11.3.3.11 Reactor Coolant Letdown Gross Activity Monitors Each of the units has its reactor coolant continuously monitored by means of a sample taken from the letdown line to the CVCS (Section 9.1). In this system, large variations in activity level are possible in the event of fuel assembly failure. This is a two-stage monitoring system consisting of a low-range channel and a high-range channel. There is one such system for each unit. After being withdrawn from the letdown line, the sample is passed through a delay line to allow N-16 to decay, then enters a sampler consisting of two gamma scintillation detectors surrounded by lead shielding, and is finally discharged to the volume control tank. Both detectors sit on a 1/2-inch removable stainless steel tube, providing flow through the sampler. Shielded lead plugs are used to convert the two detectors into either high- or low-range letdown monitors.

Normally, the low-range detector will be sitting on the 1/2-inch tubing and the high-range detector will be sitting on the shielded lead plug. In the event of a fuel element failure, the activity released could be sufficient to raise the coolant activity level above 1.0 Ci/cm3 gross fission products.

This causes the high-range monitor to begin to indicate activity level at 10-1 Ci/cm3, providing a one-decade overlap. At this point, the high-range channel provides the activity data, and the low-range monitor can be converted into a high-range monitor by inserting a shielded lead plug.

11.3.3.12 Circulating Water Discharge Tunnel Monitors Each of these identical channels (one per unit) monitors the effluent (service water, condenser circulating water, and liquid waste) in the circulating water discharge tunnel beyond the last point of possible radioactive material addition. A gamma scintillation detector slides into a capped pipe, which is then inserted directly into the discharge tunnel and acts as a well. At the top of the pipe is a waterproof support assembly that encloses a check source. The entire device is waterproof.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-14 11.3.3.13 Ventilation Vent No. 2 Particulate and Gas Monitors The two channels in Ventilation Stack No. 2 continuously sample for particulate and gas in the same way that the two Process Vent channels monitor the process vent sample, except that multi-probe samplers are provided to obtain a representative sample in the duct and both channels are equipped with silicon diode type detectors. In addition the post-accident noble gas and particulate monitor is used to obtain grab samples of the ventilation flow stream. Finally, the operability of the ventilation vent No. 2 particulate and gas monitors is relied upon in conjunction with the fuel pit bridge area monitor and communications to provide a timely and valid indication of a fuel handling accident in the spent fuel pool.

11.3.3.14 High-Range Postaccident Radiation Monitors Methods for monitoring high-level releases of noble gases, iodine, and particulates have been developed and implemented. All potential releases are monitored by instrumenting ventilation vent no. 2, the process vent stack, main steam safety valve and power operated relief valve header, and the auxiliary feedwater pump turbine exhaust. The waste gas decay tank and hydrogen purge exhaust are discharged through the process vent stack. The auxiliary building, decontamination building, fuel building, and safeguards area exhausts are discharged through the auxiliary building ventilation vent no. 2. The containment purge system, which is common to both units, discharges through the ventilation vent no. 2 during outages. The main condenser air ejectors normally discharge to the atmosphere, but flow is diverted to containment if the set radiation level limit is exceeded.

The high-range noble gas radiation monitors on the ventilation and process vents are listed in Table 11.3-8. These monitors have a range of 10-7 to 105 Ci/cm3 (Xe-133) under normal background conditions (less than 1 mR/hr). Due to shielding around the detector, the reduction of effluent detector sensitivity under maximum background conditions will not exceed the normal effluent instrument range. A multidetector system with detectors enclosed in lead shielding and sufficient range overlap is provided to ensure complete coverage for all expected background conditions. Shielded effluent detectors are needed to obtain the required sensitivities.

Accident particulate and iodine releases are determined by retrieving fixed filters for laboratory analysis. The filters are shielded to provide personnel protection during removal and reinstallation. Several filters in parallel provide for continuous sampling during filter removal.

High-range monitors, with a usable range of 0.01 mrem/hr to 10,000 R/hr, are installed on all main steam lines and the exhaust of the turbine driven AFW pump. These monitors are listed in Table 11.3-8. The 10,000 R/hr maximum reading corresponds to a noble gas concentration which is significantly greater than the maximum anticipated value, as determined by analysis, for noble gas concentrations following any design basis accident. The main steam monitors will be used in conjunction with secondary system sampling and offsite radiation monitoring.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-15 The control units for the monitors contain the electronics necessary to interpret and display detector readings. Four control units, one for each monitor, are located in the Unit 1 Containment Spray Pump House and four control units are located in the Unit 2 Containment Spray Pump House. The units provide visual alarms for failure, alert, and high radiation. A digital readout of radiation level is also provided.

The detectors are powered from the control units, called LPDUs. The LPDUs are powered from reliable sources.

The Surry Power Station has the capability of monitoring all vital areas through the process vent and ventilation vent stack samples or local grab samples, as appropriate. Accurate monitoring of iodine in the presence of high noble gas concentrations is accomplished by the use of silver zeolite sampling cartridges. The cartridges can be analyzed by the multichannel analyzer system (MCA) that is located in a concrete walled count room for shielding purposes. The MCA detectors are located within lead-lined shields that also contribute to background reduction. This combination of shielding provided in the count room is adequate to provide a low background analysis location under most emergency conditions. If background becomes too high for accurate analysis, samples can be taken to the Radwaste Facility or shipped offsite. Procedures for iodine sampling and analysis are available in the Health Physics office.

Particulate sampling is accomplished in conjunction with radioiodine sampling by using fiber filter patches positioned upstream of the iodine filter. Particulate analysis is accomplished by using a multichannel analyzer.

11.3.3.15 Storm Drain Radiation Sample System The storm drain radiation sample system is used to monitor for radioactive contamination of the effluent from the storm drain system prior to discharge into the James River. Recording flow meters and automatic wastewater samplers are installed at the four final release points of the storm drain system. Access to the equipment is through precast equipment manholes into the buried storm water drainage lines adjacent to the discharge canal. Electrical service to the equipment is provided by 120V ac power outlets in weatherproof enclosures inside the manholes.

Samples are automatically drawn on a periodic basis to be sampled by Health Physics personnel for radioactive material content.

11.3.4 Area Radiation Monitoring System 11.3.4.1 General The area radiation monitors are designed for continuous operation. Continuous, as used to describe the operation of an area radiation monitor, means that the monitor provides the required information at all times with the following exceptions: (1) the monitor is not required to be in operation because of specified plant conditions per the Technical Specifications, or (2) the system is out of service for testing or maintenance.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-16 The area radiation monitoring system reads out and records the radiation levels in selected areas throughout the station and activates alarms (audible and visible) if these levels exceed a preset value or if the detector malfunctions. With the exception of the containment gas and particulate monitors, each detector reads out and activates alarms, both in the control room and at its station location. Each channel is equipped with a check source remotely operated from the control room. The recorders provide a continuous permanent record of radiation levels while the detectors are functioning. Detectors monitoring Unit 1 are supplied with power from the emergency bus for Unit 1. Detectors monitoring Unit 2 are supplied with power from the emergency bus for Unit 2. Detectors monitoring areas common to both units have the capability of being supplied with power from the emergency bus for either Unit 1 or Unit 2.

Additions subsequent to the original station area radiation monitoring system include the spent resin handling area, laundry, and radwaste facilities. These systems are powered from reliable power supplies and are indicated and source tested locally.

The alarm setpoint of each area monitor is variable and is set at a level slightly above the normal background radiation level in the respective area.

The area radiation monitoring system consists of the detectors listed in Table 11.3-7, plus each units containment particulate and gas monitors, described below.

Criticality monitors are not required in the spent fuel and new fuel storage and handling areas. An exemption from the criticality monitoring requirements specified in 10 CFR 70.24(a) was received from the NRC for the storage and handling of fuel assemblies enriched up to 4.3 weight percent U-235 (Reference 8). The exemption was based on station design features and procedural controls that are in place to preclude an inadvertent criticality. Area radiation monitors are provided in these areas which would alert personnel to excessive radiation levels and would initiate appropriate response actions.

11.3.4.2 Containment Particulate Monitors This channel continuously withdraws a sample from the containment atmosphere into a closed, shielded system exterior to the containment. The sample is passed through a moving filter paper with a collection efficiency of 99% for particle sizes greater than 1.0 micron. The amount of deposited activity is continuously scanned by a lead-shielded beta scintillation detector with a sensitivity of 1 x 10-11 Ci/cm3 for particulates in a background of 2.5 mR/hr. The sample system, which is common to both the particulate and gas monitors, includes a pump with a 0.75-hp motor, a flow meter, automatic pressure protecting valves, a flow regulating valve, and isolation valves. The pump and motor are located outside the containment. A sample point is available for taking a sample of the containment atmosphere after an incident for spectrum analysis in the laboratory. During refueling a high-activity alarm automatically trips the containment purge air supply fans and closes the purge system butterfly valves, thus isolating the purge system. This automatic function is not credited in the fuel handling accident nor is it required to be functional. The operability of the containment particulate monitor is relied upon in

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-17 conjunction with the containment gas monitor, the manipulator crane area monitor, and communications to provide a timely and valid indication of a fuel handling accident in the containment. The counting rate of the limiting isotopes, I-131 and Cs-137, is nominally 2.6 x 1012 cpm/Ci/cm3 (2 scfm and 1"/hour tape speed).

11.3.4.3 Containment Gas Monitors This channel takes the continuous containment atmosphere sample, after it has passed through the particulate filter paper, and draws it through an in-line, easily removable, charcoal cartridge arrangement to the containment gas monitor assembly, which is a fixed-volume, lead-shielded sampler enclosing a beta scintillation detector. The sensitivity of this detector is 1 x 10-6 Ci/cm3 for noble gases in a background of 2.5 mR/hr. The sample activity is measured, and then the sample is returned to the containment.

During refueling, a high-activity alarm automatically trips the containment purge air supply fans and closes the purge system butterfly valves, thus isolating the purge system. This automatic function is not credited in the fuel handling accident nor is it required to be functional. The operability of the containment gas monitor is relied upon in conjunction with the containment particulate monitor, the manipulator crane area monitor, and communications to provide a timely and valid indication of a fuel handling accident in the containment.

A purge valve arrangement blocks the normal sample flow to permit purging the detector with a clean sample for calibration. Purged gases are discharged to the containment. Protection and isolation are provided as described in Section 11.3.4.2.

The counting rates of the limiting isotopes Xe-133, and Kr-85 are nominally 3.145 x 107, and 1.09 x 108 cpm/Ci/cm3, respectively.

11.3.4.4 Other Area Radiation Monitoring Equipment This equipment consists of fixed-position, ion-chamber-type gamma detectors and associated electronic equipment. These channels warn personnel of any increase in radiation level at locations where personnel may be expected to remain for extended periods of time. The channels and their ranges are listed in Table 11.3-7.

In addition, if the dose rate at the manipulator crane area monitor exceeds the alarm setpoint during refueling, the alarm automatically trips the containments purge air supply fans and closes the purge system butterfly valves, thus isolating the purge system. This automatic function is not credited in the fuel handling accident nor is it required to be functional. The operability of the manipulator crane area monitor is relied upon in conjunction with the containment gas and particulate monitors, and communications to provide a timely and valid indication of a fuel handling accident in the containment.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-18 Finally, the operability of the fuel pit bridge area monitor is relied upon in conjunction with the ventilation vent No. 2 particulate and gas monitors and communications to provide a timely and valid indication of a fuel handling accident in the spent fuel pool.

11.3.4.5 High-Range Postaccident Containment Monitors A high-range reactor containment area monitor is located outside the containment structure.

The detector is permanently mounted and aimed at the personnel hatch. The monitor has a range of 0.1 to 107 mR/hr to measure the expected high gamma dose rate in the containment following a LOCA.

An additional set of two high-range containment radiation monitors is installed at separate locations on the crane wall above the operating deck level inside containment. The monitors are single ion chamber detectors that measure photons over the range of 1 to 107 R/hr. The system is sensitive to photon energies from 60 keV to 3 MeV, with +/- 20% accuracy for 0.1 to 3 MeV photons.

The readout for the in-containment monitors is located in the control room and consists of a rate meter and recorder that starts at a present value. Each redundant monitor is powered by a separate vital instrument bus.

The in-containment monitors meet the Seismic Class 1 requirements of Regulatory Guide 1.100 (Reference 5), will withstand the LOCA conditions specified by Regulatory Guide 1.89 (Reference 6), and have been environmentally qualified in accordance with the requirements of IEEE 323-1974 and Regulatory Guide 1.97, Revision 3.

11.3.4.6 Technical Support Center Area Monitor The Technical Support Center (TSC) radiation monitoring system is a localized system and satisfies the guidelines established in NUREG-0696. The radiation monitoring system components consist of one particulate, iodine, and noble gas monitor, two area monitors and a remote alarm panel.

This monitoring system provides continuous indication of the dose rate and airborne activity in the TSC during an emergency, as well as alerting personnel of adverse conditions. It is totally contained within the TSC and is in no way connected to the control room or any safety-related systems.

11.3.5 Environmental Survey Program The information in this section gives the programmatic elements of the Radiological Environmental Monitoring Program (REMP) for Surry Power Station. Historical data is provided on the pre-operational radiological surveillance performed in support of the Applicants Environmental Report. Current requirements of the REMP program contained in Technical Specifications are implemented through the Offsite Dose Calculation Manual (ODCM), which

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-19 describes the specific elements of the present radiological environmental surveillance program at Surry Power Station.

The Surry Power Station Applicants Environmental Report contains the preoperational radiological surveillance program covering the period from May 1968 through June 1970.

Comments made by the U.S. Fish and Wildlife Service concerning the preoperational phase of the station were taken into account by:

1. Holding a conference on June 21, 1968, to discuss the pre- and postoperational surveys for the Surry Power Station. Representatives from the following agencies were present:

a. Federal Water Pollution Control Authority.

b. Bureau of Commission of Fisheries, Radiobiological Laboratory.

c. Virginia State Water Control Board.

d. Commonwealth of Virginia, Bureau of Industrial Hygiene.

e. Virginia Commission of Game and Inland Fisheries.

f. U.S. Fish and Wildlife Service, Bureau of Sport Fisheries and Wildlife.

g. Virginia Institute of Marine Science.

The meeting adjourned with the understanding that all agencies represented at the meeting were completely satisfied with the Companys program for pre- and postoperational radiological and ecological surveillance programs.

2. Conducting the preoperational radiological surveillance program in such a manner that indigenous species that concentrate radionuclides are routinely sampled and analyzed.

3. Preparing reports on the radiological program for distribution to interested agencies before station operation.

4. Holding discussions with appropriate state officials and personnel from the U.S. Fish and Wildlife Service regarding thermal pollution.

5. Installing of seven platforms in the James River upon which are seven instruments continuously measuring water temperature. (Instruments to measure river salinity were also installed on those same towers.)

6. Working in conjunction with personnel from the Virginia Institute of Marine Science regarding their grant from the AEC concerning thermal pollution of the river.

7. Designing the intake water facility with a 1.0-fps intake velocity at the screen surface so as to prevent significant damage to fishery resources.

A postoperational radiation surveillance program was developed using the knowledge and information obtained from the preoperational surveillance program. The latter was in effect over a

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-20 2-year period and served to train plant personnel in sampling and analytical techniques; aid in identifying those indicator samples that may be an indication of a slow buildup of radioactivity in the environment; establish the degree of variability between measurements resulting from seasonal changes in the weather (since fluctuations do occur and are expected); generate meaningful environmental data based primarily on scientific and technical requirements; and establish a correlation of data between the consulting service and the stations laboratory group.

The ultimate objective of the postoperational surveillance program is the verification of the adequacy of radiation source control. Therefore, analytical efforts are directed toward those samples that have the ability to concentrate the radioelements of concern and afford an integrated and sensitive sampling mechanism. Milk, shellfish (oysters, crabs, and clams), and silt are considered indicator samples and are indicative of radioactivity levels in the environment.

Commercial and/or recreationally important species of fish (catfish, eel and perch) are selected for sampling in the vicinity of the discharge point. The consumption of fish present a direct ingestion pathway to man. Because certain species of fish, such as bottom feeders, concentrate radionuclides which may be taken up from the water and aquatic sediments or may accumulate in the fish directly via internal deposition, it is important to include these organisms in the radiological environment sampling program. Samples are collected from the environment surrounding the station at various intervals throughout the year. Radioanalysis of these samples indicates conditions both in time and space, and thus a slow buildup of radioactivity can be determined. In addition, comparison in the trends of radioactivity levels are more meaningful, since the biological variation caused by using many different media has been eliminated.

Since the programs origination, sampling has been expanded to include obtaining representative samples of the sediment, shoreline silt, food products, oysters, fish and clams from the surrounding area. These samples are analyzed for radioisotopes.

11.3.5.1 Air Monitoring To meet the surveillance objective previously stated, it is desirable for the surveillance methods to indicate changes in radioactivity above-background levels. Therefore, continuous-duty air particulate samplers are also used to confirm that the station presents no hazard to the public. An air-sampling station is located at different locations including a control location which provides background data. The establishment of the air particulate network takes into account the following four general considerations:

1. The average meteorological conditions in the vicinity of the site.

2. The current and projected population densities near the station.

3. The proximity of other nuclear facilities.

4. The proximity of the site boundary location of the highest calculated annual average ground level D/Q.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-21 Air particulate matter is accumulated for a 1-week period on appropriate filter media using a low-volume air sampler. The particulate filters are analyzed in accordance with Radiological Environmental Monitoring Program requirements.

11.3.5.2 Milk The dose consequence to an individual is from both a direct and indirect exposure pathway.

The direct exposure pathway is from the inhalation of radioactive material and the indirect exposure pathway is from the grass-cow-milk pathway. In this pathway radioactive material is deposited on the plants consumed by the dairy animals. The radioactive material is them passed on to the individual via the milk.

It has been estimated that, since a cow grazes over an area of 160 m2 per day (Reference 7),

cows milk affords a good integrated sample. Since milk is one of the best and most direct biosamplers for determining the radiocesium, radiostrontium, and radioiodine levels in the environment, samples are collected from the local dairy farms in the vicinity of the station.

11.3.5.3 Shellfish, Crabs, and Fish Shellfish have the ability to concentrate certain stable elements and radionuclides far above the normal concentrations found in their saline water environment. Oysters and clams, Crasostrea virginica and Mercenaria mercenaria respectively, found in the James River are thus one of the more sensitive mechanisms for the determination of radioactivity released from the station.

Oysters and clams are collected in the vicinity of the station. Crab and fish samples are obtained in the vicinity of the station. Radioanalysis of these sample media are in accordance with the Radiological Environmental Monitoring Program.

Additional aquatic vectors useful as integrating samples have been investigated, but none have proven to be as sensitive or indicative as shellfish.

11.3.5.4 Silt and Sediment Since shellfish do not concentrate all radionuclides, river bottom sediment samples are collected from downstream areas which have or potentially may have recreational value. Silt samples are obtained in the vicinity of the station. Because of the interaction of a number of mechanisms, radionuclides accumulate in silt and bottom sediments. Because of this, silt is one of the few environmental media in which radioactive effluents from nuclear power stations are usually detected. These samples thus afford an integrated indication of average water concentrations.

11.3.5.5 Water Samples Water samples are collected for analysis from the following sources: surface water samples from two different locationsone upstream and one downstream from the station and ground (well) water samples, which are collected from various wells in the vicinity of the station. The

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-22 well water and surface water samples are analyzed in accordance with Radiological Environmental Monitoring Program requirements.

11.3.5.6 Food Products The dose consequence to an individual from food products is via the ingestion exposure pathway. Samples of food products are obtained from farms in the vicinity of the station and analyzed in accordance with Radiological Environmental Monitoring Program requirements.

11.3.5.7 Equipment Analytical equipment used for the Radiological Environmental Monitoring Program includes the following:

1. Gas-flow proportional counter (alpha-beta counter).

2. Multichannel gamma spectroscopy analyzer.

3. Liquid scintillation counter.

4. Thermoluminescent Dosimeter (TLD) readers.

5. Other analytical equipment which meet the Lower Limit of Detection (LLD) requirements set forth in the Offsite Dose Calculation Manual.

11.3.5.8 Environmental Dosimetry Normally, the gaseous wastes discharged from the station consist almost entirely of the noble gases, xenon and krypton. The radiation hazard from these gases, due to their inertness, is external radiation exposure. Therefore, radiation surveillance can be maintained by using devices to measure total external body radiation levels in the station environs. The TLDs measure external radiation exposure from several sources including naturally occurring radionuclides in the air, radiation from cosmic origin, fallout from atomic weapons testing as well as potential radioactive airborne releases from Surry Power Station. An inner ring of TLDs are located in the five mile range from the site with a station in each of the 16 sectors of each ring. Other TLDs are located in special program interest areas (residences, schools, etc.) and in control locations. The TLDs are collected and analyzed for gamma radiation at a frequency which optimizes statistical sensitivity and characterizes seasonal fluctuations.

11.3.5.9 Interlaboratory Comparison Program Surry Power Station participates in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring. The program is applicable for the radioanalyses performed in support of the Radiological Environmental Monitoring Program.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-23 11.3.6 Control Room The control area is described in Section 7.7. The design basis for radiation protection in the control room under accident conditions is that the whole body radiation dose to personnel accessing and occupying the control room is limited to less than or equal to 5 rem TEDE for the duration of the accident. This dose includes the 30-day dose from ECCS leakage (including RWST backleakage) and the contribution from the postulated radioactive plume leaking from the containment (as discussed in Section 14.5.5) until engineered safeguards equipment returns the containment to subatmospheric pressure and terminates the leakage.

The control room doses for the design basis LOCA are discussed in Section 14.5.5.3.

The control room walls, which are a minimum of 24-inch thick for tornado missile protection, provide more than adequate shielding from radiation. Special consideration has been given to the design of penetrations and structural details of the control room so as to establish an acceptable condition of leaktightness.

The air conditioning systems are installed within the spaces served and are designed to provide uninterrupted service under accident conditions. Upon a safety injection signal, the normal replenishment air and exhaust systems are isolated automatically from the control room by tight closures in the ductwork. The control and relay rooms are provided with an emergency ventilation system fitted with particulate and impregnated charcoal filters to introduce cleaned outside air into the protected spaces. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of control room envelope isolation, procedures require the alignment of the control room emergency ventilation system to provide a filtered breathing air supply to the control room envelope. This can continue indefinitely to hold the area pressure above atmospheric to ensure outflow leakage.

The radiation level in the control room is measured by a gamma monitor to verify safe operating conditions.

As a secondary precaution, personnel air-packs are available in the control area.

11.3 REFERENCES

1. IE Bulletin 79-01, Environmental Qualification of Class 1E Equipment, 1979.

2. J. J. Dinunno et al., Calculation of Distance Factors for Power and Test Reactors, TID -

14844, U.S. Atomic Energy Commission, 1962.

3. Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors, 1974.

4. NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, 1979.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-24

5. Regulatory Guide 1.100, Seismic Qualification of Electric Equipment for Nuclear Power Plants, 1977.

6. Regulatory Guide 1.89, Qualification of Class 1E Equipment for Nuclear Power Plants, 1974.

7. G. Wortley, Contamination of Milk with Radionuclides Dispersed into the Biosphere, a report submitted to the Joint FAO/WHO Expert Committee on Milk Hygiene, WHO Headquarters, Geneva, Switzerland, 1969.

8. Letter from G. Edison of the USNRC to J.P. OHanlon of Virginia Electric and Power Company dated July 15, 1998 (Serial No.98-427), Issuance of Revised Exemption from the Requirements of 10 CFR 70.24(a) - Surry Power Station, Units 1 and 2 (TAC Nos. MA0657 and MA0658).

9. Letter from NRC to Vepco,

Subject:

Steam Generator Program, Surry Power Station, Units 1 and 2 dated August 17, 1977, Ser. No. 351.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-25 11.3 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11448-FM-066A Flow/Valve Operating Numbers Diagram: Auxiliary Steam and Air Removal System, Unit 1 11548-FM-066A Flow/Valve Operating Numbers Diagram: Auxiliary Steam and Air Removal System, Unit 2

2. 11448-FM-090B Flow/Valve Operating Numbers Diagram: Gaseous Waste Disposal System, Unit 1

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-26 Table 11.3-1 ALLOWABLE DOSE RATES Maximum Dose Rate Zone Description (mrem/hr) Typical Locations Full-power operation Control room, outside surface of containment, Continuous access (Zone I) 0.75 and all turbine plant and administration areas Auxiliary and fuel building passageways in general and inside reactor containment Periodic access (Zone II) 2.5 personnel lock Limited access (Zone III) 15 Outside surface of shielded tank shields Controlled access (Zone IV) 15 Inside shielded equipment compartments Annulus between crane wall and containment Access to incore instrumentation 100 wall Vicinity of incore instrumentation transfer Access to incore instrumentation 40 devices Access to incore instrumentation 20 Vicinity of incore instrumentation motors Hot shutdown (after 15-min decay)

Reactor containment above charging floor and Limited access (Zone III) 15 outside of crane wall Controlled access (Zone IV) 15 Inside shielded equipment compartments Cold shutdown for maintenance (after 8-hr decay)

Reactor containment above charging floor and Periodic access (Zone II) 2.5 outside of crane wall Controlled access (Zone IV) 15 Inside shielded equipment compartments Cold shutdown for refueling (after 4-day decay)

Reactor containment above charging floor, outside of crane wall, and adjacent to fuel transfer canal near incore instrumentation Periodic access (Zone II) 2.5 devices Controlled access (Zone IV) 15 Inside shielded equipment compartments Surface of water over raised fuel Fuel element above up-ender, and above other elements 50 fuel elements in fuel building

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-27 Table 11.3-2 CONTAINMENT SHIELDING

SUMMARY

Thickness Symbol Figure Shield Description Materiala (inches)

Water 34 A 11.3-2 Neutron shield tank Steel 3 B 11.3-2 Primary shield Concrete 54 C 11.3-2 Supplementary neutron shield Benelex 70 14 D 11.3-2 Streaming shield ring Benelex 401 5 Steel 1-1/2 E 11.3-2 Neutron shield tank support Lead 2 11.3-1 &

F 11.3-2 Cubicle - crane support wall Concrete 33 G 11.3-2 Crane support wall Concrete 24 11.3-1 &

H 11.3-2 Containment wall Concrete 54 I 11.3-2 Containment dome Concrete 30 J 11.3-2 Floor elevation, 3' 6" Concrete 24 K 11.3-2 Charging floor Concrete 24 11.3-1 &

L 11.3-2 Refueling cavity wall Concrete 36 Concrete 11.3-2 (Unit 1) 24 Steel M 11.3-3 Missile shield (Unit 2) 2 N 11.3-2 Refueling cavity water Water -

Concrete O 11.3-2 Removable block wall (Ilmenite) 12 P 11.3-1 Fuel trans. canal wall Concrete 54 Q 11.3-1 Fuel trans. canal wall Concrete 72 R 11.3-1 Fuel trans. tube shielding Concrete 36 S 11.3-1 Fuel trans. canal wall Concrete 72 T 11.3-1 Incore inst. cubicle wall Concrete 42 U 11.3-1 Cubicle wall Concrete 30 V 11.3-1 Regen. heat exchanger wall Concrete 24 W 11.3-1 Cable vault wall Concrete 24 X 11.3-1 Auxiliary feed pump cubicle wall Concrete 36 Y 11.3-1 Safeguards area wall Concrete 24 Z 11.3-3 Incore Sump Room Drain Steel 3.5 AA 11.3-2 Incore Sump Room Drain Steel 3.5

a. All concrete is reinforced with steel.

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-28 Table 11.3-3 N-16 AND ACTIVATED CORROSION PRODUCT ACTIVITY USED IN THE ORIGINAL PLANT SHIELDING DESIGN ANALYSIS Activity Isotope (Ci/cm3 at 500°F)

Mn-54 2.7 x 10-3 Mn-56 5.7 x 10-2 Fe-59 8.3 x 10-3 Co-58 2.3 x 10-4 Co-60 9.2 x 10-4 N-16a 64.0

a. At the reactor vessel outlet nozzle at 2546 MWt (power level for the original plant shielding design analysis).

Table 11.3-4 POSTACCIDENT MITIGATION PHASE SOURCE TERMS Source Source Term Basis Sump water 0% noble gas Sump water is degassed 50% halogens TID-14844 1% solid fission products TID-14844 Primary coolant sample 100% noble gas R.G. 1.4/TID-14844 50% halogens TID-14844 1% solid fission products TID-14844 Containment atmosphere 100% noble gas R.G. 1.4/TID-14844 25% halogens R.G. 1.4 0% solid fission products R.G. 1.4

Revision 52Updated Online 09/30/20 Table 11.3-5 PROCESS RADIATION MONITORING SYSTEMa Maximum Type of Sensitivity Range Background Monitor Number Detector Medium Limiting Isotopes (ci/cm3) (decades) (mR/hr)

Process Vent Particulate Silicon (1-GW-RM-130A) 1 Diode Gas, air 1-131 1 x 10-9 (I-131) 3 0.75 Process Vent Gas Silicon (1-GW-RM-130B) 1 Diode Gas, air I-131, Xe-133, Kr-85 5 x 10-6 (Kr-85) 4 0.75 Ventilation Vent No. 1 - Gas (1-VG-RM-104) 1 Beta scint. Air 1-131. Xe-133, Kr-85 5 x 10-6 (Kr-85) 3 0.75 RF Vent Particulatec 1.92E-12 (1-RRM-RE-100) 1 Beta scint. Gas, air Cs-137 (Cs-137) 5 <1 RF Vent Gasc (1-RRM-RE-101) 1 Beta scint. Gas, air Xe-133 5.12E-7 (Xe-133) 5 <1 Component Cooling Water Gamma (1-CC-RM-105, 106) 2 scint. Water Co-60, Cs-137 1 x 10-5 (Cs-137) 3 0.75 Comp. Cooling Hx Service Water Gamma SPS UFSAR (1-SW-RM-107A, B, C, D) 4 scint. Water Co-60, Cs-137 1 x 10-5 (Cs-137) 3 0.75 RF Liquid Waste Disposalb Gamma (1-RRM-RE-131) 1 scint. Water CS-137 1 x 10-7 (Cs-137) 5 <1 Condenser Air Ejector Gamma 1.74 x 10-5 (1/2-SV-RM-111/211) 2 scint. Vapor I-131, Xe-133, Kr-85 (Kr-85) 5 0.75 Steam Generator Blowdownb (1/2-SS-RM-112/212 Gamma

&113/213) 4 scint. Water Co-60, Cs-137 4 x 10-6 (Cs-137) 3 2.5

a. This table contains information based on the requirements stated in the various specifications for the listed monitors. The actual monitors meet or exceed these minimum requirements except 1-RRM-RE-100 and 1-RRM-RE-101.

11.3-29

b. These channels also have local readout and alarm.

c. Based upon Sensitivity Calculated IAWX-S-001376-001, Revision 0, Specification for Radiation Monitoring System.

Table 11.3-5 (CONTINUED)

Revision 52Updated Online 09/30/20 PROCESS RADIATION MONITORING SYSTEMa Maximum Type of Sensitivity Range Background Monitor Number Detector Medium Limiting Isotopes (ci/cm3) (decades) (mR/hr)

Recirculation Spray Cooler (1/2-SW-RM-114/214, Gamma 115/215, 116/216, & 117/217) 8 scint. Water Co-60, Cs-137 3 x 10-4 (Cs-137) 2 5.0 R.C. Letdown High Rangeb Gamma Co-60, mixed fission (1/2-CH-RM-118/218) 2 scint. Water products 1 x 10-1 (Co-60) 4 2.5 R.C. Letdown Low Rangeb Gamma Co-60, mixed fission (1/2-CH-RM-119/219) 2 scint. Water products 1 x 10-4 (Co-60) 4 2.5 C.W. Discharge Tunnel Gamma (1/2-SW-RM-120/220) 2 scint. Water Co-60, Cs-137 2 x 10-7 (Cs-137) 3 0.05 Ventilation Vent No. 2 -

Particulate Silicon (1-VG-RM-131A) 1 Diode Air I-131 1 x 10-9 (I-131) 3 0.75 Ventilation Vent No. 2 - Gas Silicon (1-VG-RM-131B) 1 Diode Air I-131, Xe-133, Kr-85 5 x 10-6 (Kr-85) 3 0.75 SPS UFSAR Laundry Facility Continuous Effluent -

Particulate (1-RM-RMS-RIC3) 1 Beta scint. Air I-131, Cs-137 1 x 10-10 (I-131) 5 0.5 Continuous Effluent - Iodine (1-RM-RMS-RIC4) 1 Beta scint. Air I-131 1 x 10-10 (I-131) 5 0.5

a. This table contains information based on the requirements stated in the various specifications for the listed monitors. The actual monitors meet or exceed these minimum requirements except 1-RRM-RE-100 and 1-RRM-RE-101.

b. These channels also have local readout and alarm.

c. Based upon Sensitivity Calculated IAWX-S-001376-001, Revision 0, Specification for Radiation Monitoring System.

11.3-30

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-31 Table 11.3-6 PROCESS RADIATION MONITORING SYSTEM COUNTING RATES OF LIMITING ISOTOPES (cpm/Ci/cm3)

Monitor I-131 Xe-133 Kr-85 Cs-137 Co-60 Process Vent Particulate 1.04E+9 - - 1.15E+9 7.1E+8 Process Vent Gas 2.28E-6 1.7E-6 2.92E-6 - -

Ventilation Vent No. 1 - Gas - 2.2 x 107 9.5 x 107 - -

RF Vent Particulate - - - 6.41E10 -

RF Vent Gas - 3.35E7 9.40E7 - -

Ventilation Vent No. 2 -

Part. 1.04E+9 - - 1.15E+9 7.1E+8 Ventilation Vent No. 2 - Gas - 1.7E-6 2.92E-6 - -

Component Cooling Water - - - 1.0 x 108 2.0 x 108 Component Cooling HX Service Water - - - 2.8 x 108 7.6 x 108 RF Liquid Waste Disposal - - - 3.4 x 108 9.9 x 108 Condenser Air Ejector 1.79 x 109 2.32 x 107 3.22 x 106 - -

Steam Generator Blowdown - - - 1.0 x 108 2.0 x 108 Recirculating Spray Cooler - - - 1.0 x 108 2.0 x 108 R.C. Letdown High Range - - - - 9.9 x 102 R.C. Letdown Low Range - - - - 9.6 x 105 C.W. Discharge Tunnel - - - 4.2 x 108 1.1 x 109 Laundry Facility Continuous Effluent Part. 3.6 x 1010 - - 3.6 x 1010 -

Continuous Effluent Iodine 4.1 x 109 - - - -

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-32 Table 11.3-7 AREA RADIATION MONITORING LOCATIONS, NUMBER, AND RANGES Channel Location (Number) Range (mR/hr)

Containment personnel hatch area (2) 1/2-RM-RMS-161/261 10-1107 Manipulator crane (2) 1/2-RM-RMS-162/262 10-1107 Reactor containment area (2) 1/2-RM-RMS-163/263 10-1107 Incore instrument transfer area (2) 1/2-RM-RMS-164/264 10-1107 New fuel storage area (1) 1-RM-RMS-152 10-1107 Fuel pit bridge (1) 1-RM-RMS-153 10-1107 Auxiliary Building control area (1) 1-RM-RMS-154 10-1107 Solid waste drum storage and handling area (1) 1-RM-RMS-155 10-1107 Sample room (1) 1-RM-RMS-156 10-1107 Main control room (1) 1-RM-RMS-157 10-1107 Laboratory (1) 1-RM-RMS-158 10-1107 Decontamination area (1) 1-RM-RMS-151 10-1107 Spent resin handling area (2) 1-RM-RMS-138, 139 10-2103 Laundry Facility (2) 1-RM-RMS-RIC8, RIC9 10-2103 Radwaste Facility (10) 1-RRM-RE-121 (RF control room) 1-RRM-RE-122 (RF laboratory) 1-RRM-RE-123 (RF DAW truck area) 1-RRM-RE-124 (RF DAW sorting/compactor area) 1-RRM-RE-125 (RF LSA box storage area) 1-RRM-RE-126 (RF HIC storage and handling Area) 1-RRM-RE-127 (RF hot machine shop truck bay) 1-RRM-RE-128 (RF hot machine shop area) 1-RRM-RE-129 (RF local control panel area) 1-RRM-RE-130 (RF bitumen control room) 10-1104 Containment high range radiation monitor system (4) 1/2-RM-RMS-127/227 & 128/228 100107 R/hr Technical support center (2) 1-RM-RMS-136, 137 10-1104

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-33 Table 11.3-8 HIGH-RANGE POST-ACCIDENT RADIATION MONITORS Normal Range Noble Gas Effluent Monitors Process Vent (1-GW-RM-130B)

Ventilation Vent No. 2 (1-VG-RM-131B)

High Range Noble Gas Effluent Monitors Process Vent (1-GW-RM-130C)

Ventilation Vent No. 2 (1-VG-RM-131C)

High Range Effluent Monitors Main Steam Lines 1-MS-RM-124 (1A) 1-MS-RM-125 (1B) 1-MS-RM-126 (1C) 2-MS-RM-224 (2A) 2-MS-RM-225 (2B) 2-MS-RM-226 (2C)

Auxiliary Feedwater Turbine Exhaust 1-MS-RM-129 2-MS-RM-229

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-34 Figure 11.3-1 SHIELD ARRANGEMENT, PLAN

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-35 Figure 11.3-2 SHIELD ARRANGEMENT, ELEVATION, UNIT 1

Revision 52Updated Online 09/30/20 SPS UFSAR 11.3-36 Figure 11.3-3 SHIELD ARRANGEMENT, ELEVATION, UNIT 2

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-i Appendix 11A Radiation Exposure Evaluation for Expected Radioactive Effluents

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-ii Intentionally Blank

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-1 APPENDIX 11A RADIATION EXPOSURE EVALUATION FOR EXPECTED RADIOACTIVE EFFLUENTS 11A.1 ANALYTICAL BASIS Surry Units 1 and 2 were analyzed and evaluated using the parameters and methodology set forth in Regulatory Guides 1.109 (Reference 1), 1.111 (Reference 2), and 1.112 (Reference 3),

and NUREG-0017 (Reference 4). Maximum individual doses resulting from gaseous and liquid effluents were calculated.

Radioactive source terms, both liquid and gaseous, were calculated in a manner consistent with Regulatory Guide 1.112 and NUREG-0017. Specific data used for the generation of the sources terms are presented in Section 11A.2.1.

Meteorological information used in the calculation of doses was developed consistent with the methodology described in Regulatory Guide 1.111. Information related to the meteorological inputs is contained in Section 11A.2.2. The dispersion factors (/Q) and ground depositions factors (D/Q) from the release points at Surry Units 1 and 2 to the various receptions are contained in Section 11A.2.3.

A plant and animal census is provided in Section 11A.3.

Dose calculations were performed in a manner consistent with Regulatory Guide 1.109. The NRC computer codes LADTAP and GASPAR were used to perform the calculations. The results of the analyses presented in Section 11A.4 support the Surry Power Stations capability of keeping the levels of radioactivity in effluents as low as reasonably achievable.

The liquid waste disposal system described Section 11.2.3 reflects changes in the liquid waste design used in the original Appendix I evaluation. Table 11.2-4, however, maintains the parameters used in the evaluation. This table was unchanged because the new liquid waste design was to be, at a minimum, equal to the previous system. In reality, improved performance was expected. Therefore, this analysis represents a conservative evaluation of offsite doses.

The gaseous waste disposal system is as described in 11.2.5.

11A.2 INPUT INFORMATION General plant information, meteorological information, dispersion factors (/Q), and ground deposition factors (D/Q) are given in the following sections.

11A.2.1 General Plant Information Plant information required by Appendix B of Regulatory Guide 1.112 is contained in Table 11A-1.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-2 11A.2.2 Meteorological Information Information concerning the onsite meteorological measurements program is found in Section 2.2.1.2.

Joint frequency distributions of wind speed and wind direction by atmospheric stability class are prepared monthly and annually, based on the format of Table 1 in Regulatory Guide 1.23 (Reference 5). Monthly and annual joint frequency distributions of 35-foot wind and T150 ft. - 35 ft. data are used as input for /Q and D/Q calculations for ground-level releases.

Monthly and annual joint frequency distributions of 150-foot wind and T150 ft. - 35 ft. data are used as input for qualifying elevated release calculations of /Q and D/Q. The 2-year (1974-1976) data set was chosen as the most recent representative data set.

The meteorological data for the period 1974-1976 are considered to be representative of atmospheric transport and diffusion conditions of the site region on a long-term basis. The stability distribution based on T150 ft. - 35 ft. for the period March 3, 1974, to March 2, 1975, is consistent with the 2-year data period as used in this report (Table 11A-2). Comparison of annual wind roses for both the 35- and 150-foot levels indicates that the annual 2-year wind roses are consistent with the first year of data and are in general agreement with Richmond and Norfolk wind roses for the period January 1, 1969, to December 31, 1973 (Reference 6). The representativeness of the first year of the 2-year data set to the long-term meteorological conditions is discussed in Reference 6.

11A.2.3 Dispersion Factors and Ground Deposition Factors Table 11A-3 provides /Q and D/Q values for ground-level and mixed-mode releases for the special appropriate distances as indicated in Section 11A.3 for each downwind sector.

Tables 11A-4 and 11A-5 provide /Q and D/Q values associated with surface-level releases from the containment (considered as entrained in the building wake, and therefore a ground-level release) for the standard population distances. Tables 11A-6 and 11A-7 provide /Q and D/Q values associated with a process vent release from 3.048 m above one of the containment structures, or approximately 43 m above grade (considered as a mixed-mode release) for the standard population distances.

Dispersions factors ( /Q) were calculated using a sector-average, straight-line model specified in Regulatory Guide 1.111. Ground deposition (D/Q) values were calculated according to Regulatory Guide 1.111. The mixed release mode was used as applicable for a release height of 3.05 m above the 40.1-m containment from a 0.08-m-diameter vent at an exit velocity of 30.5 m/s. Qualifying elevation release heights were adjusted for momentum stack downwash, and terrain rise, as described in Regulatory Guide 1.111.

The open terrain correction factor for /Q and D/Q values was applied in accordance with Figure 2 of Regulatory Guide 1.111. As described in References 2 and 3 and in Table 11A-8, the terrain is flat and rises to less than about 170 feet out to a distance of 50 miles near Richmond.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-3 Therefore, straight-line airflow trajectory regimes are considered to reasonably represent dispersion conditions as related to annual /Q values in the vicinity of the Surry Power Station.

The calculation /Q and D/Q values were based on onsite meteorological data during the period March 3, 1974, to March 2, 1975, and May 1, 1975, to April 30, 1976. Representative joint frequency distributions were developed for ground-level or elevated releases from the station as follows:

1. Ground-level release /Q and D/Q calculations were based on meteorological tower observations of wind speed and direction at the 35-foot level and of temperature differential (delta T) between the 150- and 35-foot levels. These levels were selected to conservatively represent the transport and diffusion of surface releases in the vicinity of the plant, or for vent releases entrained in the building wake. The z diffusion parameter was based on the curves in Figure 1 of Regulatory Guide 1.111.

2. Qualifying elevated release /Q and D/Q calculations were based on meteorological tower observations of wind speed and direction at the 150-foot level and of the same temperature differential (delta T) between the 150- and 35-foot levels, as representing the environment of the plume between its release height and the ground.

11A.3 PLANT AND ANIMAL CENSUS The plant and animal census is conducted annually in order to determine the current land use of the area surrounding Surry Power Station. The purpose of the census is to locate the distance to the nearest milk cow or other bovine, milk goat, vegetable garden, and residence within 5 miles of the Surry Power Station. The annual land use is detailed in the current radiological environmental monitoring report.

11A.4 DOSE CALCULATIONS 11A.4.1 Doses From Liquid Effluents Liquid source terms were calculated for two specific cases using the GALE Code: the liquid radwaste system as presently operating and as the system operated with the blowdown treatment system, which is no longer used. These cases are indicated below:

1. Dirty wastes treated by a system consisting of two mixed-bed demineralizers and no treatment of steam generator blowdown.

2. Same as 1 above, only steam generator blowdown treated by two mixed-bed demineralizers.

(This equipment is no longer used.)

Inputs to the GALE Code were based on (1) station operating experience, (2) information supplied previously in Chapter 11 and the Environmental Report (ER) (References 8 & 9), and (3) NUREG-0017. Source terms for each of the two cases outlined above are presented in

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-4 Tables 11A-9 and 11A-10.

Liquid radioactive wastes from the units are released to the James River via the discharge canal. Possible pathways of exposure for release from the station include ingestion of fish and invertebrates and shoreline activities. The irrigated food pathway does not exist at this location, nor does the potable water pathway. For all pathways, a river dilution factor of 5 was assumed as appropriate per Regulatory Guide 1.109.

Doses from liquid pathways were calculated for the maximum individual, based on the models given in Regulatory Guide 1.109, using the computer code LADTAP. Dose factors, bioaccumulation factors, and shorewidth factors given in Regulatory Guide 1.109 and in the LADTAP code were used, as were usage terms for shoreline activities and ingestion of fish and invertebrates.

Tables 11A-11 and 11A-12 present the LADTAP input data and the maximum individual doses for both cases indicated above.

During normal station operations, doses from liquid effluents are calculated according to the Offsite Dose Calculation Manual (ODCM). Calculations from the ODCM demonstrate compliance with this section.

11A.4.2 Doses From Gaseous Effluents Gaseous source terms were calculated using the GALE Code, and are presented in Table 11A-13. Inputs to the GALE Code were based on (1) plant operating experience, (2) information supplied in Chapter 11 and the ER, and (3) NUREG-0017.

Doses to the maximum individual from gaseous effluents were calculated by the NRC GASPAR Code, using the models of Regulatory Guide 1.109. Dose factors, annual air intake, intakes of food products, and parameters for calculating radionuclide concentrations in food products as given in Regulatory Guide 1.109 and in the GASPAR Code were used.

Dose contributions from the following pathways were calculated and analyzed in the assessment of the maximally exposed individual:

1. Immersion in the plume.

2. Ground contamination.

3. Inhalation.

4. Consumption of vegetables, meat, and milk.

For dose calculation purposes, the source terms of Table 11A-2 were divided into mixed-mode releases (i.e., those released from the Surry process vent that could be considered to be elevated at certain times and at ground level at others) and ground-level releases (i.e., those released from the ventilation vents, steam generator flash tank, and turbine building). The sources

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-5 of releases for the Surry process vent are the gaseous waste and containment vacuum systems. For dose calculation purposes, these releases were considered mixed mode, and the /Q and D/Q values, as presented in Table 11A-3, reflect this.

The sources of releases from the ventilation vents of the Surry units are the auxiliary, decontamination, and spent fuel buildings, safeguards areas, condenser air ejector, and containment purge systems. For dose calculation purposes, these releases were considered ground level, and the /Q and D/Q values, as presented in Table 11A-3, reflect this.

Based on the /Q and D/Q values in Table 11A-3, specific locations were analyzed for the location of the maximally exposed individual. When the principal locations were determined, dose calculations were performed, incorporating the pathways specific to each location. By adding the doses resulting from both mixed-mode and ground-level releases for each of the pathways existing at these locations, the location of the maximum individual was determined.

After evaluating the special locations, the maximum organ dose occurred to an infant who resides 3.75 miles north-northwest of the power station and drinks milk from a cow raised at this location. The maximum total body dose occurred to an individual 1.53 miles south of the Surry Power Station. Table 11A-14 presents the doses at the location of the maximally exposed individuals for Surry Units 1 and 2.

Table 11A-15 presents the doses to the above-mentioned individual based on the cooling of steam generator blowdown below 212°F. Operation in this manner eliminates the gaseous releases of I-131 and I-133 from the blowdown vent offgas and results in a reduction of the maximum organ dose by a factor of 5.

During normal station operations, doses from gaseous effluents are calculated according to the Offsite Dose Calculation Manual (ODCM). Calculations from the ODCM demonstrate compliance with this section.

11A REFERENCES

1. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, 1976.

2. Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, 1976.

3. Regulatory Guide 1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors, 1976.

4. NUREG-0017, Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Pressurized Water Reactors (PWR-GALE Code), 1976.

5. Regulatory Guide 1.23, Onsite Meteorological Program (Safety Guide 23), 1972.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-6

6. Virginia Electric and Power Company, Surry 3 and 4 Annual Meteorological Report, Docket Nos. 50-434 and 40-435, 1975.

7. U.S. Atomic Energy Commission, Surry Power Station Units 3 and 4 Final Environmental Statement, Docket Nos. 50-434 and 50-435, 1974.

8. U.S. Atomic Energy Commission, Surry Power Station Unit 1 Final Environmental Statement, Docket No. 50-280, 1972.

9. U.S. Atomic Energy Commission, Surry Power Station Unit 2 Final Environmental Statement, Docket No. 50-281, 1972.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-7 Table 11A-1 GENERAL PLANT INFORMATION*

Units Value Source General Maximum core thermal power evaluated for safety considerations in SAR MWt 2441 Section 1.1 Quantity of liquid tritium GALE Code released Ci/yr 480 calculations Quantity of gaseous tritium GALE Code released Ci/yr 490 calculations Primary system Mass of coolant in primary Calculations based on system, excluding pressurizer information in and primary coolant purification Tables 4.1-2, 4.1-3, system, at full power 103 lb 367 4.1-4, and 4.1-6 Average primary system letdown rate to primary coolant purification system gpm 60 Table 9.1-3 Average flow rate through the primary coolant purification system demineralizer gpm 6 Table 9.1-5 Unit operating Average shim bleed flow gpm 1.8 experience Secondary system Number of steam generators 3 Table 4.1-4 Type of steam generators U-tube Table 4.1-4 Carryover factor used for 1% iodine evaluation of iodine and 0.1%

non-volatiles nonvolatiles NUREG-0017 Total steam flow in secondary system 106 lb/hr 11.2 Figure 10.2-1 Calculations based on Mass of liquid in each steam information in generator at full power 103 lb 90.7 Table 4.1-4 Primary-to-secondary system leakage rate used in evaluation lb/day 100 NUREG-0017 Average steam generator blowdown rate used in Unit operating evaluation total 103 lb/hr 30.3 experience

• This information was developed for the Appendix I Report, which was submitted to the NRC in Letter SN 247, dated June 17, 1977. This information was based on the original plant licensed power of 2441 MWt and has never been updated. Current compliance to 10 CFR 50 Appendix I requirements are documented through the Offsite Dose Calculation Manual and reported in accordance with Technical Specifications.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-8 Table 11A-2 MONTHLY T150 ft - 35 ft STABILITY DISTRIBUTION (%)

Month A B C D E F G January 1975a 2.13 2.63 2.79 36.29 41.22 7.88 7.06 2-yr 10.07 2.19 2.42 32.08 39.73 5.93 7.57 February 1975 7.54 3.28 4.10 35.57 31.97 9.02 8.52 2-yr 11.88 3.83 3.35 25.68 34.53 9.65 11.08 March 1974-1975 20.68 4.46 5.36 32.29 28.12 4.02 5.06 2-yr 17.04 4.12 4.12 29.75 32.42 4.77 7.80 April 1974 14.92 2.44 4.26 21.77 40.79 8.37 7.46 2-yr 20.35 2.59 3.50 18.18 33.78 9.65 11.96 May 1974 21.26 4.42 4.42 24.25 32.24 7.42 5.99 2-yr 15.06 3.35 4.28 21.98 34.33 9.92 11.06 June 1974 17.26 4.99 3.12 25.57 34.51 5.41 9.15 2-yr 12.85 3.68 3.37 17.95 30.56 9.72 21.87 July 1974 15.47 5.56 6.91 19.67 31.83 8.56 12.01 2-yr 7.67 3.33 4.27 21.49 38.13 9.84 15.27 August 1974 10.01 3.65 5.28 29.50 37.89 6.77 6.90 2-yr 6.73 3.05 4.58 28.36 34.88 11.10 11.30 September 1974 14.12 5.24 4.37 26.93 33.77 7.57 8.01 2-yr 7.77 3.78 3.63 23.46 41.03 7.63 12.71 October 1974 15.80 2.66 2.24 17.62 25.73 14.13 21.82 2-yr 10.98 3.08 2.87 18.72 31.71 12.12 20.52 November 1974 16.74 4.08 3.52 21.38 29.11 13.22 11.95 2-yr 11.72 3.07 3.97 21.48 31.94 13.88 13.95 December 1974 16.95 4.89 3.02 27.01 34.48 7.61 6.03 2-yr 13.27 3.76 2.77 29.03 36.48 1.95 6.74

a. The first-year data period is March 3, 1974, to March 2, 1975, and the 2-year data period is the combined years of March 3, 1974, to March 2, 1975, and May 1, 1975, to April 30, 1976.

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-9 Table 11A-3

/Q AND D/Q VALUES AT SPECIAL DISTANCES AND RELEASE MODES FOR A 2-YEAR DATA PERIODa Mixed-Mode Release (joint ground-level and elevated Ground-Level Release release)

Receptor Direction Distance (m) /Q (sec/m3) D/Q (m-2) /Q (sec/m3) D/Q (m-2)

NNE 2414 4.7 (-06)b 1.0 (-08) 3.9 (-07) 2.2 (-09)

NNE 3058 2.9 (-06) 5.9 (-09) 2.8 (-07) 1.3 (-09)

NE 2333 5.2 (-06) 1.1 (-08) 4.2 (-07) 3.3 (-09)

S 503 3.1 (-05) 1.7 (-07) 8.0 (-07) 3.5 (-08)

S 2470 1.5 (-06) 5.5 (-09) 3.1 (-07) 2.5 (-09)

SSW 3492 3.8 (-07) 1.4 (-09) 1.2 (-07) 7.7 (-10)

SW 2881 6.7 (-07) 2.0 (-09) 1.2 (-07) 9.7 (-10)

SW 3379 4.8 (-07) 1.4 (-09) 1.1 (-07) 6.6 (-10)

WSW 4828 2.0 (-07) 5.9 (-10) 6.8 (-08) 3.3 (-10)

NNW 6034 5.9 (-07) 5.9 (-10) 5.5 (-08) 1.2 (-10)

N 274 2.7 (-04) 5.9 (-07) 5.8 (-07) 2.6 (-08)

N 503 9.5 (-05) 2.2 (-07) 4.5 (-07) 1.7 (-08)

SSE 4747 4.1 (-07) 1.3 (-09) 7.5 (-08) 4.8 (-10)

a. Data period is March 3, 1974, to March 2, 1975, and May 1, 1975, to April 30, 1976. Open terrain corrective factors of Regulatory Guide 1.111 are incorporated.

b. 4.7 (-06) = 4.7 x 10-6.

Revision 52Updated Online 09/30/20 Table 11A-4 ANNUAL AVERAGE /Q (sec/m ) VALUES BASED ON A GROUND-LEVEL RELEASE FOR A 2-YEAR DATA PERIODa 3

Receptor Distance in Meters Direction 805 2414 4023 5633 7242 12070 24140 40234 56327 72420 N 4.4 (-5)b 4.7 (-6) 1.7 (-6) 9.4 (-7) 6.2 (-7) 2.6 (-7) 1.1 (-7) 5.5 (-8) 3.6 (-8) 2.7 (-8)

NNE 4.4 (-5) 4.7 (-6) 1.7 (-6) 9.2 (-7) 6.1 (-7) 2.6 (-7) 1.0 (-7) 5.4 (-8) 3.6 (-8) 2.6 (-8)

NE 4.5 (-5) 4.8 (-6) 1.7 (-6) 9.5 (-7) 6.2 (-7) 2.6 (-7) 1.1 (-7) 5.6 (-8) 3.7 (-8) 2.7 (-8)

ENE 2.0 (-5) 2.1 (-6) 7.5 (-7) 4.1 (-7) 2.7 (-7) 1.1 (-7) 4.6 (-8) 2.4 (-8) 1.6 (-8) 1.1 (-8)

E 1.8 (-5) 1.9 (-6) 6.6 (-7) 3.6 (-7) 2.4 (-7) 9.9 (-8) 4.0 (-8) 2.1 (-8) 1.4 (-8) 1.0 (-8)

ESE 1.5 (-5) 1.5 (-6) 5.3 (-7) 2.9 (-7) 1.9 (-7) 7.8 (-8) 3.1 (-8) 1.6 (-8) 1.1 (-0) 7.8 (-9)

SE 1.5 (-5) 1.6 (-6) 5.5 (-7) 3.0 (-7) 1.9 (-7) 8.1 (-8) 3.3 (-8) 1.7 (-8) 1.1 (-8) 8.1 (-9)

SSE 1.6 (-5) 1.6 (-6) 5.6 (-7) 3.0 (-7) 1.9 (-7) 8.0 (-8) 3.2 (-8) 1.6 (-8) 1.1 (-8) 7.8 (-9)

S 1.5 (-5) 1.6 (-6) 5.3 (-7) 2.9 (-7) 1.8 (-7) 7.6 (-8) 3.0 (-8) 1.5 (-8) 1.0 (-8) 7.3 (-9)

SSW 8.1 (-6) 8.2 (-7) 2.8 (-7) 1.5 (-7) 9.4 (-8) 3.8 (-8) 1.5 (-8) 7.6 (-9) 4.9 (-9) 3.6 (-9)

SW 9.4 (-6) 9.5 (-7) 3.3 (-7) 1.0 (-7) 1.2 (-7) 4.8 (-8) 1.9 (-8) 9.7 (-9) 6.4 (-9) 4.6 (-9)

WSW 8.2 (-6) 8.2 (-7) 2.8 (-7) 1.5 (-7) 9.5 (-8) 3.9 (-8) 1.5 (-8) 7.8 (-9) 5.0 (-9) 3.7 (-9)

SPS UFSAR W 1.3 (-5) 1.3 (-6) 4.4 (-7) 2.4 (-7) 1.5 (-7) 6.3 (-8) 2.5 (-8) 1.3 (-8) 8.2 (-9) 6.0 (-9)

WNW 1.8 (-5) 1.9 (-6) 6.4 (-7) 3.5 (-7) 2.3 (-7) 9.4 (-8) 3.8 (-8) 2.0 (-8) 1.3 (-8) 9.3 (-9)

NW 2.1 (-5) 2.2 (-6) 7.8 (-7) 4.2 (-7) 2.8 (-7) 1.2 (-7) 4.6 (-8) 2.4 (-8) 1.6 (-8) 1.2 (-8)

NNW 3.3 (-5) 3.6 (-6) 1.3 (-6) 7.0 (-7) 4.6 (-7) 2.0 (-7) 7.9 (-8) 4.1 (-8) 2.7 (-8) 2.0 (-8) 0.5 miles 1.5 miles 2.5 miles 3.5 miles 4.5 miles 7.5 miles 15.0 miles 25.0 miles 35.0 miles 45.0 miles

a. Data period is March 3, 1974 to March 2, 1975, and May 1, 1975, to April 30, 1976. Open terrain correction factors of Regulatory Guide 1.111 are incorpo-rated.

b. 4.4 (-05) = 4.4 x 10-5.

11A-10

Revision 52Updated Online 09/30/20 Table 11A-5 ANNUAL AVERAGE D/Q (M-2) VALUES BASED ON A GROUND-LEVEL RELEASE FOR A 2-YEAR DATA PERIODa Receptor Distance in Meters Direction 805 2414 4023 5633 7242 12070 24140 40234 56327 72420 N 9.7 (-08)b 7.3 (-09) 2.1 (-09) 9.9 (-10) 5.9 (-10) 2.0 (-10) 5.4 (-11) 2.0 (-11) 1.0 (-11) 6.2 (-12)

NNE 1.4 (-07) 1.0 (-08) 3.0 (-09) 1.4 (-09) 8.4 (-10) 2.8 (-10) 7.8 (-11) 2.9 (-11) 1.5 (-11) 8.9 (-12)

NE 1.4 (-07) 1.0 (-08) 3.0 (-09) 1.4 (-09) 8.3 (-10) 2.8 (-10) 7.6 (-11) 2.8 (-11) 1.4 (-11) 8.8 (-12)

ENE 6.4 (-08) 4.8 (-09) 1.4 (-09) 6.5 (-10) 3.9 (-10) 1.3 (-10) 3.6 (-11) 1.3 (-11) 6.7 (-12) 4.1 (-12)

E 6.0 (-08) 4.5 (-09) 1.3 (-09) 6.1 (-10) 3.6 (-10) 1.2 (-10) 3.3 (-11) 1.2 (-11) 6.2 (-12) 3.8 (-12)

ESE 6.0 (-08) 4.5 (-09) 1.3 (-09) 6.2 (-10) 3.7 (-10) 1.2 (-10) 3.4 (-11) 1.2 (-11) 6.3 (-12) 3.9 (-12)

SE 7.4 (-08) 5.5 (-09) 1.6 (-09) 7.6 (-10) 4.5 (-10) 1.5 (-10) 4.1 (-11) 1.5 (-11) 7.7 (-12) 4.7 (-12)

SSE 8.4 (-08) 6.3 (-09) 1.8 (-09) 8.6 (-10) 5.1 (-10) 1.7 (-10) 4.7 (-11) 1.7 (-11) 8.8 (-12) 5.4 (-12)

S 7.7 (-08) 5.7 (-09) 1.7 (-09) 7.9 (-10) 4.6 (-10) 1.5 (-10) 4.3 (-11) 1.6 (-11) 8.0 (-12) 4.9 (-12)

SSW 4.5 (-08) 3.3 (-09) 9.6 (-10) 4.6 (-10) 2.7 (-10) 9.0 (-11) 2.5 (-11) 9.1 (-12) 4.6 (-12) 2.9 (-12)

SW 4.0 (-08) 3.0 (-09) 8.8 (-10) 4.2 (-10) 2.5 (-10) 8.2 (-11) 2.3 (-11) 8.3 (-12) 4.2 (-12) 2.6 (-12)

WSW 4.1 (-08) 3.1 (-09) 8.8 (-10) 4.2 (-10) 2.5 (-10) 8.2 (-11) 2.3 (-11) 8.3 (-12) 4.3 (-12) 2.6 (-12)

SPS UFSAR W 6.4 (-08) 4.8 (-09) 1.4 (-09) 6.5 (-10) 3.9 (-10) 1.3 (-10) 3.6 (-11) 1.3 (-11) 6.7 (-12) 4.1 (-12)

WNW 7.0 (-08) 5.2 (-09) 1.5 (-09) 7.1 (-10) 4.2 (-10) 1.4 (-10) 3.9 (-11) 1.4 (-11) 7.3 (-12) 4.5 (-12)

NW 7.2 (-08) 5.4 (-09) 1.6 (-09) 7.3 (-10) 4.3 (-10) 1.4 (-10) 4.0 (-11) 1.5 (-11) 7.5 (-12) 4.6 (-12)

NNW 7.0 (-08) 5.3 (-09) 1.5 (-09) 7.2 (-10) 4.3 (-10) 1.4 (-10) 3.9 (-11) 1.4 (-11) 7.3 (-12) 4.5 (-12) 0.5 miles 1.5 miles 2.5 miles 3.5 miles 4.5 miles 7.5 miles 15.0 miles 25.0 miles 35.0 miles 45.0 miles

a. Data period is March 3, 1974, to March 2, 1975, and May 1, 1915, to April 30, 1976. Open terrain correction factors of Regulatory Guide 1.111 are incor-porated.

b. 9.7 (-08) = 9.7 x 10-8.

11A-11

Revision 52Updated Online 09/30/20 Table 11A-6 ANNUAL AVERAGE /Q (SEC/M ) VALUES BASED ON A MIXED-MODE RELEASE FOR A 2-YEAR DATA PERIODa 3

Receptor Distance in Meters Direction 805 2414 4023 5633 7242 12070 24140 40234 56327 72420 N 5.6 (-7)b 2.8 (-7) 1.4 (-7) 8.9 (-8) 6.8 (-8) 3.5 (-8) 1.2 (-8) 6.3 (-9) 4.1 (-9) 3.0 (-9)

NNE 8.8 (-7) 3.9 (-7) 1.8 (-7) 1.2 (-7) 8.9 (-8) 4.4 (-8) 1.4 (-8) 7.3 (-9) 4.8 (-9) 3.5 (-9)

NE 1.0 (-6) 3.9 (-7) 1.8 (-7) 1.1 (-7) 8.2 (-8) 4.1 (-8) 1.4 (-8) 6.9 (-9) 4.5 (-9) 3.3 (-9)

ENE 6.6 (-7) 2.3 (-7) 1.1 (-7) 6.6 (-8) 4.7 (-8) 2.4 (-8) 8.1 (-9) 4.1 (-9) 2.7 (-9) 2.0 (-9)

E 6.8 (-7) 2.3 (-7) 1.0 (-7) 6.4 (-8) 4.6 (-8) 2.3 (-8) 8.0 (-9) 4.1 (-9) 2.6 (-9) 1.9 (-9)

ESE 6.7 (-7) 1.9 (-7) 8.2 (-8) 5.3 (-8) 3.7 (-8) 1.8 (-8) 7.2 (-9) 3.7 (-9) 2.4 (-9) 1.7 (-9)

SE 7.5 (-7) 2.1 (-7) 9.1 (-8) 5.6 (-8) 4.0 (-8) 1.9 (-8) 7.1 (-9) 3.6 (-9) 2.3 (-9) 1.7 (-9)

SSE 7.3 (-7) 2.2 (-7) 9.5 (-8) 5.8 (-8) 4.0 (-8) 1.9 (-8) 6.7 (-9) 3.4 (-9) 2.2 (-9) 1.6 (-9)

S 8.7 (-7) 3.1 (-7) 1.4 (-7) 8.4 (-8) 5.9 (-8) 2.8 (-8) 8.6 (-9) 4.4 (-9) 2.8 (-9) 2.1 (-9)

SSW 6.2 (-7) 2.0 (-7) 8.9 (-8) 5.9 (-8) 4.3 (-8) 2.0 (-8) 6.6 (-9) 3.4 (-9) 2.2 (-9) 1.6 (-9)

SW 5.4 (-7) 1.6 (-7) 7.8 (-8) 4.8 (-8) 3.5 (-8) 1.7 (-8) 5.6 (-9) 2.9 (-9) 1.9 (-9) 1.4 (-9)

WSW 6.0 (-7) 1.7 (-7) 7.8 (-8) 5.3 (-8) 3.7 (-8) 1.9 (-8) 6.1 (-9) 3.1 (-9) 2.0 (-9) 1.5 (-9)

SPS UFSAR W 7.1 (-7) 2.3 (-7) 1.0 (-7) 6.7 (-8) 5.0 (-8) 2.3 (-8) 7.5 (-9) 3.8 (-9) 2.5 (-9) 1.8 (-9)

WNW 6.0 (-7) 2.3 (-7) 1.1 (-7) 6.6 (-8) 4.7 (-8) 2.3 (-8) 8.9 (-9) 4.5 (-9) 3.0 (-9) 2.2 (-9)

NW 7.5 (-7) 2.2 (-7) 1.0 (-7) 6.4 (-8) 4.6 (-8) 2.5 (-8) 9.2 (-9) 4.7 (-9) 3.1 (-9) 2.2 (-9)

NNW 4.4 (-7) 1.9 (-7) 9.6 (-8) 6.3 (-8) 4.7 (-8) 2.8 (-8) 9.8 (-9) 5.0 (-9) 3.3 (-9) 2.4 (-9) 0.5 miles 1.5 miles 2.5 miles 3.5 miles 4.5 miles 7.5 miles 15.0 miles 25.0 miles 35.0 miles 45.0 miles

a. Data period is March 3, 1974, to March 2, 1975, and May 1, 1915, to April 30, 1976. Open terrain correction factors of Regulatory Guide 1.111 are incor-porated.

b. 5.6 (-7) = 5.6 x 107.

11A-12

Revision 52Updated Online 09/30/20 Table 11A-7 ANNUAL AVERAGE D/Q (M-2) VALUES BASED ON A MIXED-MODE RELEASE FOR A 2-YEAR DATA PERIODa Receptor Distance in Meters Direction 805 2414 4023 5633 7242 12070 24140 40234 56327 72420 N 1.0 (-08)b 1.1 (-09) 3.4 (-10) 1.8 (-10) 1.2 (-10) 5.8 (-11) 2.4 (-11) 1.0 (-11) 5.9 (-12) 4.1 (-12)

NNE 2.0 (-08) 2.2 (-09) 6.7 (-10) 3.4 (-10) 2.2 (-10) 9.8 (-11) 3.8 (-11) 1.6 (-11) 9.3 (-12) 6.4 (-12)

NE 2.9 (-08) 3.0 (-09) 9.0 (-10) 4.5 (-10) 2.9 (-10) 1.2 (-10) 4.3 (-11) 1.9 (-11) 1.1 (-11) 7.0 (-12)

ENE 1.6 (-08) 1.7 (-09) 5.2 (-10) 2.6 (-10) 1.6 (-10) 6.4 (-11) 2.3 (-11) 9.9 (-12) 5.6 (-12) 3.7 (-12)

E 1.7 (-08) 1.8 (-09) 5.3 (-10) 2.6 (-10) 1.6 (-10) 6.4 (-11) 2.3 (-11) 9.7 (-12) 5.5 (-12) 3.6 (-12)

ESE 2.1 (-08) 2.2 (-09) 6.5 (-10) 3.2 (-10) 1.9 (-10) 7.2 (-11) 2.4 (-11) 1.0 (-11) 5.7 (-12) 3.7 (-12)

SE 2.5 (-08) 2.6 (-09) 7.6 (-10) 3.7 (-10) 2.3 (-10) 8.5 (-11) 2.9 (-11) 1.2 (-11) 6.7 (-12) 4.4 (-12)

SSE 2.1 (-08) 2.3 (-09) 6.9 (-10) 3.4 (-10) 2.1 (-10) 7.8 (-11) 2.7 (-11) 1.2 (-11) 6.5 (-12) 4.3 (-12)

S 2.3 (-08) 2.6 (-09) 7.9 (-10) 3.9 (-10) 2.4 (-10) 9.1 (-11) 3.2 (-11) 1.4 (-11) 7.9 (-12) 5.2 (-12)

SSW 1.5 (-08) 1.8 (-09) 5.3 (-10) 2.6 (-10) 1.6 (-10) 6.1 (-11) 2.1 (-11) 9.2 (-12) 5.2 (-12) 3.4 (-12)

SW 1.3 (-08) 1.4 (-09) 4.2 (-10) 2.1 (-10) 1.3 (-10) 4.8 (-11) 1.5 (-11) 7.0 (-12) 3.9 (-12) 2.6 (-12)

WSW 1.5 (-08) 1.6 (-09) 4.9 (-10) 2.4 (-10) 1.5 (-10) 5.4 (-11) 1.9 (-11) 8.0 (-12) 4.5 (-12) 2.9 (-12)

SPS UFSAR W 1.7 (-08) 1.9 (-09) 5.7 (-10) 2.8 (-10) 1.7 (-10) 6.5 (-11) 2.3 (-11) 9.7 (-12) 5.5 (-12) 3.6 (-12)

WNW 1.2 (-08) 1.3 (-09) 4.0 (-10) 2.0 (-10) 1.3 (-10) 5.2 (-11) 1.9 (-11) 8.2 (-12) 4.7 (-12) 3.2 (-12)

NW 2.5 (-08) 2.4 (-09) 7.1 (-10) 3.5 (-10) 2.2 (-10) 8.2 (-11) 2.8 (-11) 1.2 (-11) 6.5 (-12) 4.3 (-12)

NNW 9.3 (-08) 9.6 (-10) 2.9 (-10) 1.5 (-10) 9.9 (-11) 4.5 (-11) 1.8 (-11) 7.5 (-12) 4.3 (-12) 3.0 (-12) 0.5 miles 1.5 miles 2.5 miles 3.5 miles 4.5 miles 7.5 miles 15.0 miles 25.0 miles 35.0 miles 45.0 miles

a. Data period is March 3, 1974 to March 2, 1975, and May 1, 1975, to April 30, 1976. Open terrain correction factors of Regulatory Guide 1.111 are incor-porated

b. 1.0 (-08) = 1.0 x 10-8.

11A-13

Revision 52Updated Online 09/30/20 SPS UFSAR 11A-14 Table 11A-8 0- TO 5-MILE HIGHPOINTS BY MILE AND 5- TO 10-MILE HIGHPOINTS FOR 16 CARDINAL POINTS FROM SURRY POWER STATIONa Section 0-1 1-2 2-3 3-4 4-5 5-10 N 37 - - 12 55 80 NNE 37 5 2 - 71 100 NE 38 4 - - 52 80 ENE 43 10 - - 30 80 E 38 3 - - - 60 ESE 38 37 - - 11 30 SE 39 36 - - - -

SSE 39 33 39 37 37 50 S 37 39 50 51 60 80 SSW 42 38 42 70 85 90 SW 39 34 70 72 84 95 WSW 40 - 55 81 83 130 W 38 - - 70 88 90 WNW 39 - - - - 10 NW 38 - - 6 5 80 NNW 39 - - 12 47 110

a. All highpoints are measured in feet. Dash (-) denotes sea level.

Sources: Surry Power Station Units 3 and 4 Environment Report, Figures 2.6-8 through 2.6-11; and U.S. Geological Survey 7.5 minute topographic maps.

Revision 52Updated Online 09/30/20 Table 11A-9 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITHOUT STEAM GENERATOR BLOWDOWN PROCESSING Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Corrosion And Activation Products CR51 2.78E+01 1.68E-03 2.11E-07 0.00257 0.00844 0.02643 0.00000 0.03744 0.03757 0.00000 0.03800 MN54 3.03E+02 2.74E-04 5.06E-08 0.00052 0.00138 0.00634 0.00000 0.00824 0.00827 0.00036 0.00860 FE55 9.50E+02 1.41E-03 1.77E-07 0.00271 0.00713 0.02214 0.00000 0.03199 0.03210 0.00000 0.03200 FE59 4.50E+01 8.84E-04 1.30E-07 0.00148 0.00445 0.01624 0.00000 0.02217 0.02225 0.00000 0.02200 CO58 7.13E+01 1.41E-02 1.80E-06 0.02492 0.07121 0.22509 0.00000 0.32124 0.32239 0.00144 0.32000 C060 1.92E+03 1.76E-03 2.27E-07 0.00340 0.00891 0.02845 0.00000 0.04077 0.04091 0.00313 0.04400 NP239 2.35E+00 1.09E-03 1.14E-07 0.00022 0.00512 0.01426 0.00000 0.01960 0.01967 0.00000 0.02000 SPS UFSAR Fission Products BR83 1.00E-01 4.97E-03 1.96E-07 0.00000 0.00498 0.02458 0.00000 0.02956 0.02967 0.00000 0.03000 BR84 2.21E-02 2.77E-03 3.29E-08 0.00000 0.00003 0.00412 0.00000 0.00415 0.00416 0.00000 0.00420 BR85 2.08E-03 3.22E-04 3.53E-10 0.00000 0.00000 0.00004 0.00000 0.00004 0.00004 0.00000 0.00004 RB86 1.87E+01 7.47E-05 1.07E-08 0.00018 0.01868 0.00134 0.00000 0.02019 0.02027 0.00000 0.02000 RB88 1.24E-02 2.14E-01 1.36E-06 0.00000 0.00218 0.17036 0.00000 0.17304 0.17367 0.00000 0.17000 SR89 5.20E+01 3.09E-04 5.17E-08 0.00053 0.00156 0.00647 0.00000 0.00856 0.00859 0.00000 0.00860 SR90 1.03E+04 8.82E-06 1.26E-09 0.00002 0.00004 0.00016 0.00000 0.00022 0.00022 0.00000 0.00022 Y90 2.67E+00 1.09E-06 7.31E-10 0.00002 0.00001 0.00009 0.00000 0.00011 0.00012 0.00000 0.00012 11A-15 SR91 4.03E-01 6.35E-04 4.74E-08 0.00000 0.00210 0.00594 0.00000 0.00805 0.00808 0.00000 0.00810

Table 11A-9 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITHOUT STEAM GENERATOR BLOWDOWN PROCESSING Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Fission Products Y91M 3.47E-02 3.82E-04 4.73E-08 0.00000 0.00136 0.00593 0.00000 0.00728 0.00731 0.00000 0.00730 Y91 5.88E+01 5.66E-05 7.73E-09 0.00011 0.00029 0.00097 0.00000 0.00137 0.00137 0.00000 0.00140 Y93 4.25E-01 3.31E-05 3.09E-09 0.00000 0.00011 0.00039 0.00000 0.00050 0.00050 0.00000 0.00050 ZR95 6.50E+01 5.30E-05 7.71E-09 0.00009 0.00027 0.00097 0.00000 0.00133 0.00133 0.00000 0.00130 NB95 3.50E+01 4.42E-05 7.84E-09 0.00009 0.00022 0.00098 0.00000 0.00129 0.00130 0.00000 0.00130 MO99 2.79E+00 7.61E-02 1.08E-05 0.02037 0.36131 1.35526 0.00010 1.73705 1.74331 0.00000 1.70000 TC99M 2.50E-01 4.80E-02 2.91E-05 0.01948 0.28162 3.64351 0.00019 3.94479 3.95902 0.00000 4.00000 RU103 3.96E+01 3.98E-05 5.20E-09 0.00007 0.00020 0.00065 0.00000 0.00092 0.00092 0.00005 0.00097 SPS UFSAR RH103M 3.96E-02 4.76E-05 3.10E-08 0.00007 0.00020 0.00389 0.00000 0.00416 0.00417 0.00000 0.00420 RU106 3.67E+02 8.63E-06 1.26E-09 0.00002 0.00004 0.00016 0.00000 0.00022 0.00022 0.00086 0.00110 RH106 3.47E-04 1.08E-05 8.89E-09 0.00002 0.00004 0.00111 0.00000 0.00118 0.00118 0.00000 0.00120 TE125M 5.80E+01 2.56E-05 2.32E-09 0.00004 0.00013 0.00029 0.00000 0.00046 0.00047 0.00000 0.00047 TE127M 1.09E+02 2.47E-04 2.30E-08 0.00045 0.00125 0.00288 0.00000 0.00457 0.00459 0.00000 0.00460 TE127 3.92E-01 8.32E-04 1.60E-07 0.00045 0.00315 0.02007 0.00000 0.02367 0.02375 0.00000 0.02400 TE129M 3.40E+01 1.24E-03 1.57E-07 0.00197 0.00622 0.01966 0.00000 0.02786 0.02796 0.00000 0.02800 TE129 4.79E-02 1.69E-03 9.07E-07 0.00127 0.00417 0.11370 0.00000 0.11914 0.11957 0.00000 0.12000 11A-16

Table 11A-9 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITHOUT STEAM GENERATOR BLOWDOWN PROCESSING Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Corrosion and Activation Products I130 5.17E-01 2.03E-03 1.77E-07 0.00000 0.00736 0.022112 0.00001 0.02950 0.02961 0.00000 0.03000 TE131M 1.25E+00 2.32E-03 2.38E-07 0.00011 0.01021 0.02985 0.00000 0.04018 0.04033 0.00000 0.04000 TE131 1.74E-02 1.17E-03 8.36E-07 0.00002 0.00187 0.10471 0.00000 0.10660 0.10699 0.00000 0.11000 I131 8.05E+00 2.41E-01 3.20E-05 0.02153 1.18990 4.01574 0.00312 5.23029 5.24915 0.00002 5.20000 TE132 3.25E+00 2.44E-02 2.77E-06 0.00825 0.11676 0.34691 0.00003 0.47196 0.47366 0.00000 0.47000 I132 9.58E-02 1.04E-01 1.46E-05 0.00851 0.19360 1.82326 0.00026 2.02562 2.03292 0.00000 2.00000 I133 8.75E-01 3.58E-01 3.61E-05 0.00060 1.48638 4.52219 0.00295 6.01212 6.03380 0.00000 6.00000 I134 3.67E-02 4.98E-02 8.93E-07 0.00000 0.00449 0.11191 0.00000 0.11640 0.11640 0.00000 0.12000 SPS UFSAR CS134 7.49E+02 2.18E-02 2.99E-06 0.07167 5.50343 0.37523 0.00003 5.95036 5.97182 0.00468 6.00000 I135 2.79E-01 1.89E-01 1.30E-05 0.00000 0.52153 1.62652 0.00069 2.14874 2.15649 0.00000 2.20000 CS136 1.30E+01 1.15E-02 1.37E-06 0.02321 2.85474 0.17198 0.00001 3.04995 3.06094 0.00000 3.10000 CS137 1.10E+04 1.57E-02 1.99E-06 0.05202 3.96259 0.24974 0.00002 4.26437 4.27975 0.00864 4.30000 BA137M 1.77E-03 1.72E-02 1.41E-05 0.04864 3.70502 1.77272 0.00002 5.52641 5.54634 0.00000 5.50000 BA140 1.28E+01 1.95E-04 2.49E-08 0.00023 0.00097 0.00312 0.00000 0.00433 0.00434 0.00000 0.00430 LA140 1.68E+00 1.38E-04 3.40E-08 0.00025 0.00072 0.00426 0.00000 0.00524 0.00526 0.00000 0.00530 CE141 3.24E+01 6.19E-05 7.86E-09 0.00010 0.00031 0.00098 0.00000 0.00139 0.00140 0.00000 0.00140 11A-17 CE143 1.38E+00 3.70E-05 4.13E- 09 0.00000 0.00016 0.00052 0.00000 0.00068 0.00069 0.00000 0.00069

Table 11A-9 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITHOUT STEAM GENERATOR BLOWDOWN PROCESSING Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Corrosion and Activation Products PR143 1.37E+01 4.44E-05 5.51 E-09 0.00006 0.00022 0.00069 0.00000 0.00097 0.00098 0.00000 0.00098 CE144 2.84E+02 2.91E-05 5.06E-09 0.00006 0.00015 0.00063 0.00000 0.00084 0.00084 0.00187 0.00270 PR144 1.20E-02 3.53E-05 3.41E-08 0.00006 0.00015 0.00427 0.00000 0.00447 0.00449 0.00000 0.00450 All Others 0.00 0.00 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 Total (Except Tritium) 1.41E+00 1.67E-04 0.31637 20.35739 20.91034 0.00748 41.59157 41.74157 0.02244 42.00000 Tritium Release 480 Curies Per Year SPS UFSAR 11A-18

Revision 52Updated Online 09/30/20 Table 11A-10 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITH STEAM GENERATOR BLOWDOWN PROCESSINGa Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Corrosion And Activation Products CR51 2.78E+01 1.68E-03 2.11E-07 0.00257 0.00044 0.00003 0.00000 0.01104 0.01118 0.00000 0.01100 MN54 3.03E+02 2.74E-04 5.06E-08 0.00058 0.00138 0.00001 0.00000 0.00191 0.00198 0.00036 0.00230 FE55 9.50E+02 1.41E-03 1.77E-07 0.00271 0.00713 0.00002 0.00000 0.00987 0.000994 0.00000 0.00990 FE59 4.50E+01 8.84E-04 1.30E-07 0.00148 0.00448 0.00002 0.00000 0.00594 0.00599 0.00000 0.00600 CO58 7.13E+01 1.41E-02 1.80E-06 0.02492 0.07121 0.00023 0.00002 0.09639 0.09707 0.00144 0.09900 C060 1.98E+03 1.76E-03 2.27E-07 0.00340 0.00891 0.00003 0.00000 0.01235 0.01243 0.00313 0.01600 NP239 2.35E+00 1.09E-03 1.14E-07 0.00022 0.00512 0.00001 0.00000 0.00536 0.00540 0.00000 0.00540 SPS UFSAR Fission Products BR83 1.00E-01 4.97E-03 1.96E-07 0.00000 0.00498 0.00002 0.00000 0.00501 0.00505 0.00000 0.05000 BR84 2.21E-02 2.77E-03 3.29E-08 0.00000 0.00003 0.00000 0.00000 0.00003 0.00003 0.00000 0.00003 RB86 1.87E+01 7.47E-03 1.07E-08 0.00018 0.01848 0.00001 0.00000 0.01887 0.01901 0.00000 0.01900 RB88 1.24E-02 2.14E-01 1.38E-06 0.00000 0.00218 0.00171 0.00000 0.00389 0.00398 0.00000 0.00390 SR89 5.40E+01 3.09E-04 5.17E-08 0.00053 0.00156 0.00001 0.00000 0.00209 0.00211 0.00000 0.00210 SR90 1.03E+04 6.82E-06 1.26E-09 0.00002 0.00004 0.00000 0.00000 0.00006 0.00006 0.00000 0.00006 Y90 2.67E+00 1.09E-06 7.31E-10 0.00002 0.00001 0.00000 0.00000 0.00002 0.00002 0.00000 0.00002 SR91 4.03E-01 6.35E-04 4.74E-08 0.00000 0.00210 0.00001 0.00000 0.00211 0.00213 0.00000 0.00210

a. Steam generator blowdown treatment system is no longer used. 11A-19

Table 11A-10 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITH STEAM GENERATOR BLOWDOWN PROCESSINGa Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

Fission Products Y91M 1.47E-02 3.82E-04 4.73E-08 0.00000 0.00136 0.00001 0.00000 0.00136 0.00137 0.00000 0.00140 Y91 5.88E+01 5.66E-05 7.73E-09 0.00011 0.00029 0.00000 0.00000 0.00040 0.00040 0.00000 0.00040 Y95 4.45E-01 3.31 E-05 3.09E-09 0.00000 0.00011 0.00000 0.00000 0.00011 0.00011 0.00000 0.00011 ZR95 6.50E+01 5.30E-05 7.71E-09 0.00009 0.00027 0.00000 0.00000 0.00036 0.00036 0.00000 0.00036 NB95 3.50E+01 4.42E-05 7.84E-09 0.00009 0.00022 0.00000 0.00000 0.00031 0.00031 0.00000 0.00031 MO99 2.79E+00 7.61E-02 1.08E-05 0.02037 0.36131 0.00136 0.00010 0.38314 0.38591 0.00000 0.39000 TC99M 2.50E-01 4.80E-02 2.91E-05 0.01948 0.28162 0.00364 0.00019 0.30493 0.30713 0.00000 0.31000 RU103 3.96E+01 3.98E-05 5.20E-09 0.00007 0.00020 0.00000 0.00000 0.00027 0.00027 0.00005 0.00032 SPS UFSAR RH103M 3.96E-02 4.76E-05 3.10E-08 0.00007 0.00020 0.00000 0.00000 0.00027 0.00027 0.00000 0.00027 RH106 3.67E+02 8.83E-06 1.26E-09 0.00008 0.00004 0.00000 0.00000 0.00006 0.00006 0.00086 0.00093 RH106 3.47E-04 1.08E-05 8.89E-09 0.00002 0.00004 0.00000 0.00000 0.00006 0.00006 0.00000 0.00027 TE125M 5.80E+01 4.56E-05 4.32E-09 0.00004 0.00013 0.00000 0.00000 0.00017 0.00017 0.00000 0.00170 TE127M 1.09E+02 2.47E-04 2.30E-08 0.00045 0.00125 0.00000 0.00000 0.00170 0.00171 0.00000 0.00360 TE127 4.92E-01 8.32E-04 1.60E-07 0.00045 0.00315 0.00004 0.00000 0.00362 0.00365 0.00000 0.00830 TE129M 3.40E+01 1.24E-03 1.57E-07 0.00197 0.00622 0.00004 0.00000 0.00822 0.00828 0.00000 0.00560 TE130 5.17E-01 2.03E-03 1.77E-07 0.00000 0.00736 0.00002 0.00001 0.00740 0.00745 0.00000 0.00750

a. Steam generator blowdown treatment system is no longer used. 11A-20

Table 11A-10 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITH STEAM GENERATOR BLOWDOWN PROCESSINGa Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

TE131M 1.25E+00 2.32E-03 2.38E-07 0.00011 0.01021 0.00003 0.00000 0.01036 0.01043 0.00000 0.01000 TE131 1.74E-02 1.17E-03 8.36E-07 0.00004 0.00187 0.00010 0.00000 0.00199 0.00401 0.00000 0.00200 Corrosion and Activation Products T131 8.05E+00 2.41E-01 3.20E-07 0.02153 1.18990 0.00402 0.00318 1.21857 1.28739 0.00002 1.20000 TE132 3.25E+00 2.44E-02 2.77E-06 0.00025 0.11876 0.00035 0.00003 0.12539 0.12630 0.00000 0.13000 T132 4.58E-04 1.04E-01 1.46E-05 0.00851 0.19360 0.00182 0.00026 0.20419 0.20567 0.00000 0.21000 T135 8.75E-01 3.58E-01 3.61E-05 0.00060 1.48538 0.00452 0.00295 1.49445 1.50328 0.00000 1.50000 T134 3.67E-02 4.98E-02 8.93E-07 0.00000 0.00449 0.00011 0.00000 0.00460 0.00463 0.00000 0.00460 CS134 7.49E+02 2.18E- 02 2.99E-06 0.07167 5.50343 0.00175 0.00003 5.57888 5.61929 0.00468 5.60000 SPS UFSAR T135 2.79E-01 1.892E-01 1.30E-05 0.00000 0.52153 0.00165 0.00089 0.52385 0.52764 0.00000 0.53000 CS136 1.40E+01 1.15E-02 1.47E-06 0.02321 2.85474 0.00172 0.00001 2.87968 2.90054 0.00000 2.90000 CS137 1.10E+04 1.57E-02 1.99E-06 0.05202 3.98459 0.00250 0.00004 4.01713 4.04625 0.00864 4.10000 BA137M 1.77E-03 1.78E-02 1.41E-05 0.04864 3.70502 0.00177 0.00002 3.75546 3.78266 0.00000 3.80000 BA140 1.28E+01 1.95E-04 2.49E-08 0.00023 0.00097 0.00000 0.00000 0.00121 0.00121 0.00000 0.00120 LA140 1.68E+00 1.38E-04 3.40E-08 0.00023 0.00072 0.00000 0.00000 0.00098 0.00099 0.00000 0.00099 CE 41 3.24E+01 6.19E-05 7.86E-09 0.00010 0.00031 0.00000 0.00000 0.00041 0.00041 0.00000 0.00041 CE143 1.38E+00 3.70E-05 4.13E- 09 0.00000 0.00016 0.00000 0.00000 0.00017 0.00017 0.00000 0.00017

a. Steam generator blowdown treatment system is no longer used. 11A-21

Table 11A-10 (CONTINUED)

Revision 52Updated Online 09/30/20 LIQUID SOURCE TERMS FROM SURRY UNITS 1 & 2 (PER UNIT BASIS)

WITH STEAM GENERATOR BLOWDOWN PROCESSINGa Annual Releases To Discharge Canal Coolant Concentrations Primary Secondary Misc. Total Adjusted Detergent Half-life (Micro (Micro Boron Rs Wastes Secondary Turb Bldg LWS Total Wastes Total Nuclide (Days) Ci/ml) Ci/ml) (Curies) (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) (Ci/yr)

PR143 1.37+01 4.44E-05 5.51E-09 0.00006 0.00022 0.00000 0.00000 0.00028 0.00028 0.00000 0.00028 CE144 2.84E+04 2.91E-05 5.06E-09 0.00006 0.00015 0.00000 0.00000 0.00020 0.00024 0.00187 0.00210 PR144 1.20E-02 3.53E-05 3.41E-08 0.00006 0.00015 0.00000 0.00000 0.00021 0.00021 0.00000 0.00021 All Others 3.22E-04 3.53E-10 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 Total (Except Tritium) 1.41E+00 1.67E-04 0.31537 20.35739 0.02963 0.00748 20.71086 20.86086 0.02244 21.00000 Tritium Release 480 Curies Per Year

a. Steam generator blowdown treatment system is no longer used.

SPS UFSAR 11A-22

Revision 52Updated Online 09/30/20 Table 11A-11 LADTAP INPUT DATA AND RESULTS - MAXIMALLY EXPOSED INDIVIDUAL DOSE CALCULATIONS FOR SURRY UNITS 1 AND 2 (DEMINERALIZER RADWASTE SYSTEM WITHOUT STEAM GENERATOR BLOWDOWN TREATMENT)

Usage Rates (kg/yr or hr/yr)

Exposure Pathway Dilution Factor Transit Time (hr) Adult Teen Child Fish ingestion 5 24 21.0 16.0 6.9 Invertebrate ingestion 5 24 5.0 3.8 1.7 Shoreline use 5 0 12.0 67.0 14.0 Dose Results (mrem/yr per unit)a Adults Teenagers Exposure Pathway Total Body GI-LLI Skin Total Body GI-LLI Skin b

Fish ingestion 1.15 (-1) 1.74 (-2) - 6.63 (-2) 1.27 (-2) -

Invertebrate ingestion 6.12 (-2) 2.02 - 5.47 (-2) 1.59 -

Shoreline use 3.28 (-3) 3.28 (-3) 3.82 (-3) 1.83 (-2) 1.83 (-2) 2.13 (-2) 1.79 (-1) 2.04 3.82 (-3) 1.39 (-1) 1.62 2.13 (-2)

a. Doses to other individuals and organs are smaller than those presented.

SPS UFSAR

b. 1.15 (-1) = 1.15 x 10-1.

11A-23

Revision 52Updated Online 09/30/20 Table 11A-12 LADTAP INPUT DATA AND RESULTS - MAXIMALLY EXPOSED INDIVIDUAL DOSE CALCULATIONS FOR SURRY UNITS 1 AND 2 (MODIFIED LIQUID RADWASTE SYSTEM WITH STEAM GENERATOR BLOWDOWN TREATMENT)a Usage Rates (kg/yr or hr/yr)

Exposure Pathway Dilution Factor Transit Time (hr) Adult Teen Child Fish ingestion 5 24 21.0 16.0 6.9 Invertebrate ingestion 5 24 5.0 3.8 1.7 Shoreline use 5 0 12.0 67.0 14.0 Dose Results (mrem/yr per unit)b Adults Teenagers Exposure Pathway Total Body GI-LLI Skin Total Body GI-LLI Skin c

Fish ingestion 1.06 (-1) 7.22 (-3) - 6.06 (-2) 5.16 (-3) -

Invertebrate ingestion 2.81 (-2) 5.60 (-1) - 2.15 (-2) 4.40 (-1) -

Shoreline use 3.06 (-3) 3.06 (-3) 3.57 (-3) 1.71 (-2) 1.71 (-2) 2.00 (-2) 1.37 (-1) 5.70 (-1) 3.57 (-3) 9.92 (-2) 4.62 (-1) 2.00 (-2)

a. Steam generator blowdown treatment system is no longer used.

SPS UFSAR

b. Doses to other individuals and organs are smaller than those presented.

c. 1.06 (-1) = 1.06 x 10-1.

11A-24

Revision 52Updated Online 09/30/20 Table 11A-13 GASEOUS SOURCE TERMS FROM SURRY UNITS 1 AND 2 (PER UNIT)

Gaseous Release Rate - Curies Per Year Primary Secondary Gas Stripping Building Ventilation Coolant Coolant Blowdown Air Ejector (Micro Ci/Gm) (Micro Ci/Gm) Shutdown Continuous Reactor Auxiliary Turbine Vent Offgas Exhaust Total Kr-83m 2.251E-02 8.325E-09 7.0E+00 0.0 0.0 0.0 0.0 0.0 0.0 7.0E+00 Kr-85m 1.173E-01 4.426E-08 3.9E+01 7.0E+00 1.0E+00 2.0E+00 0.0 0.0 2.0E+00 5.1E+01 Kr-85 6.233E-02 2.337E-08 2.1E+01 2.5E+02 3.0E+01 1.0E+00 0.0 0.0 0.0 3.0E+02 Kr-87 6.438E-02 2.299E-08 2.1E+01 1.0E+00 0. 1.0E+00 0.0 0.0 0.0 2.3E+01 Kr-88 2.139E-01 7.879E-08 7.1E+01 8.0E+00 2.0E+00 5.0E+00 0.0 0.0 3.0E+00 8.9E+01 Kr-89 5.379-03 2.017E-09 2.0E+00 0.0 0.0 0.0 0.0 0.0 0.0 2.0E+00 Xe-131m 8.252E-02 3.115E-08 2.7E+01 2.1E+02 2.9E+01 2.0E+00 0.0 0.0 1.0E+00 2.7E+02 Xe-133m 2.139E-01 8.074E-08 7.1E+01 1.6E+02 2.8E+01 5.0E+00 0.0 0.0 3.0E+00 2.7E+02 Xe-133 1.577E+01 5.865E-08 5.8E+03 2.5E+04 3.9E+03 3.3E+02 0.0 0.0 2.1E+02 3.5E+04 Xe-135m 1.398E-02 5.185E-09 5.0E+00 0.0 0.0 0.0 0.0 0.0 0.0 5.0E+00 SPS UFSAR Xe-135 3.695E-01 1.371E-07 1.2E+02 4.7E+01 9.0E+00 8.0E+00 0.0 0.0 5.0E+00 1.9E+02 Xe-137 9.662E-03 3.602E-09 3.0E+00 0.0 0.0 0.0 0.0 0.0 0.0 3.0E+00 Xe-138 4.731E-02 1.728E-08 1.6E+01 0.0 0.0 1.0E+00 0.0 0.0 0.0 1.7E+01 Total Noble Gases 3.6E+04 I-131 2.408E-01 3.207E-05 0.0 0.0 2.8E-03 3.8E-03 1.7E-03 1.6E-01 2.4E-02 1.9E-01 I-133 3.582E-01 3.611E-05 0.0 0.0 7.9E-04 5.7E-03 2.0E-03 1.8E-01 3.6E-02 2.2E-01 Tritium Gaseous Release 490 Ci/yr 0.0 appearing in the table indicates release is less than 1.0 Ci/yr for noble gas, 0.0001 Ci/yr for I AR-41 25 Curies/yr 11A-25 C-14 8 Curies/yr

Table 11A-13 (CONTINUED)

Revision 52Updated Online 09/30/20 GASEOUS SOURCE TERMS FROM SURRY UNITS 1 AND 2 (PER UNIT)

Airborne Particulate Release Rate - Curies Per Year Waste Gas System Building Ventilation Nuclide Reactor Auxiliary Total Mn-54 4.5E-05 6.0E-05 1.8E-04 2.9E-04 Fe-59 1.5E-05 2.1E-05 6.0E-05 9.6E-05 Co-58 1.5E-04 2.1E-04 6.0E-04 9.6E-04 Co-60 7.0E-05 9.3E-05 2.7E-04 4.3E-04 Sr-89 3.3E-06 4.7E-06 1.3E-05 2.1E-05 Sr-90 6.0E-07 8.2E-07 2.4E-06 3.8E-06 Cs-134 4.5E-05 6.0E-05 1.8E-04 2.9E-04 Cs-137 7.5E-05 1.0E-04 3.0E-04 4.8E-04 SPS UFSAR 11A-26

Revision 52Updated Online 09/30/20 Table 11A-14 MAXIMUM DOSES TO AN INDIVIDUAL RESULTING FROM GASEOUS EFFLUENTS FROM SURRY UNITS 1 AND 2 WITH STEAM GENERATOR FLASH TANK (MREM/YR PER UNIT)

Location 2.17 miles SSW Location 3.75 miles NNW Total Body Organ Dose (thyroid)

A. Radioiodines and particulatesa Ground 7.80 (-4)b 3.11 (-4)

Ingestion of vegetables 5.01 (-2) -

Inhalation 3.31 (-3) 9.96 (-2)

Milk - 1.67 5.42 (-2) 1.77 Total Body Skin B. Noble gases Plume (1.53 miles S) 1.32 (-1) 3.53 (-1)

C. Air doses (1.53 miles S) Annual beta 5.67 (-1) mrad/yr Annual gamma 2.20 (-1) mrad/yr (Site boundary 0.31 miles N) Annual beta 15.4 mrad/yr Annual gamma 6.26 mrad/yr SPS UFSAR

a. Maximum organ dose occurs to an infant.

b. 7.80 (-4) = 7.80 x 10-4 11A-27

Revision 52Updated Online 09/30/20 Table 11A-15 MAXIMUM DOSES TO AN INDIVIDUAL RESULTING FROM GASEOUS EFFLUENTS FROM SURRY UNITS 1 AND 2 WITH STEAM GENERATOR BLOWDOWN COOLED (MREM/YR PER UNIT)

Location 2.17 miles SSW Location 3.75 miles NNW Total Body Organ Dose (thyroid)

D. Radioiodines and particulatesa Ground 7.09 (-4)b 2.81 (-4)

Ingestion of vegetables 4.97 (-2) -

Inhalation 3.25 (-3) 2.08 (-2)

Milk - 3.00 (-1) 5.37 (-2) 3.21 (-1)

Total Body Skin E. Noble gases Plume (1.53 miles S) 1.32 (-1) 3.53 (-1)

F. Air doses (1.53 miles S) Annual beta 5.67 (-1) mrad/yr Annual gamma 2.20 (-1) mrad/yr (Site boundary 0.31 miles N) Annual beta 15.4 mrad/yr Annual gamma 6.26 mrad/yr SPS UFSAR

a. Maximum organ dose occurs to an infant.

b. 7.09 (-4) = 7.09 x 10-4 11A-28