ML16166A136

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Updated Final Safety Analysis Report, Chapter 11, Radioactive Waste Management
ML16166A136
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Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 05/25/2016
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South Carolina Electric & Gas Co
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Office of Nuclear Reactor Regulation
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RC-16-0083
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{{#Wiki_filter:11-i Reset May 2013 TABLE OF CONTENTS Section Title Page 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1-1 11.1 SOURCE TERMS 11.1-1 11.1.1 RADIOACTIVITIES IN SYSTEMS AND COMPONENTS, CONSERVATIVE MODEL 11.1-1 11.1.1.1 Reactor Coolant Activity 11.1-1 11.1.1.2 Gaseous Waste Processing System Activity 11.1-3 11.1.1.3 Volume Control Tank Activity 11.1-4 11.1.1.4 Pressurizer Activity 11.1-4 11.1.1.5 Liquid Waste Processing System Activity 11.1-4 11.1.1.6 Solid Waste 11.1-4 11.1.2 LEAKAGE RATES 11.1-4 11.1.3 RADIOACTIVITIES IN THE FLUID SYSTEMS, REALISTIC MODEL 11.1-5 11.1.4 TRITIUM 11.1-5 11.1.4.1 System Sources 11.1-5 11.1.4.2 Design Bases 11.1-6 11.1.4.3 Design Evaluation 11.1-7 11.

1.5 REFERENCES

11.1-7 11.2 LIQUID WASTE SYSTEMS 11.2-1 11.2.1 DESIGN OBJECTIVES 11.2-1 11.2.2 SYSTEMS DESCRIPTIONS 11.2-1 11.2.2.1 Waste Holdup Tank 11.2-2 11.2.2.2 Floor Drain Tank 11.2-3 11.2.2.3 Laundry and Hot Shower Tank 11.2-3 11.2.2.4 Excess Liquid Waste Processing System (ELWS) 11.2-3 11.2.2.5 Laboratory Drain System 11.2-4 11.2.2.6 Waste from Spent Resin 11.2-4 11.2.3 SYSTEM DESIGN 11.2-4 11.2.3.1 Component Design 11.2-4 11.2.3.2 Instrumentation Design 11.2-11 11.2.3.3 Tank Overflow Protection 11.2-11 11.2.4 OPERATING PROCEDURES 11.2-11 11.2.4.1 Normal Operation 11.2-12 11.2.4.2 Faults of Moderate Frequency 11.2-15 11.2.5 PERFORMANCE TESTS 11.2-16 11.2.6 ESTIMATED RELEASES 11.2-16 11.2.6.1 Reactor Grade Demineralizers 11.2-16 11.2.6.2 Liquid Waste Processing System 11.2-17 11.2.6.3 Detergent Wastes 11.2-18 11.2.6.4 Secondary System 11.2-18 11.2.6.5 Adjustments to Liquid Radwaste Source Term for Anticipated Operational Occurrences 11.2-19 11.2.6.6 Criteria for Reuse, Discharge and Recycle 11.2-20 11.2.7 RELEASE POINTS 11.2-20 11.2.8 DILUTION FACTORS 11.2-21 11.2.9 ESTIMATED DOSES 11.2-22 11.2.10 REFERENCES 11.2.23 RN 03-038 RN 03-038

11-ii Reset May 2013 TABLE OF CONTENTS (Continued) Section Title Page 11.3 GASEOUS WASTE SYSTEM 11.3-1 11.3.1 DESIGN OBJECTIVES 11.3-1 11.3.2 SYSTEM DESCRIPTION 11.3-1 11.3.3 SYSTEM DESIGN 11.3-3 11.3.3.1 Component Design 11.3-3 11.3.3.2 Instrumentation and Control Design 11.3-4 11.3.4 OPERATING PROCEDURES 11.3-6 11.3.4.1 General Description 11.3-6 11.3.4.2 Startup Operation 11.3-6 11.3.4.3 Normal Operations 11.3-7 11.3.4.4 Shutdown 11.3-7 11.3.5 PERFORMANCE TESTS 11.3-7 11.3.6 ESTIMATED RELEASES 11.3-8 11.3.6.1 Gaseous Waste Processing System 11.3-8 11.3.6.2 Reactor Building Purge 11.3-8 11.3.6.3 Auxiliary Building Ventilation 11.3-9 11.3.6.4 Secondary System 11.3-9 11.3.6.5 Release Criteria 11.3-10 11.3.7 RELEASE POINTS 11.3-10 11.3.8 DILUTION FACTORS 11.3-10 11.3.9 ESTIMATED DOSES 11.3-11 11.3.10 REFERENCES 11.3-13 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.4-1 11.4.1 DESIGN OBJECTIVES 11.4-1 11.4.2 CONTINUOUS MONITORING 11.4-2 11.4.3 SAMPLING 11.4-10 11.4.4 INSERVICE INSPECTION, CALIBRATION AND MAINTENANCE 11.4-12 11.

4.5 REFERENCES

11.4-13 11.5 SOLID WASTE SYSTEM 11.5-1 11.5.1 DESIGN OBJECTIVES 11.5-1 11.5.2 SYSTEM INPUTS 11.5-1 11.5.3 EQUIPMENT DESCRIPTION 11.5-2 11.5.3.1 Processing 11.5-2 11.5.3.2 Equipment 11.5-2 11.5.4 EXPECTED VOLUMES 11.5-6 11.5.4.1 Activity Levels 11.5-6 11.5.4.2 Processed Wastes 11.5-7 11.5.4.3 Filter Cartridges 11.5-7 11.5.4.4 Miscellaneous Solid Wastes 11.5-7

11-iii Reset May 2013 TABLE OF CONTENTS (Continued) Section Title Page 11.5.5 PACKAGING 11.5-7 11.5.5.1 Evaporator Bottoms and Chemical Samples -No Longer In Service-11.5-7 11.5.5.2 Spent Resin 11.5-7 11.5.5.3 Filter Disposal 11.5-9 11.5.5.4 Radioactive Hardware 11.5-10 11.5.5.5 Compacted Wastes 11.5-10 11.5.6 STORAGE 11.5-10 11.5.7 SHIPMENT 11.5.11 11.5.8 POTENTIAL FOR RELEASES 11.5-11 11.5.8.1 Potential for Release during Container Filling 11.5-11 11.5.8.2 Potential for Release from Storage Tanks 11.5-12 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 11.6-1 11.7 Deleted by RN 12-005 RN 08-003 RN 12-005

11-iv Reset May 2013 LIST OF TABLES Table Title Page 11.1-1 Parameters used in the Calculation of Reactor Coolant Fission and Corrosion Product Activities 11.1-8 11.1-2 Reactor Coolant Equilibrium Fission and Corrosion Product Activities 11.1-11 11.1-3 Volume Control Tank Equilibrium Activities 11.1-12 11.1-4 Pressurizer Activities 11.1-13 11.1-5 Normal Plant Operation Source Terms 11.1-14 11.1-6 Parameters used to Describe the Reactor System - Realistic Model 11.1-17 11.1-7 Tritium Production 11.1-18 11.2-1 Isotopic Values in the Liquid Waste Processing System 11.2-24 11.2-2 Equipment Principal Design Parameters 11.2-28 11.2-3 Liquid Waste Processing System Major Component Inventories 11.2-39 11.2-4 Range of Measured Demineralizer Decontamination Factors for Selected Isotopes 11.2-42 11.2-5 Liquid Waste Processing System Instrumentation Principal Design Parameters 11.2-43 11.2-6 Tank Overflow Protection 11.2-49 11.2-6a Comparison of Tanks Outside Containment With Provisions of Branch Technical Position ETSB 11-1 (Rev. 1), Paragraph B.1.b 11.2-51 11.2-7 Parameters Used in the Calculation of Estimated Activity in Liquid Wastes 11.2-54 11.2-8 PWR-GALE Code Input Parameters Used in Calculating Releases of Radioactive Materials in Liquid Effluents 11.2-55 11.2-9 Deleted by RN 03-038 11.2-10 Deleted by RN 03-038 11.2-11 Deleted by RN 03-038 11.2-12 Liquid Effluents Annual Releases 11.2-57 11.2-13 Comparison of Radionuclide Concentrations in Liquid Effluents to the Limits of 10 CFR 20 11.2-58 11.2-14 Summary of Calculated Liquid Pathway Doses Virgil C. Summer Nuclear Station 11.2-59 11.2-15 Appendix I Conformance Summary Table Virgil C. Summer Nuclear Station Liquid Effluents 11.2-60 02-01 02-01 99-01 RN 03-038 RN 03-038

11-v Reset May 2013 LIST OF TABLES (Continued) Table Title Page 11.3-1 Design Basis Accumulated Radioactivity per Unit in the Gaseous Waste Processing System after Forty Years Operation 11.3-14 11.3-2 Expected Accumulated Radioactivity per Unit in the Gaseous Waste Processing System after Forty Years Operation 11.3-15 11.3-3 Reduction in Reactor Coolant System Gaseous Fission Products Resulting from Normal Operation of the Gaseous Waste Processing System 11.3-16 11.3-4 Process Parameters for Gaseous Waste Processing System 11.3-17 11.3-5 Gaseous Waste Processing System Component Data 11.3-21 11.3-6 Gaseous Waste Processing System Instrumentation Design Parameters 11.3-22 11.3-7 PWR-GALE Code Input Parameters Used In Calculating Releases of Radioactive Materials in Gaseous Effluents 11.3-25 11.3-8 Calculated Releases of Radioactive Materials in Gaseous Effluents from the Plant 11.3-26 11.3-9 Stack Release Information 11.3-27 11.3-9a Comparison of Normal Ventilation Exhaust System Air Filtration and Adsorption Units With Branch Technical Position ESTB 11-2. 11.3-30 11.3-10 Comparison of Radionuclide Concentrations in Gaseous Effluents to the Limits of 10 CFR 20 11.3-35 11.3-11 Summary of Calculated Gaseous Pathway Doses Virgil C. Summer Nuclear Station 11.3-36 11.3-12 Appendix I Conformance Summary Table Virgil C. Summer Nuclear Station Gaseous Effluents 11.3-37 11.4-1 Process and Effluent Radiological Monitors 11.4-14 11.4-2 Discharge Monitoring and Analysis 11.4-17 11.5-1 Spent Resin Volumes 11.5-13 11.5-2 Anticipated Total Solid Waste Generated per Year 11.5-14 11.5-3 Maximum Expected Concentrations of Waste to be Packaged 11.5-15 11.5-4 Maximum Expected Activities of Expended Filter Cartridge 11.5-16 11.5-5 Solid Waste System Equipment Design Parameters 11.5-17 11.5-6 Solid Radioactive Waste Processed from Westinghouse Designed Operating Reactors 11.5-18 11.5-7 Valves (GAI); Valves (Westinghouse); Equipment; Piping (GAI); Piping (Westinghouse) 11.5-19 02-01

11-vi Reset May 2013 LIST OF FIGURES Figure No. Title 11.2-1 Liquid Waste Processing System Process Flow Diagram 11.2-2 Deleted by RN 03-038 11.2-3 Location of Liquid Release Points 11.2-4 Fairfield Pumped Storage Facility 11.2-5 Duratek System Typical Lineup 11.3-1 Gaseous Waste Processing System Fission Gas Accumulation Based on Continuous Core Operation at 2958 MWt with 1% Fuel Defects and 60 gpm Continuous Letdown 11.3-2 Estimated Gaseous Waste Processing System Fission Gas Accumulation Based on Table 11.1-5 and Full Power Operation at 2958 MWt and 60 gpm Continuous Letdown 11.3-3 Gaseous Waste Processing System Process Flow Diagram 11.3-4 Piping System Flow Diagram - Waste Processing (3 Sheets) 11.3-5 Waste Gas Compressor Package 11.3-6 Catalytic Hydrogen Recombiner Package 11.3-7 Gaseous Waste Release Points 11.3-8 Potentially Radioactive Gaseous Waste Release Points 11.4-1 Radiation Monitoring System Interlocks 11.4-2 Location of High Range Effluent Monitors RM-A13 and RM-A14 11.5-1 Deleted by RN 08-003 RN 02-025 RN 08-003 RN 03-038 RN 03-038

LIST OF EFFECTIVE PAGES (LEP) The following list delineates pages to Chapter 11 of the Virgil C. Summer Nuclear Station Final Safety Analysis Report which are current through May 2013. The latest changes to pages and figures are indicated below by Revision Number (RN) in the Amendment column along with the Revision Number and date for each page and figure included in the Final Safety Analysis Report. Page/Fig. No. Amend. No. Date Page/Fig.No. Amend. No. Date 11-vii Reset May 2013 Page 11-i Reset May 2013 11-ii Reset May 2013 11-iii Reset May 2013 11-iv Reset May 2013 11-v Reset May 2013 11-vi Reset May 2013 11-vii Reset May 2013 11-viii Reset May 2013 11-ix Reset May 2013 Page 11.1-1 99-01 June 1999 11.1-2 99-01 June 1999 11.1-3 99-01 June 1999 11.1-4 99-01 June 1999 11.1-5 99-01 June 1999 11.1-6 99-01 June 1999 11.1-7 99-01 June 1999 11.1-8 99-01 June 1999 11.1-9 99-01 June 1999 11.1-10 99-01 June 1999 11.1-11 99-01 June 1999 11.1-12 99-01 June 1999 11.1-13 99-01 June 1999 11.1-14 99-01 June 1999 11.1-15 99-01 June 1999 11.1-16 99-01 June 1999 11.1-17 99-01 June 1999 11.1-18 99-01 June 1999 Page 11.2-1 RN03-038 November 2011 11.2-2 RN03-038 November 2011 11.2-3 RN03-038 November 2011 11.2-4 RN03-038 November 2011 11.2-5 RN03-038 November 2011 11.2-6 RN03-038 November 2011 Page 11.2-7 RN03-038 November 2011 11.2-8 RN03-038 November 2011 11.2-9 RN03-038 November 2011 11.2-10 RN03-038 November 2011 11.2-11 RN03-038 November 2011 11.2-12 RN03-038 November 2011 11.2-13 RN03-038 November 2011 11.2-14 RN03-038 November 2011 11.2-15 RN03-038 November 2011 11.2-16 RN03-038 November 2011 11.2-17 RN03-038 November 2011 11.2-18 RN03-038 November 2011 11.2-19 RN03-038 November 2011 11.2-20 RN03-038 November 2011 11.2-21 RN03-038 November 2011 11.2-22 RN03-038 November 2011 11.2-23 RN03-038 November 2011 11.2-24 RN03-038 November 2011 11.2-25 RN03-038 November 2011 11.2-26 RN03-038 November 2011 11.2-27 RN03-038 November 2011 11.2-28 RN03-038 November 2011 11.2-29 02-01 May 2002 11.2-30 02-01 May 2002 11.2-31 02-01 May 2002 11.2-32 97-01 August 1997 11.2-33 RN03-038 November 2011 11.2-34 RN03-038 November 2011 11.2-35 RN03-038 November 2011 11.2-36 RN03-038 November 2011 11.2-37 97-01 August 1997 11.2-38 RN03-038 November 2011 11.2-39 RN03-038 November 2011

LIST OF EFFECTIVE PAGES (LEP) Page/Fig.No. Amend. No. Date Page/Fig.No. Amend. No. Date 11-viii Reset May 2013 Page 11.2-40 RN03-038 November 2011 11.2-41 RN03-038 November 2011 11.2-42 02-01 May 2002 11.2-43 02-01 May 2002 11.2-44 RN03-038 November 2011 11.2-45 02-01 May 2002 11.2-46 02-01 May 2002 11.2-47 RN04-037 October 2005 11.2-48 02-01 May 2002 11.2-49 RN03-038 November 2011 11.2-50 RN03-038 November 2011 11.2-51 02-01 May 2002 11.2-52 02-01 May 2002 11.2-53 02-01 May 2002 11.2-54 RN03-038 November 2011 11.2-55 RN03-038 November 2011 11.2-56 RN03-038 November 2011 11.2-57 RN03-038 November 2011 11.2-58 RN03-038 November 2011 11.2-59 RN03-038 November 2011 11.2-60 RN03-038 November 2011 Fig. 11.2-1 RN03-038 November 2011 11.2-2 (Sh. 1) deleted November 2011 11.2-2 (Sh. 2) deleted November 2011 11.2-2 (Sh. 3) deleted November 2011 11.2-2 (Sh. 4) deleted November 2011 11.2-2 (Sh. 5) deleted November 2011 11.2-3 0 August 1984 11.2-4 0 August 1984 11.2-5 RN03-038 November 2011 Page 11.3-1 00-01 December 2000 11.3-2 RN02-025 July 2002 11.3-3 00-01 December 2000 11.3-4 00-01 December 2000 11.3-5 02-01 May 2002 11.3-6 02-01 May 2002 Page 11.3-7 RN03-023 March 2004 11.3-8 RN02-028 July 2002 11.3-9 RN02-028 July 2002 11.3-10 RN02-028 July 2002 11.3-11 RN02-028 July 2002 11.3-12 00-01 December 2000 11.3-13 00-01 December 2000 11.3-14 RN02-025 July 2002 11.3-15 RN02-025 July 2002 11.3-16 RN02-025 July 2002 11.3-17 02-01 May 2002 11.3-18 02-01 May 2002 11.3-19 02-01 May 2002 11.3-20 02-01 May 2002 11.3-21 02-01 May 2002 11.3-22 RN07-010 May 2007 11.3-23 02-01 May 2002 11.3-24 02-01 May 2002 11.3-25 RN02-028 July 2002 11.3-26 02-01 May 2002 11.3-27 02-01 May 2002 11.3-28 02-01 May 2002 11.3-29 02-01 May 2002 11.3-30 02-01 May 2002 11.3-31 02-01 May 2002 11.3-32 02-01 May 2002 11.3-33 02-01 May 2002 11.3-34 RN02-034 May 2003 11.3-35 RN02-028 July 2002 11.3-36 02-01 May 2002 11.3-37 02-01 May 2002 11.3-38 02-01 May 2002 Fig. 11.3-1 RN02-025 July 2002 11.3-2 RN02-025 July 2002 11.3-3 0 August 1984 11.3-4 (Sh. 1) RN04-020 February 2005

LIST OF EFFECTIVE PAGES (LEP) Page/Fig.No. Amend. No. Date Page/Fig.No. Amend. No. Date 11-ix Reset May 2013 Fig. 11.3-4 (Sh. 2) RN05-041 October 2005 11.3-4 (Sh. 3) RN09-011 May 2009 11.3-5 0 August 1984 11.3-6 0 August 1984 11.3-7 0 August 1984 11.3-8 0 August 1984 Page 11.4-1 02-01 May 2002 11.4-2 RN07-037 November 2011 11.4-3 02-01 May 2002 11.4-4 02-01 May 2002 11.4-5 RN07-037 November 2011 11.4-6 RN03-011 April 2006 11.4-7 RN02-019 June 2003 11.4-8 RN03-011 April 2006 11.4-9 RN03-011 April 2006 11.4-10 02-01 May 2002 11.4-11 02-01 May 2002 11.4-12 02-01 May 2002 11.4-13 02-01 May 2002 11.4-14 RN07-037 November 2011 11.4-15 02-01 May 2002 11.4-16 RN03-011 April 2006 11.4-17 02-01 May 2002 11.4-18 02-01 May 2002 Fig. 11.4-1 RN07-037 November 2011 11.4-2 0 August 1984 Page 11.5-1 RN08-003 January 2011 11.5-2 RN08-003 January 2011 11.5-3 RN08-003 January 2011 11.5-4 RN08-003 January 2011 11.5-5 RN08-003 January 2011 11.5-6 RN08-003 January 2011 11.5-7 RN08-003 January 2011 11.5-8 RN08-003 January 2011 11.5-9 RN08-003 January 2011 11.5-10 RN08-003 October 2010 RN10-022 November 2011 Page 11.5-11 RN08-003 January 2011 11.5-12 02-01 May 2002 11.5-13 RN08-003 January 2011 11.5-14 RN08-003 January 2011 11.5-15 RN08-003 January 2011 11.5-16 RN08-003 January 2011 11.5-17 RN08-003 January 2011 11.5-18 RN08-003 January 2011 11.5-19 RN08-003 January 2011 11.5-20 RN08-003 January 2011 Fig. 11.5-1 Deleted January 2011 Page 11.6-1 97-01 August 1997

11.1-1 Reformatted Per Amendment 99-01 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS The fission product inventory in the reactor core and the diffusion to the fuel pellet/cladding gap are presented in Chapter 15. Source terms and models used in the evaluation of radwaste treatment systems and effluent releases are based on operating plant data where available[1]. Two (2) source terms are presented. The first is a conservative model which utilizes a conventional fuel clad defect model. This conservative model serves as a basis for calculations of the maximum offsite doses resulting from postulated accidents. The second source term is a realistic model used to predict expected long term average concentrations of radionuclides and expected releases. This realistic model is based on ANSI/ANS-18.1-1984. 11.1.1 RADIOACTIVITIES IN SYSTEMS AND COMPONENTS, CONSERVATIVE MODEL 11.1.1.1 Reactor Coolant Activity The parameters used in the calculation of the reactor coolant fission product inventories together with the pertinent information concerning the expected coolant cleanup flowrate and demineralizer effectiveness, are summarized in Table 11.1-1. Calculated reactor coolant radionuclide concentrations, based on the assumptions of Table 11.1-1, are presented in Table 11.1-2. In these calculations the defective fuel rods are assumed to be present at the initial core loading and to be uniformly distributed throughout the core; thus, the fission product escape rate coefficients are based upon average fuel temperature. For fuel failure and burnup experience see Chapter 4. The fission product activities in the reactor coolant during operation with small cladding defects (fuel rods containing pinholes or fine cracks) are computed using the following differential equations: For parent nuclides in the coolant: dNc dt Nc i i          


  RN M D Q M DF DF i Fi C i i L C i i i   1 (11.1-1 ) 99-01 99-01

11.1-2 Reformatted Per Amendment 99-01 For daughter nuclides in the coolant: dNc dt Nc Nc j j i j j j j j j            


   RN M f D Q M DF DF Fj C i i L C    1 (11.1-2 ) Where: NC = Concentration of nuclide in the reactor coolant (atoms/gram) NF = Inventory of nuclide in the fuel (atoms) t = Operating time(seconds) R = Nuclide release coefficient (1/sec) = F v F = Fraction of fuel rods with defective cladding v = Fission product escape rate coefficient (1/sec) MC = Mass of reactor coolant (grams)  = Nuclide decay constant (1/sec) DF = Nuclide demineralizer decontamination factor QL = Purification or letdown mass flow rate (grams/sec)  = Nuclide volume control tank stripping fraction f = Fraction of parent nuclide decay events that result in the formation of the daughter nuclide D = Dilution coefficient for feed and bleed (1/sec) =   B - O t DF

  • 1 BO

= Initial boron concentration (ppm)  = Boron concentration reduction rate (ppm/sec) and where subscript i refers to the parent nuclide. subscript j refers to the daughter nuclide. The above equations are based on the assumption that there is no activity reduction due to pressurizer operation and that the nuclide concentration in the volume control tank can be approximated by: N DF N VL C   1 

  • where NVL = concentration of nuclide in the volume control tank (atoms/gram) 99-01 99-01

11.1-3 Reformatted Per Amendment 99-01 The corrosion product activities in the Reactor Coolant System are based on measurements at operating reactors[1]. The reactor coolant inventories of corrosion products (which are independent of fuel defect level) are given in Table 11.1-2. Another potential source of primary coolant activity is activation products from the silver-indium-cadmium control rods. The presence of Ag-110m has been noted in the primary coolant or plant discharges at a few of the operating plants. Several mechanisms can contribute to the presence of this isotope; namely, fission product decay of the mass 109 chain followed by activation to Ag-110m, surface contamination of the control rods with absorber material, or escape of activation products from the control rods. The control rods are designed to contain products formed in the absorber material and are not expected to contribute to the coolant activity. Current information, however, is insufficient to determine the source of the silver, and thus no model is available for predicting coolant activity levels due to the above sources. Investigations of this will be continued as more data becomes available. 11.1.1.2 Gaseous Waste Processing System Activity The stripping fractions used in determining the amount of fission gases removed from the reactor coolant in the volume control tank and collected by the Gaseous Waste Processing System (GWPS) are calculated as follows:             


1 KQ KQ KL V P (11.1-3) Where:  = Nuclide volume control tank stripping fraction K = RT / MH R = Gas constant = 45.59 atm/cc per gram-mole /R T = Nominal volume control tank temperature (R) M = Molecular weight of water = 18.0 grams/gram-mole H = Henrys Law constant Q = Letdown or purification flow rate (grams/sec)  = Nuclide decay constant (1/sec) L = Volume control tank liquid mass (grams) V = Volume control tank vapor volume (cc) P = Volume control tank purge rate to the gaseous waste processing system (cc/sec at volume control tank conditions) An activity balance is performed on the Reactor Coolant System and volume control tank to obtain the Reactor Coolant System activity, volume control tank activity, and stripping fraction. Stripping fractions are shown in Table 11.1-1. Gaseous waste source terms are discussed in section 11.3. 99-01

11.1-4 Reformatted Per Amendment 99-01 11.1.1.3 Volume Control Tank Activity Table 11.1-3 lists the maximum activities in the volume control tank using the assumptions summarized in Table 11.1-1. The liquid activity is assumed to be the same as the letdown coolant activity for the halogen and particulate activity. 11.1.1.4 Pressurizer Activity The specific activity for major nuclides in the pressurizer are discussed below.  Pressurizer Liquid Phase Source Strengths - The pressurizer liquid specific activity is assumed to be the same as that of the reactor coolant. Table 11.1-4 lists only those nuclides that are the major contributors to total source strength.  Pressurizer Steam Phase - Pressurizer steam phase radiogas concentrations (Table 11.1-4) are based on the stripping of radiogases from the continuous 2-gpm pressurizer spray and the subsequent buildup of these radiogases in the steam space. The buildup time is assumed to be 480 days. Decay credit has been taken during spray line transit. The radiogases are assumed to be completely stripped from the spray, except for Kr-85 and Xe-133, which are in Henrys Law Equilibrium with the liquid in the pressurizer. Pressurizer steam phase iodine concentrations are obtained from the liquid phase nuclide activities and measured values of the partition coefficient for I-131. A large partition coefficient was chosen to maximize the activities. It was assumed to apply to all radioiodines. The activities in the pressurizer are separated between the liquid and the steam phase and the results obtained are given in Table 11.1-4 using the above assumptions and those summarized in Table 11.1-1. 11.1.1.5 Liquid Waste Processing System Activity Liquid waste source terms are discussed in Section 11.2. 11.1.1.6 Solid Waste Solid waste source terms are discussed in Section 11.5. 11.1.2 LEAKAGE RATES As a necessary part of the effort to reduce effluent of radioactive liquid wastes, Westinghouse has been surveying various Pressurized Water Reactor (PWR) facilities which are in operation, to identify design and operating problems influencing reactor coolant and non-reactor grade leakage and hence the load on the LWPS. 99-01 99-01 99-01 99-01

11.1-5 Reformatted Per Amendment 99-01 Leakage sources have been identified in connection with pump shaft seals and valve stem leakage. Where packed glands are provided, a leakage problem may be anticipated, while mechanical shaft seals provide essentially 0 leakage. Valve stem leakage was experienced where the originally specified packing was used. A combination of a graphite filament yarn packing sandwiched with asbestos sheet packing is used with improved results in several plants. A bellows seal is being utilized in later plants which eliminates all stem leakage. In addition, seat leakage was experienced on some pressurized power operated relief valves. However, this was found to be due to a manufacturing error and has been corrected. 11.1.3 RADIOACTIVITIES IN THE FLUID SYSTEMS, REALISTIC MODEL The parameters used to describe the Virgil C. Summer Nuclear Station reactor are given in Table 11.1-6. Specific activities in the primary coolant, steam generator water, and steam generator steam are based on the parameters of Table 11.1-6 and are given in Table 11.1-5. 11.1.4 TRITIUM The release of tritium to the environment from operating Westinghouse PWRs has always been well below 10 CFR 20 limits. This Section discusses the reduced tritium production in the plant as a result of employing Zircaloy clad fuel and silver-indium-cadmium control rods. 11.1.4.1 System Sources There are 2 principal contributors to tritium production within the PWR system: the ternary fission source, and the dissolved boron in the reactor coolant. Additional contributions are made by Li6, Li7, and deuterium in the reactor water. Tritium production from various sources is shown in Table 11.1-7. 11.1.4.1.1 Fission Source This tritium is formed within the fuel material and may: 1. Remain in the fuel rod uranium matrix, 2. Diffuse into the cladding and become hydrided and fixed there, 3. Diffuse through the clad for release into the primary coolant, 4. Release to the coolant through macroscopic cracks or failures in the fuel cladding. 99-01 99-01 99-01

11.1-6 Reformatted Per Amendment 99-01 Previous Westinghouse designs conservatively assumed that the ratio of fission tritium released into the coolant to the total fission tritium formed was approximately 0.30 for Zircaloy clad fuel. The operating experience at the Robert Emmett Ginna Station of the Rochester Gas and Electric Corporation, and at other operating reactors using Zircaloy clad fuel, has shown that the tritium release through the Zircaloy fuel cladding is less than the earlier estimates. Consequently, a tritium release rate into the primary coolant of 0.001 curies per megawatt-day or less can be anticipated. 11.1.4.1.2 Control Rod Source The full length rods for the Virgil C. Summer Nuclear Station are silver-indium-cadmium. There are no reactions in these absorber materials which would produce tritium, thus eliminating any contribution from this source. Activation products from control rods are discussed in Section 11.1.1.1. 11.1.4.1.3 Boric Acid Source A direct contribution to the reactor coolant tritium concentration is made by neutron reaction with the boron in solution. The concentration of boric acid varies with core life and load follow so that this is a steadily decreasing source during core life. The principal boron reactions are the B10 (n, 2) H3 and B10 (n, ) Li7 (n, n) H3 reactions. The Li7 reaction is controlled by limiting the overall lithium concentration to approximately 2 ppm during operation. Li6 is essentially excluded from the system by utilizing 99.9% Li7. 11.1.4.1.4 Burnable Shim Rod Source These rods are in the core only during the first operating cycle and their potential tritium contribution occurs only during this period. 11.1.4.1.5 Lithium and Deuterium Lithium and deuterium reactions contribute only minor quantities to the tritium inventory as shown in Table 11.1-7. These sources are due to the activation of the lithium and deuterium in the Reactor Coolant System as they pass through the reactor. 11.1.4.2 Design Bases The design intent is to reduce the tritium sources in the Reactor Coolant System to a practical minimum to permit longer retention of the reactor coolant within the plant without compromising operator exposures. Reduction of source terms is provided by utilizing silver-indium-cadmium control rods and the determination that the quantity of tritium released from the fuel rods with Zircaloy cladding is less than originally expected. 99-01

11.1-7 Reformatted Per Amendment 99-01 11.1.4.3 Design Evaluation Table 11.1-7 lists the present expected release of tritium to the reactor coolant. It will be noted that there are two principle contributors to the tritium production: ternary fission source and the dissolved boron in the reactor coolant. For a leakage from the Reactor Coolant System into the Reactor Building atmosphere of 50 pounds per day with an assumed tritium concentration of 3.5 µCi/gm, the tritium concentration in the Reactor Building atmosphere would be low enough to permit access with no Reactor Building purge and without protective equipment by plant maintenance personnel for an average of 2 hours per week. During refueling operations, a refueling water concentration activity of 2.5 µCi/gm is expected to result in Reactor Building air concentrations at or below the 10 CFR 20 occupational Maximum Permissible Concentration (MPC) value. This concentration would permit 40 hours per week access to the Reactor Building. Although the actual relationship between reactor coolant activities and Reactor Building air concentrations will be determined by the particular operating conditions inside the Reactor Building (temperature, relative humidity, ventilation purge rate, etc.), field measurements indicate that the design objective of 3.5 µCi/gm in the reactor coolant and 2.5 µCi/gm in the refueling water are reasonable values. 11.

1.5 REFERENCES

1. "Source Term Data for Westinghouse Pressurized Water Reactors," WCAP-8253, Revision 1, July, 1975. 2. ANSI/ANS-18.1-1984, "Radioactive Source Terms for Normal Operation of Light Water Reactors". 3. "Radiation Analysis Manual," Virgil C. Summer (CGE), CGE / 3-1, Revision 0, 12/92. 99-01 99-01 99-01

11.1-8 Reformatted Per Amendment 99-01 TABLE 11.1-1 PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION AND CORROSION PRODUCT ACTIVITIES 1. Ultimate core thermal power, MWt 2958 2. Clad defects, as a percent of rated core thermal power being generated by rods with clad defects 1.0 3. Reactor coolant liquid volume, ft3 8830 4. Reactor coolant full power average temperature, F 592.8 5. Reactor coolant density at system operating temperature and pressure, g/cc 0.7 6. Purification flowrate, gpm 60 7. Effective cation demineralizer flow, gpm 6.0 8. Volume control tank volumes

a. Vapor, ft3
b. Liquid, ft3 150 150 9.

Fission product escape rate coefficients:

a. Noble gas isotopes, sec-1 6.5 x 10-8
b. Br, Rb, I, and Cs isotopes, sec-1 1.3 x 10-8
c. Te isotopes, sec-1 1.0 x 10-9
d. Mo, Tc, and Ag isotopes, sec-1 2.0 x 10-9
e. Sr and Ba isotopes, sec-1 1.0 x 10-11
f. Y, Zr, Nb, Ru, Rh, La, Ce, and Pr isotopes, sec-1 1.6 x 10-12 10.

Mixed bed demineralizers decontamination factors:

a. Br, I, Sr, and Ba 10.0
b. Noble gases and all other isotopes 1.0 11.

Cation bed demineralizer decontamination factors

a. Kr and Xe isotopes 1
b. Sr and Ba isotopes 1
c. Rb-86, Cs-134, and Cs-137 10
d. Rb-88, Rb-89, Cs-136, and Cs-138 1
e. Other isotopes 1

99-01

11.1-9 Reformatted Per Amendment 99-01 TABLE 11.1-1 (Continued) PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION AND CORROSION PRODUCT ACTIVITIES 12. Volume control tank noble gas stripping fractions: Isotope Stripping Fraction (1) Kr-85 6.0 x 10-5 Kr-85m 5.6 x 10-1 Kr-87 8.2 x 10-1 Kr-88 6.7 x 10-1 Xe-131m 1.3 x 10-2 Xe-133 3.0 x 10-2 Xe-133m 6.8 x 10-2 Xe-135 3.0 x 10-1 Xe-135m 9.4 x 10-1 Xe-138 9.4 x 10-1 13. Initial boron concentrations High (2) Low (3) 2100 1200 14. Pressurizer volumes

a. Vapor, ft3
b. Liquid, ft3 577 865 15.

Spray line flow, gpm 2.0 16. Pressurizer stripping fractions

a. Noble gases (except Kr-85)
b. Kr-85
c. All other elements 1.0 0.9 0

17. Fuel cycle times (effective full power days) Equilibrium cycle 480 18. Number of reactor coolant loops 3 (1) Assuming no volume control tank purge. (2) High value is assumed where high boron concentration is conservative. (i.e., tritium production). (3) Low value is assumed where low feed and bleed is conservative. (e.g., coolant fission product activities). 99-01

11.1-10 Reformatted Per Amendment 99-01 TABLE 11.1-1 (Continued) PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION AND CORROSION PRODUCT ACTIVITIES 19. Corrosion product parameters Core wetted areas, effective (in2)

a. Zirlo 7.6 x 106
b. Stainless steel 4.9 x 105
c. Inconel 7.7 x 105 Out of core wetted area, Inconel (in2) 3.2 x 107 Coolant velocity (ft/sec)
a. Core
b. Steam generator 14.0 18.0 Nominal base metal release rates (mg/dm2-mo)
a. Zirlo 0.0
b. Stainless steel 0.5
c. Inconel 1.0 99-01

11.1-11 Reformatted Per Amendment 99-01 TABLE 11.1-2 REACTOR COOLANT EQUILIBRIUM FISSION AND CORROSION PRODUCT ACTIVITIES (Based on parameters given in Table 11.1-1) Isotope Activity Ci/gm Fission Products Br-84 4.2 x 10-2 Rb-88 3.8 Sr-89 4.0 x 10-3 Sr-90 2.0 x 10-4 Sr-91 5.3 x 10-3 Sr-92 1.2 x 10-3 Y-90 5.7 x 10-5 Y-91 5.4 x 10-4 Y-92 1.1 x 10-3 Zr-95 6.7 x 10-4 Nb-95 6.7 x 10-4 Mo-99 7.9 x 10-1 I-131 3.0 I-132 3.1 I-133 4.6 I-134 6.0 x 10-1 I-135 2.4 Te-132 2.9 x 10-1 Te-134 2.8 x 10-2 Cs-134 4.4 Cs-136 4.5 Cs-137 2.1 Cs-138 9.7 x 10-1 Ba-140 4.4 x 10-3 La-140 1.4 x 10-3 Ce-144 4.7 x 10-4 Pr-144 4.7 x 10-4 Kr-85 7.6 Kr-85m 1.8 Kr-87 1.1 Kr-88 3.2 Xe-131m 2.3 Xe-133 2.9 x 102 Xe-133m 1.9 x 101 Xe-135 8.6 Xe-135m 5.2 x 10-1 Xe-138 6.4 x 10-1 Mn-54 4.1 x 10-4 Mn-56 2.2 x 10-2 Co-58 1.4 x 10-2 Co-60 1.3 x 10-3 Fe-59 5.2 x 10-4 Cr-51 5.5 x 10-3 99-01

11.1-12 Reformatted Per Amendment 99-01 TABLE 11.1-3 VOLUME CONTROL TANK EQUILIBRIUM ACTIVITIES (Based on parameters given in Table 11.1-1) Isotope Vapor Activity (Ci/cc) Kr-83m 3.3 x 100 Kr-85 1.6 x 102 Kr-85m 1.9 x 101 Kr-87 4.9 x 100 Kr-88 2.5 x 101 Xe-131m 3.2 x 101 Xe-133 4.2 x 103 Xe-133m 2.7 x 102 Xe-135 1.1 x 102 Xe-135m 9.4 x 100 Xe-138 6.5 x 10-1 Isotope Liquid Activity (Ci/gm) Kr-88 8.9 x 10-1 Rb-88 3.8 x 100 Xe-133 1.7 x 102 Xe-135 4.3 x 100 Cs-134 4.4 x 100 Cs-136 4.5 x 100 Cs-137 2.1 x 100 Cs-138 9.7 x 10-1 I-131 3.0 x 100 I-132 3.1 x 100 I-133 4.6 x 100 I-134 6.0 x 10-1 I-135 2.4 x 100 99-01

11.1-13 Reformatted Per Amendment 99-01 TABLE 11.1-4 PRESSURIZER ACTIVITIES (Based on parameters given in Table 11.1-1) Isotope Vapor Activity (Ci/cc) N-16 1.5 x 10-3 Kr-83m 1.8 x 10-2 Kr-85 2.3 x 101 Kr-85m 2.1 x 10-1 Kr-87 2.8 x 10-2 Kr-88 2.2 x 10-1 I-131 3.0 x 10-2 I-132 3.1 x 10-2 I-133 4.6 x 10-2 I-134 6.0 x 10-3 I-135 2.4 x 10-2 Xe-131m 5.1 Xe-133 6.3 x 102 Xe-133m 2.8 x 101 Xe-135 2.1 Xe-135m 7.8 x 10-4 Xe-138 7.9 x 10-4 Isotope Liquid Activity (Ci/gm) N-16 (maximum) 2.9 Rb-88 3.8 I-132 3.1 I-133 4.6 I-135 2.4 Cs-134 4.4 Cs-136 4.5 Cs-138 9.7 x 10-1 Kr-88 3.2 Xe-133 2.9 x 102 Xe-135 8.6 NOTE: See Section 11.1.1.4 for additional information. 99-01

11.1-14 Reformatted Per Amendment 99-01 TABLE 11.1-5 Normal Plant Operation Source Terms Group I - Noble Gases (Based on ANSI / ANS-18.1-1984) No VCT Purge Nuclide Y Parameter Reactor Coolant Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) Kr-85m 5.2 x 10-1 1.6 x 10-1 3.9 x 10-8 Kr-85 5.1 x 10-5 6.1 x 10-1 1.5 x 10-7 Kr-87 7.9 x 10-1 1.7 x 10-1 3.9 x 10-8 Kr-88 6.3 x 10-1 2.9 x 10-1 7.2 x 10-8 Xe-131m 1.1 x 10-2 7.4 x 10-1 1.8 x 10-7 Xe-133m 5.9 x 10-2 6.6 x 10-2 1.7 x 10-8 Xe-133 2.5 x 10-2 2.5 x 100 6.1 x 10-7 Xe-135m 9.3 x 10-1 1.6 x 10-1 3.8 x 10-8 Xe-135 2.7 x 10-1 8.3 x 10-1 2.1 x 10-7 Xe-137 9.8 x 10-1 4.2 x 10-2 1.0 x 10-8 Xe-138 9.3 x 10-1 1.5 x 10-1 3.5 x 10-8 VCT Purge of 2.36 x 102 cm3/sec Nuclide Y Parameter Reactor Coolant Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) Kr-85m 7.0 x 10-1 1.5 x 10-1 3.7 x 10-8 Kr-85 5.7 x 10-1 1.1 x 10-2 2.7 x 10-9 Kr-87 8.3 x 10-1 1.7 x 10-1 3.9 x 10-8 Kr-88 7.5 x 10-1 2.8 x 10-1 7.0 x 10-8 Xe-131m 4.7 x 10-1 7.6 x 10-2 1.8 x 10-8 Xe-133m 4.9 x 10-1 2.4 x 10-2 6.0 x 10-9 Xe-133 4.8 x 10-1 4.8 x 10-1 1.2 x 10-7 Xe-135m 9.3 x 10-1 1.6 x 10-1 3.8 x 10-8 Xe-135 5.5 x 10-1 6.8 x 10-1 1.7 x 10-7 Xe-137 9.8 x 10-1 4.2 x 10-2 1.0 x 10-8 Xe-138 9.4 x 10-1 1.5 x 10-1 3.5 x 10-8 99-01

11.1-15 Reformatted Per Amendment 99-01 TABLE 11.1-5 (Continued) Normal Plant Operation Source Terms Groups II, III, IV, and V (Based on ANSI / ANS-18.1-1984) Group II - Halogens Nuclide Reactor Coolant Activity (Ci/gram) Steam Generator Liquid Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) Br-84 2.0 x 10-2 1.2 x 10-7 1.2 x 10-9 I-131 5.0 x 10-2 3.6 x 10-6 3.6 x 10-8 I-132 2.6 x 10-1 5.4 x 10-6 5.4 x 10-8 I-133 1.6 x 10-1 9.3 x 10-6 9.3 x 10-8 I-134 4.2 x 10-1 4.0 x 10-6 4.0 x 10-8 I-135 3.1 x 10-1 1.2 x 10-5 1.2 x 10-7 Group III - Rubidium, Cesium Nuclide Reactor Coolant Activity (Ci/gram) Steam Generator Liquid Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) Rb-88 2.3 x 10-1 8.6 x 10-7 4.2 x 10-9 Cs-134 7.5 x 10-3 5.5 x 10-7 2.9 x 10-9 Cs-136 9.3 x 10-4 6.7 x 10-8 3.4 x 10-10 Cs-137 9.9 x 10-3 7.4 x 10-7 3.7 x 10-9 Group IV - N-16 Nuclide Reactor Coolant Activity (Ci/gram) Steam Generator Liquid Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) N-16 4.0 x 101 1.3 x 10-6 1.3 x 10-7 Group V - Tritium Nuclide Reactor Coolant Activity (Ci/gram) Steam Generator Liquid Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) H-3 1.0 x 100 1.0 x 10-3 1.0 x 10-3 99-01

11.1-16 Reformatted Per Amendment 99-01 TABLE 11.1-5 (Continued) Normal Plant Operation Source Terms Group VI - Miscellaneous Nuclides (Based on ANSI / ANS-18.1-1984) Nuclide Reactor Coolant Activity (Ci/gram) Steam Generator Liquid Activity (Ci/gram) Steam Generator Steam Activity (Ci/gram) Na-24 5.3 x 10-2 2.8 x 10-6 1.4 x 10-8 Cr-51 3.3 x 10-3 2.5 x 10-7 1.2 x 10-9 Mn-54 1.7 x 10-3 1.2 x 10-7 6.2 x 10-10 Fe-55 1.3 x 10-3 9.3 x 10-8 4.7 x 10-10 Fe-59 3.2 x 10-4 2.3 x 10-8 1.2 x 10-10 Co-58 4.9 x 10-3 3.6 x 10-7 1.8 x 10-9 Co-60 5.6 x 10-4 4.2 x 10-8 2.1 x 10-10 Zn-65 5.4 x 10-4 4.0 x 10-8 1.9 x 10-10 Sr-89 1.5 x 10-4 1.1 x 10-8 5.5 x 10-11 Sr-90 1.3 x 10-5 9.3 x 10-10 4.7 x 10-12 Sr-91 1.1 x 10-3 5.2 x 10-8 2.6 x 10-10 Y-90 1.5 x 10-6 1.1 x 10-10 5.7 x 10-13 Y-91m 5.6 x 10-4 5.3 x 10-9 2.7 x 10-11 Y-91 5.5 x 10-6 4.0 x 10-10 2.1 x 10-12 Y-93 4.8 x 10-3 2.2 x 10-7 1.1 x 10-9 Zr-95 4.2 x 10-4 3.0 x 10-8 1.5 x 10-10 Nb-95 3.0 x 10-4 2.1 x 10-8 1.1 x 10-10 Mo-99 7.0 x 10-3 4.7 x 10-7 2.3 x 10-9 Tc-99m 5.5 x 10-3 2.0 x 10-7 1.0 x 10-9 Ru-103 8.0 x 10-3 5.9 x 10-7 3.0 x 10-9 Ru-106 9.6 x 10-2 7.0 x 10-6 3.4 x 10-8 Rh-103m 9.2 x 10-3 5.2 x 10-7 2.7 x 10-9 Rh-106 1.1 x 10-1 6.0 x 10-6 2.9 x 10-8 Ag-110m 1.4 x 10-3 1.0 x 10-7 5.1 x 10-10 Te-129m 2.0 x 10-4 1.5 x 10-8 7.4 x 10-11 Te-129 2.9 x 10-2 3.7 x 10-7 1.9 x 10-9 Te-131m 1.7 x 10-3 1.0 x 10-7 5.1 x 10-10 Te-131 9.5 x 10-3 4.8 x 10-8 2.5 x 10-10 Te-132 1.8 x 10-3 1.2 x 10-7 6.2 x 10-10 Ba-137m 9.4 x 10-3 7.0 x 10-7 3.5 x 10-9 Ba-140 1.4 x 10-2 9.8 x 10-7 4.9 x 10-9 La-140 2.7 x 10-2 1.8 x 10-6 8.7 x 10-9 Ce-141 1.6 x 10-4 1.2 x 10-8 5.9 x 10-11 Ce-143 3.1 x 10-3 1.9 x 10-7 9.6 x 10-10 Ce-144 4.3 x 10-3 3.0 x 10-7 1.6 x 10-9 Pr-143 3.7 x 10-3 2.3 x 10-7 1.2 x 10-9 Pr-144 4.9 x 10-3 2.6 x 10-7 1.3 x 10-9 W-187 2.8 x 10-3 1.6 x 10-7 8.3 x 10-10 Np-239 2.4 x 10-3 1.6 x 10-7 7.9 x 10-10 99-01

11.1-17 Reformatted Per Amendment 99-01 TABLE 11.1-6 PARAMETERS USED TO DESCRIBE THE REACTOR SYSTEM-REALISTIC MODEL Parameter Symbol Units Thermal power P MWt 2958 Steam flowrate FS lbs/hr 1.3 x 107 Weight of water in the Reactor Coolant System WP lbs 3.9 x 105 Weight of water in all steam generators WS lbs 3.4 x 105 Reactor coolant letdown flow (purification) FD lbs/hr 3.0 x 104 Reactor coolant letdown flow (yearly average for boron control) FB lbs/hr 300 Steam generator blowdown flow (total) FBD lbs/hr 4.2 x 104 Fraction of radioactivity in blowdown stream which is not returned to the secondary coolant system NBD 1.0 Flow through the purification system cation demineralizer FA lbs/hr 3.0 x 103 Ratio of condensate demineralizer flowrate to the total steam flowrate NC 0.0 Ratio of the total amount of noble gases routed to gaseous radwaste from the purification system to the total amount of noble gases routed to the primary coolant system from the purification system (not including the Boron Recycle System) Y See Table 11.1-5 Primary to Secondary Leak Rate lbs/day 100 99-01

11.1-18 Reformatted Per Amendment 99-01 TABLE 11.1-7 TRITIUM PRODUCTION Tritium Source Expected Release to Reactor Coolant Curies/Cycle Produced in Core Ternary Fissions 1420 IFBAs 237 Produced in Coolant Soluble Boron 805 Soluble Lithium 142 Deuterium 3.24 Total 2607 Note: Power level = 2958 MWt IFBA B-10 mass = 1730 gm Initial cycle reactor coolant boron concentration = 1200 ppm Equilibrium cycle reactor coolant boron concentration = 2100 ppm Lithium concentration (99.9 atom percent Li7) = 2.2 ppm 99-01

11.2-1 Reformatted November 2011 11.2 LIQUID WASTE SYSTEMS This section describes the design and operating features of the Liquid Waste Processing System (LWPS). In addition the total plant liquid releases from all sources are estimated and summarized in Section 11.2.6. Section 11.2.2 describes an upper limit system capacity which can be maintained for an indefinite period of time. 11.2.1 DESIGN OBJECTIVES The LWPS is designed to receive, control, segregate, process, reuse, and discharge liquid wastes. The system design considers potential personnel exposure and assures that quantities of radioactive releases to the environment are in accordance with 10 CFR 50, Appendix I. Under normal plant operation, the activity from radionuclides leaving the penstock of the Fairfield Pumped Storage Facility, due to releases from the LWPS, is a small fraction of the effluent concentration limits (ECLs) as defined in Appendix B of 10 CFR 20. Section 11.2.6 establishes that the LWPS adequately meets the above listed design objectives. 11.2.2 SYSTEMS DESCRIPTIONS The LWPS primarily collects and processes potentially radioactive wastes for release to the environment. Provisions are made to sample and analyze fluids before they are discharged. Based on the laboratory analysis, these wastes are either released under controlled conditions via the penstocks of the Fairfield Pumped Storage Facility or retained for further processing. Permanent records of liquid releases are provided by analyses of known volumes of waste. Alternatively, the liquid waste may be reused in the plant. The bulk of the radioactive liquids discharged from the Reactor Coolant System are processed by the Reactor Grade Demineralizer System discussed in Section 9.3.6. This limits input to the LWPS and results in processing of relatively small quantities of generally low activity level wastes. The LWPS consists of 5 collection systems which are provided by the waste holdup tank, floor drain tank, the laundry and hot shower tank, the excess liquid waste processing system (the excess waste holdup tank and the decon pit collection tank) and the laboratory drain system. Capability for handling and storage of spent demineralizer resins is also provided. The LWPS does not include provisions for processing secondary system wastes. The Nuclear Blowdown Processing System is discussed in Section 10.4.8. The segregation of primary and secondary side wastes is maintained since ammonia from the secondary side could result in the loss of LWPS demineralizer efficiency, and condenser inleakage could lead to undesirable chemical inclusion in the LWPS. Additionally, the mixing of RN 03-038

11.2-2 Reformatted November 2011 low activity wastes (secondary side) with those of higher activity (primary side) should be avoided since a large volume of contaminated water is produced. The present design, which segregates primary and secondary wastes, minimizes the amount of water which must be processed by discharging low activity wastes directly, where permissible, with no treatment. In the event of equipment faults of moderate frequency (Section 11.2.4.2), the LWPS is capable of processing up to 1 gpm of primary coolant leakage with no change in system operation. As a practical upper limit of system operation, the LWPS can process 25 gpm not including laundry type effluents which are normally discharged without processing. This liquid may be collected in either the floor drain tank or waste holdup tank or in both tanks. Instrumentation and controls necessary for the operation of the LWPS are located on a control board in the Auxiliary Building. Any alarm on this control board is relayed to the main control board in the Control Room. 11.2.2.1 Waste Holdup Tank The waste holdup tank is provided to collect both reactor and non-reactor grade water which enters the LWPS via equipment leaks and drains, valve leakoffs, pump seal leakoffs, tank overflows, reactor building sump flows and other tritiated and aerated water sources. Deaerated tritiated water inside the Reactor Building from sources such as valve leakoffs, which is collected in the reactor coolant drain tank, need not enter the waste holdup tank. These may be routed directly to the recycle holdup tanks for processing. The basic composition of the liquid collected in the waste holdup tank is normally boric acid and water with some radioactivity. Liquid collected in this tank is normally processed through the Waste Water System (Duratek demineralizers) and released to the environment under controlled conditions. Liquid wastes are released from the waste monitor tanks through the penstocks of the Fairfield Pumped Storage Facility. The discharge valve is interlocked with a process radiation monitor and closes automatically when the radioactivity concentration in the liquid discharge exceeds a preset limit. The waste monitor tanks act as a reservoir for storing waste which is to be released from the LWPS to the Fairfield Pumped Storage Facility. Prior to entering these tanks, the liquid may pass through a waste monitor tank demineralizer and a waste monitor tank filter, if required for additional cleanup. RN 03-038

11.2-3 Reformatted November 2011 Normally, the waste monitor tank demineralizer and filter are bypassed. A sample is taken and, after analysis, the results are logged and the liquid is discharged. This radiation monitor, RM-L5, is described in Section 11.4. Liquid waste discharge flow and volume are recorded. 11.2.2.2 Floor Drain Tank The floor drain tank is provided to collect non-reactor grade (non-recyclable) liquid wastes. These include floor drains, equipment drains containing non-reactor grade water, and other non-reactor grade sources. Normally only non-radioactive water is collected in the floor drain tank which can then be sent directly to the waste monitor tank without processing and subsequently discharged. If there is activity in the floor drain tank liquid and is such that the discharge limits cannot be met without cleanup, the liquid may be processed through the Waste Water System (Duratek demineralizers) and released under controlled conditions via the penstocks of the Fairfield Pumped Storage Facility. Non-recyclable reactor coolant leakage normally enters the floor drain tank from system leaks in the Auxiliary Building via the floor drains. This liquid is not reused because it is diluted and contaminated by non-reactor grade water entering the floor drain tank from other sources. Sources of water include fan cooler leaks, secondary side steam and feedwater leaks, component cooling water leaks, and decontamination water. 11.2.2.3 Laundry and Hot Shower Tank Laundry and hot shower drains normally need no treatment for removal of radioactivity. This water is transferred to waste monitor tank number 2 via the laundry and hot shower filter. A sample is taken and, after analysis, the results are logged and the water is discharged if the activity level is below acceptable limits. 11.2.2.4 Excess Liquid Waste Processing System (ELWS) The ELWS consists of 2 storage tanks, the excess waste holdup tank and the decon pit collection tank. The excess waste holdup tank is used to accept excess liquid waste from the floor drain tank, laundry and hot shower tank, and waste holdup tank when these tanks are filled to capacity. The liquid from this tank can be recycled back to these tanks, released directly to the environment via the waste monitor tank or processed through the Duratek demineralizers prior to release from the plant. The decon pit collection tank collects liquid from the Fuel Handling Building sumps, the Radiological Maintenance Building drains, excess waste holdup tank sump, excess waste holdup area sump and decon pit drains. If the activity in this tank liquid is such that the discharge limits cannot be met without cleanup, the liquid is processed through the Duratek demineralizers and released under controlled conditions via the penstocks of the Fairfield Pumped Storage Facility. RN 03-038

11.2-4 Reformatted November 2011 This system also normally receives liquid waste from the Normal and Post Accident Sampling System waste pump. In addition, the Turbine Building Floor Drain System discharge will be directed to the ELWS when excessive radioactive discharge is detected by radiation monitor number RM-L8. 11.2.2.5 Laboratory Drain System The laboratory drain system consists of three sinks in the radiochemical laboratory and two sinks in the sample room. In the radiochemical laboratory spent reactor coolant samples, equipment rinse water and other non-reactor grade fluids are disposed of in the two sinks that drain to the floor drain tank. No liquids or wastes are intentionally disposed of in the sink that drains to the chemical drain tank. In the sample room, excess sample purges of reactor grade water and excess reactor coolant samples are drained from one sink to the waste holdup tank for processing. The other sink is used for draining non-reactor grade fluids to the nuclear blowdown holdup tank. 11.2.2.6 Waste From Spent Resin The spent resin sluice portion of the LWPS consists of a spent resin storage tank, a spent resin sluice pump, and a spent resin sluice filter. The equipment is arranged such that the resin sluice water after entering a demineralizer vessel returns to the spent resin storage tank for reuse. The purpose of this system is to transport spent resin to the spent resin storage tank without generating large volumes of waste liquid. This is accomplished by reusing the sluice water for subsequent resin sluicing operations. 11.2.3 SYSTEM DESIGN 11.2.3.1 Component Design Principal design parameters for the LWPS equipment are given in Table 11.2-2. Parts or components in contact with borated water are fabricated from or clad with austenitic stainless steel except for the Royal Flex hose provided by the vendor as an integral part of the Duratek demineralizers. The Royal Flex hose is a reinforced Thermoplastic Vinyl Nitrile hose. Pumps are provided with vent and drain connections. Component safety classes and the corresponding code and code class are shown in Table 3.2-1. Except for flanged joints and quick disconnect couplings, all-welded construction is used. RN 03-038

11.2-5 Reformatted November 2011 Table 11.2-3 lists the isotopic inventory of each major component, using the following assumptions:

1.

Concentrations based on an equivalent fuel defect level of 1 percent. Values given in the table indicate liquid activity only.

2.

For tanks upstream of the Duratek system (floor drain tank, waste holdup tank) the tanks were assumed 80 percent full of primary coolant.

3.

The tanks downstream of the Duratek system were assumed 80 percent full of processed primary coolant.

4.

The Duratek system components for a typical lineup are shown in Figure 11.2-5.

5.

The laundry and hot shower tank was assumed 80 percent full of laundry waste water. These activities are the recommended source terms for the analysis of consequences of a postulated failure of the LWPS components. 11.2.3.1.1 Pumps 11.2.3.1.1.1 Reactor Coolant Drain Tank Pumps Two (2) pumps are provided. One (1) reactor coolant drain tank pump provides sufficient flow for normal tank operation with 1 pump for standby. 11.2.3.1.1.2 Waste Evaporator Feed Pump One (1) pump of standard design is used. The waste evaporator feed pump is used to transfer liquid from the waste holdup tank. The pump trips off in auto. 11.2.3.1.1.3 Waste Evaporator Condensate Pump The waste evaporator condensate pump is a transfer pump. One (1) pump of standard design is used to transfer the contents of the waste evaporator condensate tank to the waste monitor tank. 11.2.3.1.1.4 Chemical Drain Tank Pump The chemical drain tank pump is of standard design and is used to transfer the liquid from the chemical drain tank. 11.2.3.1.1.5 Spent Resin Sluice Pump This pump is identical to the reactor coolant drain tank pumps. Its delivery flow is based on the required velocity to sluice resin. RN 03-038

11.2-6 Reformatted November 2011 11.2.3.1.1.6 Laundry and Hot Shower Tank Pump The laundry and hot shower tank pump is of standard design and is used to transfer the water to the waste monitor tank. 11.2.3.1.1.7 Floor Drain Tank Pump The floor drain tank pump is of standard design and is used to normally transfer water to the waste monitor tank. 11.2.3.1.1.8 Waste Monitor Tank Pumps A waste monitor tank pump is of standard design and is used for each tank to discharge water from the LWPS or for recycle if further processing is required. The pump may also be used for circulating the water in the waste monitor tank in order to obtain uniform tank contents and hence a representative sample before discharge. The pump can be throttled to achieve the desired discharge rate. 11.2.3.1.1.9 Excess Liquid Waste Pumps The two (2) excess liquid waste pumps are used to transport the waste fluid from the excess waste holdup tanks through the processing portion of the ELWS and to return waste to the LWPS. These pumps also circulate the waste in the tank to obtain proper mixing prior to sampling tank contents. 11.2.3.1.2 Reactor Coolant Drain Tank Heat Exchanger The reactor coolant drain tank heat exchanger is a U-tube type with 1 shell pass and 4 tube passes. Although the heat exchanger is normally used in conjunction with the reactor coolant drain tank, it can also cool the pressurizer relief tank contents from 200 to 120°F in less than 8 hours. 11.2.3.1.3 Tanks 11.2.3.1.3.1 Reactor Coolant Drain Tank One (1) tank is provided. The purpose of the reactor coolant drain tank is to collect leakoff type drains inside the Reactor Building at a central collection point for further disposition through a single penetration via the reactor coolant drain tank pumps. The tank provides surge and net positive suction head requirements to the pumps. Only water which can be directed to the recycle holdup tanks enters the reactor coolant drain tank. The water must be compatible with reactor coolant and it must not contain dissolved air. RN 03-038

11.2-7 Reformatted November 2011 Sources of water entering the reactor coolant drain tank include the reactor vessel flange leakoff, valve leakoffs, number 2 and 3 seal leakoffs from the reactor coolant pumps and the excess letdown heat exchanger flow. No continuous leakage is expected from the reactor vessel flange during operation. A constant level is maintained in the tank to minimize the amount of gas sent to the gaseous waste processing system and also to minimize the amount of hydrogen required. The level is maintained by running 1 pump continuously and using a proportional control valve in the discharge line. This valve operates on a signal from a level controller to limit the flow out of the system. The remainder of the flow is recirculated to the tank. 11.2.3.1.3.2 Waste Holdup Tank One (1) atmospheric pressure tank is provided outside the Reactor Building to collect equipment drains, valve and pump seal leakoffs, recycle holdup tank overflows, reactor building sump fluid and other water from tritiated aerated sources. 11.2.3.1.3.3 Waste Evaporator Condensate Tank One (1) tank with a diaphragm to exclude air is provided to collect low level liquid waste prior to sending it to the waste monitor tank. 11.2.3.1.3.4 Chemical Drain Tank One (1) tank is available to collect chemically contaminated tritiated water from the laboratories, however, this tank is normally not used. 11.2.3.1.3.5 Spent Resin Storage Tank The purpose of the spent resin storage tank is to provide a collection point for spent resin to allow for decay of short lived radionuclides before disposal. The tank serves also as a head tank for the spent resin sluice pump. The tank is designed so that sufficient pressure can be applied in the gas space of the tank to push resin out and to the solid waste disposal unit, which is at a higher elevation than the spent resin storage tank. The spent resin storage tank is shielded to limit the dose to personnel. 11.2.3.1.3.6 Laundry and Hot Shower Tank One (1) atmospheric pressure tank is used to collect laundry and hot shower drains. RN 03-038

11.2-8 Reformatted November 2011 11.2.3.1.3.7 Floor Drain Tank One (1) atmospheric pressure tank is used to collect floor drains from the reactor plant. 11.2.3.1.3.8 Waste Monitor Tanks The 2 atmospheric waste monitor tanks are provided for monitoring liquid discharges from the plant site. Each tank is sized to hold a volume large enough such that sampling requirements are minimized, thus minimizing further laboratory effluent. 11.2.3.1.3.9 Excess Waste Holdup Tank The excess waste holdup tank normally receives liquid waste from the floor drain tank, laundry and hot shower tank, and waste holdup tank when these tanks are filled to capacity. This system also normally receives liquid waste from the Normal and Post Accident Sampling System waste pump. In addition, the Turbine Building Floor Drain System discharge will be directed to this tank when excessive radioactive discharge is detected by radiation monitor number RM-L8. 11.2.3.1.3.10 Decontamination Pit Collection Tank The decontamination pit collection tank receives drainage from the Fuel Handling Building sumps, the Radiological Maintenance Building drains, excess waste holdup tank sump, excess waste holdup area sump and decon pit drains. 11.2.3.1.4 Demineralizers As part of a continuous PWR operating plant following, Westinghouse has obtained operational data on demineralizer decontamination factors for selected isotopes. The measured range of decontamination factors for these isotopes is given in Table 11.2-4. These values were observed across mixed bed demineralizers containing cation resin in the lithium-7 form and anion resin in the borated form. The minimum values in Table 11.2-4 were generally observed just prior to resin flushing and recharging, while during the operating life of the demineralizer, decontamination factors were consistently closer to the maximum values. Although specific operating decontamination factors have not as yet been measured for other isotopes, their behavior in a similar purification media may be inferred from this data. One would anticipate, for example, bromine to have a decontamination factor similar to that given above for the iodine and fluorine. RN 03-038

11.2-9 Reformatted November 2011 11.2.3.1.4.1 Waste Water System (Duratek Demineralizers) The Waste Water System (Duratek demineralizers) is provided to process radioactive waste prior to release to the environment. The Duratek demineralizers are located on the 447 elevation of the Auxiliary Building. Prior to entering the mixed bed and cation Duratek demineralizers, the liquid waste stream is normally processed through a charcoal vessel. The charcoal media acts as a prefilter providing mechanical filtration capability prior to the Duratek demineralizers. The charcoal filtration provides for cleanup of oil and grease (organics), cobalt 58 and 60, initial iodines, cesium and other suspended solids. The liquid waste stream enters the Duratek demineralizers through a booster pump, a mechanical filter, and an Influent Control Leg which monitors waste stream parameters and provides attachments for service air and water. From the Influent Control Leg, the waste may enter one of the five pressure vessels that contain media for cleanup of the liquid waste. The media normally include:

  • Cation resin to remove metals and transition metals.
  • Mixed bed resin that reduces undesired positive and negative ions.
  • Charcoal media for cleanup of oil and grease (organics), cobalt 58 and 60, initial iodines, and cesium and for service as a prefilter to protect the other demineralizer beds.

A typical lineup for the five Duratek demineralizers in series would be charcoal, cation, charcoal, cation, mixed bed with cation. The actual required vessel lineup during normal plant operation is determined prior to processing by sampling the tank to be processed and determining the parameter(s) that are out of specification. Selection of the appropriate vessels is made by a quick-disconnect logic arrangement. After processing, decontaminated effluent passes through an Effluent Control Leg which records waste volume, and may flow through a combined resin polisher. Polyelectrolytes may be added upstream of the second charcoal vessel to help remove colloidal cobalt. This is mainly used during an outage post reactor coolant system cleanup. 11.2.3.1.4.2 Waste Evaporator Condensate Demineralizer One (1) demineralizer vessel is provided which is used during primary demineralizer resin (CVCS) transfer and provides a means to tie the resin supply and return headers together during resin transfer to prevent resin excursions throughout the solid waste system. The demineralizer vessel is empty (no contained resin beds) and is not used for waste decontamination. RN 03-038

11.2-10 Reformatted November 2011 11.2.3.1.4.3 Waste Monitor Tank Demineralizer One (1) demineralizer is provided upstream of the waste monitor tanks. 11.2.3.1.4.4 Excess Liquid Waste Demineralizers Two (2) redundant demineralizers are provided in the ELWS process train. Liquid wastes may be passed through 1 demineralizer prior to being discharged to a waste monitor tank. 11.2.3.1.5 Filters and Strainers 11.2.3.1.5.1 Filters The filters provided are of a wound cartridge or spun cartridge type construction which relies on a tortuous path to filter particles with an absolute rating. The methods employed to change filters and screens are dependent on activity levels. Filters are valved out of service with a pressure indicator between the isolation valves to assure the valves are not leaking and the filter is not at system pressure. The filter is drained to the appropriate tank and vented locally. If the radiation level of the filter is low enough, it is changed manually. If activity levels do not permit manual change, the spent cartridge is removed remotely with temporary shielding to protect personnel. The spent cartridge is placed in a shielded pig for removal to the solid waste disposal area. A new cartridge is installed, the housing is reassembled, vent and drain valves are closed, and the filter is valved into service. Filters are normally changed because of high P rather than high radiation levels. The excess liquid waste filters are back flushable and have cylindrical woven screen cartridges. One (1) of the 2 excess liquid waste filters removes suspended solids from the waste water upstream of the demineralizers. The second filter traps resin fines downstream of the demineralizer. 11.2.3.1.5.2 Strainers The strainers provided are basket type which are of a mesh or screen construction. The nominal rating of the strainers is given in Table 11.2-2. The basket type laundry and hot shower strainer is not replaced after use, but is cleaned and put back in service. Because this screen traps only large particles, it contains only negligible activity and provides no hazard to personnel. It is cleaned regularly. The decontamination pit collection tank strainer is a duplex basket strainer that removes large suspended particles from the wastes entering the decontamination pit collection tank. This strainer is also cleaned and put back into service. RN 03-038

11.2-11 Reformatted November 2011 11.2.3.2 Instrumentation Design The system instrumentation is described in Table 11.2-5. The instrumentation readout is located mainly on the Waste Processing System (WPS) panel in the Auxiliary Building and excess liquid waste panel (ELWP) in the Fuel Handling Building. Some instruments are read where the equipment is located. Alarms are shown on the WPS panel or ELWP and further relayed to 1 common WPS annunciator on the main control board in the Control Room. Pumps are protected against loss of suction pressure by a control setpoint on the level instrumentation for the respective vessels feeding the pumps. The reactor coolant drain tank pumps and the spent resin sluice pump are in addition interlocked with flowrate instrumentation and stop operating when the delivery flows reach minimum setpoints. Pressure indicators upstream and downstream of filters, strainers, and demineralizers provide local indications of pressure drops across each component. All releases to the environment are monitored for radioactivity. 11.2.3.3 Tank Overflow Protection All tanks in the Chemical and Volume Control System (CVCS), Boron Recycle System (BRS), Nuclear Blowdown Processing System (NBS), and WPS that could potentially contain radioactive liquids are designed to provide adequate warning of potential overflow conditions. A summary of the overflow protection features is given in Table 11.2-6. These tanks are provided with level indication instrumentation which has an alarm function on high liquid level in the tank. Alarm annunciation is provided separately on the local system control panel and further relayed to a common annunciator on the main control board in the Control Room for each system. A description of the level instrumentation provided for these systems is given in Sections 9.3.4.5, 9.3.6.5, 10.4.8.2, and 11.2.3.2 for the CVCS, BRS, NBS, and LWPS, respectively. In addition to tank level monitoring and warning of potential overflow conditions, provisions are made in the systems design to collect and process overflows from tanks containing potentially radioactive liquids. The collection and processing provisions are delineated in Table 11.2-6. Table 11.2-6a presents a comparison of tanks with Branch Technical Position ETSB 11-1 (Rev. 1), Paragraph B.1.b, Items (1) through (4). 11.2.4 OPERATING PROCEDURES The LWPS is manually operated except for some functions of the reactor coolant drain tank circuit. The system includes adequate control equipment to protect the system components and adequate instrumentation and alarm functions to provide operator information to ensure proper system operation. RN 03-038

11.2-12 Reformatted November 2011 11.2.4.1 Normal Operation Operation of the LWPS is essentially the same during all phases of normal and defined off-normal reactor plant operation; the only differences are in the loads on the system. The following sections discuss the operation of the system in performing its various functions. In this discussion, the term "normal operation" should be taken to mean all phases of operation except operation under emergency or accident conditions. The LWPS is not regarded as an Engineered Safety Feature System. 11.2.4.1.1 Waste Holdup Tank Water is accumulated in the waste holdup tank until sufficient quantity exists to warrant processing through the Waste Water System (Duratek demineralizers). When this water has been sufficiently processed, it is discharged into the penstocks of the Fairfield Pumped Storage Facility at a rate so as not to exceed a small fraction of the 10 CFR 20 limits. Water leaving this system is discharged to the penstocks of the Fairfield Pumped Storage Facility and is monitored for radiation. This radiation monitor, RM-L5, is described in Section 11.4. Should the radiation monitor close the discharge valve, it must be reset/bypassed before the valve can be reopened. The monitor element can be cleared by flushing it with demineralized water from the temporary connection back to the waste monitor tank. During refueling, the load on the waste portion of the LWPS is increased but there is no change in operation. During normal operation, the reactor coolant drain tank level regulation and pressure control are automatic and require no operator action. Operation of the recycle portion of the LWPS during refueling is the same as for power operation, although the load on the system may be increased when refueling is complete. The water remaining in the fuel transfer canal following normal drain down is pumped to the suction of the refueling water purification pump by the reactor coolant drain tank pumps. When the pumps lose suction, the remainder is drained to the RB sump and pumped to the waste holdup tank for processing. 11.2.4.1.2 Floor Drain Tank The water in the floor drain tank is sampled to determine the degree of processing required. Normally the water collected in the floor drain tank is not radioactive and can be sent directly to the waste monitor tanks. If required, it can be processed through the Waste Water System. When this water has been sufficiently processed, it is discharged into the penstocks of the Fairfield Pumped Storage Facility at a rate so as not to exceed a small fraction of the 10 CFR 20 limits. Water leaving this system is discharged to the penstocks of the Fairfield Pumped Storage Facility and is monitored for radiation. This radiation monitor, RM-L5, is described in Section 11.4. Should the radiation monitor close the discharge valve, it must be reset/bypassed before the valve can be reopened. The monitor element can be cleared by flushing it with demineralized water from the RN 03-038

11.2-13 Reformatted November 2011 temporary connection back to the waste monitor tank. During refueling the load on the waste portion of the LWPS is increased but there is no change in operation. 11.2.4.1.3 Laundry and Hot Shower Tank Laundry and hot shower water enters the laundry and hot shower tank for holdup. There it is sampled, filtered, and transferred to the waste monitor tank for discharge. 11.2.4.1.4 Excess Liquid Waste Processing System (ELWS) The excess waste holdup tank is normally only used to collect excess flow from the floor drain tank, laundry and hot shower tank, and waste holdup tank when these tanks are filled to capacity. The decontamination pit collection tank receives drainage from the Fuel Handling Building sumps, the Radiological Maintenance Building drains, excess waste holdup tank sump, excess waste holdup area sump and decon pit drains. The contents of these tanks are sampled to determine the degree of processing required. The liquid from this tank can be released directly to the environment via the waste monitor tank or if required, processed through the Duratek demineralizers prior to release from the plant.. When this water has been sufficiently processed, it is discharged into the penstocks of the Fairfield Pumped Storage Facility at a rate so as not to exceed a small fraction of the 10 CFR 20 limits. Water leaving this system is discharged to the penstocks of the Fairfield Pumped Storage Facility and is monitored for radiation. This radiation monitor, RM-L5, is described in Section 11.4. Should the radiation monitor close the discharge valve, it must be reset/bypassed before the valve can be reopened. The monitor element can be cleared by flushing it with demineralized water from the temporary connection back to the waste monitor tank. During refueling the load on the waste portion of the LWPS is increased but there is no change in operation. 11.2.4.1.5 Laboratory Drain Portion The laboratory drain system consists of three sinks in the radiochemical laboratory and two sinks in the sample room. In the radiochemical laboratory spent reactor coolant samples, equipment rinse water and other non-reactor grade fluids are disposed of in the two sinks that drain to the floor drain tank. No liquids or wastes are intentionally disposed of in the sink that drains to the chemical drain tank. In the sample room, excess sample purges of reactor grade water and excess reactor coolant samples are drained from one sink to the waste holdup tank for processing. The other sink is used for draining non-reactor grade fluids to the nuclear blowdown holdup tank. RN 03-038

11.2-14 Reformatted November 2011 11.2.4.1.6 Spent Resin Handling Portion This portion of the system sluices resin from the demineralizers, transports resin from the spent resin storage tank to the solid waste disposal unit. 11.2.4.1.6.1 Resin Sluicing Before resin sluicing begins, the demineralizer is valved out of service and the flowpath is aligned from the spent resin sluice pump through the process line of the demineralizer, through the screen at the top of the demineralizer, and back to the spent resin storage tank. The spent resin sluice pump provides flush water for loosening the bed for sluicing. After about 15 minutes of back flushing, the valves in the back flush circuit are closed and the sluice line is opened. The resin then flows to the spent resin storage tank. After the spent resin sluice pump is shut off the fresh resin is added via the resin fill line and the valve is then closed. The flowpath is now aligned the same as for resin flushing, e.g., through the process line, through the screen at the top of the demineralizer, and back to the spent resin storage tank. The pump is then started to remove resin fines, should any remain. The valves are then realigned for normal process operation. Resins are never sluiced through the spent resin sluice pump 11.2.4.1.6.2 Resin Drumming No resin drumming is performed on site. 11.2.4.1.7 Analysis of System Operation In order to evaluate the expected operation of the LWPS an analysis has been performed presenting average operating parameters and expected isotopic concentrations throughout the system. The analysis is based on normal operation, including anticipated operational occurrences, of the plant and the LWPS and a realistic estimation of the potential input sources based on plant operating experience. Hence, the results are representative of the anticipated operation of the LWPS of the Virgil C. Summer Nuclear Station. The input sources assumed in the study are summarized in Table 11.2-7, and the isotopic concentrations are based on reactor coolant inventories as given in Table 11.1-5. The resulting isotopic concentrations at key locations in the LWPS are given in Table 11.2-1. The study is not indicative of maximum system capacity as the maximum processing capability is considerably greater than the values given in Table 11.2-7. The analysis is not used as a basis for the plant release evaluation, which is based on the GALE Code (see Section 11.2.6). RN 03-038

11.2-15 Reformatted November 2011 11.2.4.2 Faults of Moderate Frequency The system is designed to handle the occurrence of equipment faults of moderate frequency such as:

1.

Malfunction in the LWPS Malfunction in this system could include pump, valve, or demineralizer failure. Because of pump standardization throughout the system, a spare pump can be used to replace most pumps in the system. There is sufficient surge capacity in the system to accommodate waste until the failure can be fixed and normal plant operation resumed.

2.

Excessive Leakage in Reactor Coolant System Equipment The system is designed to handle a 1 gpm reactor coolant leak in addition to the expected leakage during normal operation. Operation of the system is almost the same as for normal operation except the load on the system is increased. A 1 gpm leak into the reactor coolant drain tank is handled automatically. If the 1 gpm leak enters the waste holdup tank, operation is the same as normal except for the increased load on the demineralizers. Abnormal reactor coolant leakage can also be accommodated by the floor drain tank and processed. As discussed in Section 11.2.2 a practical upper limit to the Duratek demineralizer processing capacity is 25 gpm. Depending on the source of the liquid the 25 gpm may be collected in either the floor drain tank or waste holdup tank or both tanks. Normally the potentially radioactive fluids are only collected in the waste holdup tank since the floor drain tank is typically used to collect non-radioactive leakage. If the flow is collected in both tanks, the effluent should be discharged after processing. It is possible to operate in this condition for an indefinite time.

3.

Excessive Leakage in Auxiliary System Equipment Leakage of this type could include water from fan cooler leaks inside the Reactor Building which are collected in the Reactor Building sump and sent to the waste holdup tank. Other sources could be Component Cooling water leaks, Service Water Leaks, and steam and secondary side leaks. This water enters the floor drain tank and is processed and discharged as during normal operation. RN 03-038

11.2-16 Reformatted November 2011 NOTE 11.2.5 Section 11.2.5 is being retained for historical purposes only. 11.2.5 PERFORMANCE TESTS Initial performance tests are performed to verify the operability of the components, instrumentation and control equipment, and applicable alarms and control setpoints. The specific objectives are to demonstrate the following:

1.

Pumps are capable of producing the flowrate and head as required.

2.

Waste filters are capable of passing the required flowrate.

3.

Waste evaporator is operable to specifications.

4.

Instrumentation, controllers, and alarms operate satisfactorily to maintain process parameters, indicate, record, and alarm as required.

5.

All sampling points are available for sampling. During reactor operation the system is used at all times and hence is under continuous surveillance. 11.2.6 ESTIMATED RELEASES Liquid releases from the Virgil C. Summer Nuclear Station are calculated using the PWR-GALE Code[1] as specified by Regulatory Guide 1.112 (see Appendix 3A). The input parameters used to calculate liquid releases are listed in Table 11.2-8 and are discussed in more detail below. Releases calculated using the parameters listed in Table 11.2-8 are presented in Tables 11.2-12 and 11.2-13. A comparison of effluent concentrations with 10 CFR 20, Appendix B, Table II, Column 2 values is presented in Section 11.2.8. 11.2.6.1 Reactor Grade Demineralizers A conservatively estimated 300 gpd of reactor grade wastes are collected by the reactor coolant drain tank (equipment drain wastes). The equipment drain wastes are processed via the reactor grade demineralizers prior to entering a waste monitor tank for monitoring and discharge. The release calculations use an assumed discharge fraction of 1.0. RN 03-038

11.2-17 Reformatted November 2011 Radioactive decay during collection of the equipment drains is conservatively calculated using a collection time of 2.08 days. Radioactive decay during processing is calculated using a process time of 0.11 days. No credit is taken for radioactive decay during discharge since the waste monitor tank may be discharged prior to complete processing. The minimum expected decontamination factors for radionuclide removal with the Duratek system are 1,000 for iodine, 200 for cesium and rubidium and 10,000 for other nuclides. These values are based upon the decontamination factors for two mixed bed and one cation demineralizer in series. In calculating the effluent releases, the above values are conservatively reduced by a factor 10. The isotopic distribution of this release is given in Table 11.2-12. 11.2.6.2 Liquid Waste Processing System 11.2.6.2.1 Waste Holdup Tank Wastes from reactor grade sample drains, valve and pump leakoffs and equipment drains outside the reactor building (clean wastes) are collected in the 10,000 gallon waste holdup tank and processed through the Waste Water System. The processed wastes are pumped to a waste monitor tank for monitoring and discharge or additional processing, if required. Based upon the information in Reference [1] for liquid waste volumes and activities and on expected volumes and activities for liquid wastes not included in Reference [1] the total flow is conservatively estimated to be 200 gpd at primary coolant activity (PCA). The release calculations conservatively use a discharge fraction of 1.0. Radioactive decay during collection in the waste holdup tank is calculated using a collection time of 2.08 days. This value is based upon filling the waste holdup tank to 40 percent of capacity. Radioactive decay during processing is calculated using a process time of 0.11 days. This value is based upon processing 80 percent of the waste holdup tank capacity at the design flow rate of 25 gpm. No credit is taken for radioactive decay during discharge since the waste monitor tank may be discharged before the waste holdup tank is completely processed. The minimum expected decontamination factors for radionuclide removal with the Waste Water System are 1,000 for iodine, 200 for cesium and rubidium and 10,000 for other nuclides. These values are based upon the decontamination factors for two mixed bed and one cation demineralizer in series. In calculating the effluent releases, the above values are conservatively reduced by a factor of 10. The isotopic distribution of this release is given in Table 11.2-12. RN 03-038

11.2-18 Reformatted November 2011 11.2.6.2.2 Floor Drain Tank Wastes from floor drains, laboratory drains and miscellaneous sources are collected in a 10,000 gallon floor drain tank, sampled to determine the degree of processing required and processed as necessary through the Waste Water System. The processed wastes are then pumped to a waste monitor tank for monitoring and discharge. Based upon the information in Reference [1] for liquid waste volumes and activities, the waste flow is estimated to be 1340 gpd at 0.05 PCA. Since all of the processed wastes are normally discharged, a discharge fraction of 1.0 is used in the release calculations. Radioactive decay during collection in the floor drain tank is calculated using a collection time of 2.97 days. This value is based upon filling the floor drain tank to 40 percent of capacity. Radioactive decay during processing is calculated using process time of 0.11 day. This value is based upon processing 80 percent of the floor drain tank capacity at the design flow rate of 25 gpm. No credit is taken for radioactive decay during discharge since the waste monitor tank may be discharged before the floor drain tank is completely processed. The minimum expected decontamination factors for radionuclide removal with the Waste Water System are 1,000 for iodine, 200 for cesium and rubidium, and 10,000 for other nuclides. These values are based upon the decontamination factors given in Reference [1] for two mixed bed and one cation demineralizer in series. In calculating the effluent releases the above values are conservatively reduced by a factor 10. The isotopic distribution of this release is given in Table 11.2-12. 11.2.6.3 Detergent Wastes The plant typically ships all protective clothing off site for cleaning. There is a washer and dryer located in the HP area that is periodically used to clean personal garments. The potential release from this mechanism is considered negligible. Therefore, since there are no full scale onsite laundry facilities, consistent with the assumptions in the PWR-GALE Code[1], there are no releases assumed from this pathway. 11.2.6.4 Secondary System 11.2.6.4.1 Turbine Building Floor Drains The discharge from the Turbine Building floor drains system is monitored by a flow meter and a radiation monitor since wastes collected by the Turbine Building floor drain system may contain small quantities of radioactive materials resulting from secondary system leakage. These wastes are not treated prior to discharge. The isotopic distribution of this release is given in Table 11.2-12. RN 03-038

11.2-19 Reformatted November 2011 The discharge flow is monitored by radiation monitor number RM-L8. If radiation discharge limits are exceeded, the sump pumps are automatically tripped and the flow path may be aligned to discharge to the Excess Liquid Waste System per plant procedure HPP-710. 11.2.6.4.2 Condensate Cleanup System The Condensate Cleanup System is used during startup and as required during condenser leakage. Scheduled, infrequent operation minimizes the possibility of producing potentially radioactive waste. Based upon PWR-GALE Code[1] criteria and infrequent operation, polishers will have no impact on release calculations. 11.2.6.4.3 Steam Generator Blowdown Blowdown from the steam generators is processed as required through the NBS and returned to the condenser. There are also provisions for the discharge of treated or untreated blowdown. If the blowdown contains any significant quantity of radioactivity, radiation-monitor-controlled valves automatically divert the blowdown flow to the NBS. With the 100 lbs/day primary-to-secondary leakage used in the PWR-GALE Code[1], blowdown flow would always be diverted to the NBS. Consequently, release calculations were performed on this basis. Blowdown that is to be processed through the NBS is routed to a 14,000 gallon holdup tank and is processed through one of two demineralizer trains, each containing a mixed bed primary demineralizer and a mixed bed polishing demineralizer. Release calculations are based upon processing 45,000 lbs/hr of blowdown through 1 of the demineralizer trains and discharging 100 percent of the treated blowdown. Radioactive decay during collection in the holdup tank is calculated using a collection time of 0.01 day. This value is based upon filling the holdup tank to 40 percent capacity. No credit is taken for radioactive decay during processing or discharge. The decontamination factors used in calculating radionuclide removal are 1,000 for iodine, 100 for cesium and rubidium, and 1,000 for other nuclides. These values are based upon the decontamination factors given in Reference[1] for 2 steam generator blowdown mixed bed demineralizers in series. The isotopic distribution of this release is given in Table 11.2-12. 11.2.6.5 Adjustments to Liquid Radwaste Source Term for Anticipated Operational Occurrences The PWR-GALE Code[1] increases the calculated source term by 0.15 Ci/yr per reactor using the same isotopic distribution as the calculated source term to account for anticipated operational occurrences. This adjustment, plus the calculated tritium release, results in the releases given in Table 11.2-12. RN 03-038

11.2-20 Reformatted November 2011 11.2.6.6 Criteria for Reuse, Discharge and Recycle Processed liquids are normally discharged under the following conditions:

1.

The processed water does not satisfy plant operating requirements for water quality and tritium buildup.

2.

The effluent concentrations are within the limits specified by 10 CFR 20, Appendix B, Table II, Column 2.

3.

The discharge does not cause the limits of 10 CFR 50, Appendix I to be exceeded. Processed liquids could be reused within the plant if desired, provided that the following criteria are satisfied:

1. The plant water inventory requires makeup.
2. The water to be reused satisfies system water quality requirements.
3.

Tritium buildup is less than plant operating requirements. 11.2.7 RELEASE POINTS Flow diagrams for the systems which have the potential to release radioactive materials in liquid effluents are shown by Figures 9.3-14, 9.3-18, 10.4-7a, 10.4-13, 10.4-14, 10.4-17. The locations of the release points for three systems are as follows: Releases from the LWPS (Figure 11.2-1) and Reactor Grade System (Figure 9.3-18) are made through waste monitor tank 1 or 2 in the LWPS. Releases of detergent wastes are made through waste monitor tank 2 in the LWPS. Releases from the waste monitor tanks are piped to the penstocks of the Fairfield Pumped Storage Facility where they are diluted by the water released to the Broad River during the generating portion of the pumped hydro cycle. Planned liquid releases are made only during the generating portion of the cycle. Figure 11.2-3 shows the location of this release point relative to the plant site. Figure 11.2-4 is a close-up of the facility. If processed steam generator blowdown (see Figure 10.4-14) is released to the environment, it is piped to the penstocks of the Fairfield Pumped Storage Facility where it is diluted by the water released through the dam during the generating portion of the cycle. If unprocessed steam generator blowdown (see Figure 10.4-13) is released to the environment, it is released through the circulating water discharge canal. During startup, blowdown may be routed to the alum sludge lagoon in the Industrial Waste System. Releases from this system are routed to the discharge canal by a 24 inch pipe which terminates at the circulating water discharge structure. The circulating water discharge canal is shown by Figure 2.4-1. RN 03-038

11.2-21 Reformatted November 2011 Liquid effluents from the Turbine Building floor drains (Figure 9.3-15) are released through a collecting sump in the Industrial Waste System which, when discharged, is routed to the circulating water discharge canal. 11.2.8 DILUTION FACTORS Concentrations of radioactive effluents in waters affected by operation of the plant were calculated according to the methods set forth in Regulatory Guide 1.113. The specific rationale utilized is as follows. During the daily generating phase of the Fairfield Pumped Storage Facility, 29,000 acre-feet of water from Monticello Reservoir will flow into Parr Reservoir. A similar amount will later be pumped back into Monticello Reservoir during periods of off-peak power demand; this volume of water is equivalent to 14,620 cfs over a 24 hour period. In contrast, the average stream flow of the Broad River at Parr Dam is about 5,600 cfs, amounting to less than 40 percent of the daily volume of water exchanged. Thus, mixing between the 2 reservoirs is expected to be relatively complete. This expectation has been confirmed by physical model studies conducted by Alden Research Laboratories[2], which indicate that effluents present in the plant discharge achieve nearly equal steady-state concentrations throughout the Parr-Monticello system. Therefore, it is preferable to consider the system as a single reservoir unit for the purposes of aquatic modeling. Several analytical models suitable for use in simulation of reservoirs and cooling ponds are described in Appendix A, Section 5, of Regulatory Guide 1.113. For the Parr-Monticello System, the pond blowdown rate, Qb, is equivalent to the flow rate of the Broad River. A flow rate of 5,600 cfs, equal to the average stream flow of the Broad River at Parr Dam, was used. The plant pumping rate, Qp, is the plant condenser water recirculation rate. A flow rate of 480,000 gpm, or 1,069 cfs, was utilized as an appropriate value under average conditions. The resulting recirculation factor, R, is determined by the ratio Qb/Qp and has a value of 5.24. Review of Figure 12 from Regulatory Guide 1.113 shows that with the above determined values of 5.24 for R and a flushing time, Vt/Qb, of 36.2 days, that the partially mixed flow model approaches the plug flow model for relatively long lived isotopes. Conservatively ignoring credit for radiological decay and use of the plug flow model is conservative for short lived isotopes. Therefore, conservatively ignoring radiological decay, the dilution factor simply relates to the final flow or 5,600 cfs, equal to the average stream flow of the Broad River at Parr Dam (FSAR Table 2.4-3). RN 03-038

11.2-22 Reformatted November 2011 In addition to the limits for each isotope, the requirements of 10 CFR 20 state that, for a mixture of radionuclide, the following relationship must hold: 1 1

= ECiL i C N i Where: Ci = concentration of radionuclide I. ECiL = effluent concentration limit (ECL) of radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2. N = number of radionuclides in the mixture. The sum of the ratios of expected radionuclide concentrations to their ECL values for the mixture defined by the second column of Table 11.2-13 is 6.8 x 10-4, which is less than unity, as required. 11.2.9 ESTIMATED DOSES Potential pathways of exposure of man to radioactive materials in liquid effluents from the Virgil C. Summer Nuclear Station are identified and discussed in Section 11.6.2. Doses to individuals in the environs of the plant from each of the potentially significant pathways were calculated; methodology for and results of the calculations are discussed in the following paragraphs. All results presented in these sections were obtained using the calculational techniques prescribed in Regulatory Guide 1.109. Except where noted in discussion of doses for specific pathways, all usage and consumption values, transport times, bioaccumulation factors, dose conversion factors, and other constants utilized were those suggested in Regulatory Guide 1.109. Dilution factors for liquid pathways were calculated according to the methods of Regulatory Guide 1.113, as discussed in Section 11.2.8. Doses to individuals were calculated for drinking water, fish consumption, and recreational activity (swimming, boating, shoreline activity) pathways. Assumptions, including point of exposure, are described for each pathway in the following paragraphs; the calculated liquid pathway doses are summarized in Table 11.2-14. Each dose was calculated at the location of the highest dose offsite at which the pathway could be assumed to exist. RN 03-038

11.2-23 Reformatted November 2011 The nearest downstream point of withdrawal of drinking water for human use is at Columbia, S. C., where the city of Columbia water supply system utilizes water from the Broad River. The dose to an individual obtaining his entire annual water requirement from this system was calculated. The maximum calculated dose to a single organ from this pathway was 1.4 x 10-1 mrem/yr to an infants liver; maximum whole body dose was 4.6 x 10-2 mrem/yr to an adult. Radionuclides released from the plant were assumed to be immediately available for uptake by fish. For purposes of determining the annual dose to man from the fish consumption pathway, a transport time of 0 hours was used. The maximum predicted dose to a single organ from the fish consumption pathway was 3.3 mrem/yr to a teens liver. Maximum total body dose was 1.35 mrem/yr to a teen. Exposure to an adult swimming or boating on the reservoir or engaging in recreational activity on the shore was evaluated by assuming that the individual receiving the maximum dose spends 100, 500, and 500 hours per year, respectively, in the 3 activities. Doses from these pathways were evaluated based on the average radionuclide concentrations in the Parr-Monticello system. The doses from recreational exposure are summarized in Table 11.2-15. Maximum predicted total dose to a single organ from recreational pathways was 0.15 mrem/yr to the skin. Maximum calculated total body dose was 0.12 mrem/yr. Maximum individual doses calculated as described above were used to evaluate the status of conformance of predicted liquid effluents from the Virgil C. Summer Nuclear Station with the requirements of Appendix I to 10 CFR 50. The assumptions and results of this evaluation are summarized in Table 11.2-15. It will be noted that the calculated doses indicate that liquid effluents from the plant will conform to the "as low as reasonably achievable" criteria established in Appendix I. 11.2.10 REFERENCES

1.

U. S. Nuclear Regulatory Commission, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors," NUREG-0017, April 1976.

2.

"Radioactive Diffusion Study: Parr Hydroelectric Project for South Carolina Electric and Gas Company," Alden Research Laboratories, 1973. RN 03-038

11.2-24 Reformatted November 2011 Table 11.2-1 ISOTOPIC VALUES IN THE LIQUID WASTE PROCESSING SYSTEM(1) Isotope Input to Waste Holdup Tank After Charcoal Processing After Mixed Bed Processing After Cation Processing Input to FD Tank After Charcoal Processing After Mixed Bed Processing After Cation Processing Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc AG-110M 4.20E-04 4.19E-06 4.19E-07 2.09E-08 4.20E-05 4.19E-07 4.19E-08 2.09E-09 BA-137M 2.82E-03 6.97E-27 6.97E-28 0.00E+00 2.82E-04 6.97E-28 6.97E-29 0.00E+00 BA-140 4.20E-03 3.95E-05 3.95E-06 1.97E-07 4.20E-04 3.95E-06 3.95E-07 1.97E-08 CE-141 4.80E-05 4.68E-07 4.68E-08 2.34E-09 4.80E-06 4.68E-08 4.68E-09 2.34E-10 CE-143 9.30E-04 5.46E-06 5.46E-07 2.73E-08 9.30E-05 5.46E-07 5.46E-08 2.73E-09 CE-144 1.29E-03 0.00E+00 0.00E+00 0.00E+00 1.29E-04 0.00E+00 0.00E+00 0.00E+00 CO-58 1.47E-03 1.45E-05 1.45E-06 7.27E-08 1.47E-04 1.45E-06 1.45E-07 7.27E-09 CO-60 1.68E-04 1.68E-06 1.68E-07 8.40E-09 1.68E-05 1.68E-07 1.68E-08 8.40E-10 CR-51 9.90E-04 9.62E-06 9.62E-07 4.81E-08 9.90E-05 9.62E-07 9.62E-08 4.81E-09 CS-134 2.25E-03 2.25E-04 2.25E-05 1.12E-06 2.25E-04 2.25E-05 2.25E-06 1.12E-07 CS-136 2.79E-04 2.63E-05 2.63E-06 1.31E-07 2.79E-05 2.63E-06 2.63E-07 1.31E-08 CS-137 2.97E-03 2.97E-04 2.97E-05 1.49E-06 2.97E-04 2.97E-05 2.97E-06 1.49E-07 FE-55 3.90E-04 3.90E-06 3.90E-07 1.95E-08 3.90E-05 3.90E-07 3.90E-08 1.95E-09 FE-59 9.60E-05 9.43E-07 9.43E-08 4.72E-09 9.60E-06 9.43E-08 9.43E-09 4.72E-10 H-3 3.00E-01 3.00E-01 3.00E-01 1.50E-01 3.00E-02 3.00E-02 3.00E-02 1.50E-02 I-131 1.50E-02 1.36E-04 1.36E-05 6.82E-06 1.50E-03 1.36E-05 1.36E-06 6.82E-07 I-132 7.80E-02 2.74E-05 2.74E-06 1.37E-06 7.80E-03 2.74E-06 2.74E-07 1.37E-07 I-133 4.80E-02 2.15E-04 2.15E-05 1.08E-05 4.80E-03 2.15E-05 2.15E-06 1.08E-06 I-134 1.26E-01 3.95E-06 3.95E-07 1.97E-07 1.26E-02 3.95E-07 3.95E-08 1.97E-08 I-135 9.30E-02 1.35E-04 1.35E-05 6.73E-06 9.30E-03 1.35E-05 1.35E-06 6.73E-07 LA-140 8.10E-03 6.64E-05 6.64E-06 3.32E-07 8.10E-04 6.64E-06 6.64E-07 3.32E-08 MN-54 5.10E-04 5.09E-06 5.09E-07 2.54E-08 5.10E-05 5.09E-07 5.09E-08 2.54E-09 MO-99 2.10E-03 1.59E-05 1.59E-06 7.95E-08 2.10E-04 1.59E-06 1.59E-07 7.95E-09 NA-24 1.59E-02 7.65E-10 7.65E-11 3.82E-12 1.59E-03 7.65E-11 7.65E-12 3.82E-13 NB-95 9.00E-05 9.08E-07 9.08E-08 4.54E-09 9.00E-06 9.08E-08 9.08E-09 4.54E-10 RN 03-038

11.2-25 Reformatted November 2011 Table 11.2-1 (Continued) ISOTOPIC VALUES IN THE LIQUID WASTE PROCESSING SYSTEM(1) Isotope Input to Waste Holdup Tank After Charcoal Processing After Mixed Bed Processing After Cation Processing Input to FD Tank After Charcoal Processing After Mixed Bed Processing After Cation Processing Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc Ci/cc NB-95m 0.00E+00 2.16E-09 2.16E-10 1.08E-11 0.00E+00 2.16E-10 2.16E-11 1.08E-12 NP-239 7.20E-04 5.21E-06 5.21E-07 2.61E-08 7.20E-05 5.21E-07 5.21E-08 2.61E-09 PR-143 1.11E-03 1.05E-05 1.05E-06 5.24E-08 1.11E-04 1.05E-06 1.05E-07 5.24E-09 PR-144 1.47E-03 0.00E+00 0.00E+00 0.00E+00 1.47E-04 0.00E+00 0.00E+00 0.00E+00 RH-103M 2.76E-03 2.36E-05 2.36E-06 1.18E-07 2.76E-04 2.36E-06 2.36E-07 1.18E-08 RH-106 3.30E-02 0.00E+00 0.00E+00 0.00E+00 3.30E-03 0.00E+00 0.00E+00 0.00E+00 RU-103 2.40E-03 2.35E-05 2.35E-06 1.18E-07 2.40E-04 2.35E-06 2.35E-07 1.18E-08 RU-106 2.88E-02 2.87E-04 2.87E-05 1.44E-06 2.88E-03 2.87E-05 2.87E-06 1.44E-07 SR-89 4.50E-05 4.43E-07 4.43E-08 2.21E-09 4.50E-06 4.43E-08 4.43E-09 2.21E-10 SR-90 3.90E-06 3.90E-08 3.90E-09 1.95E-10 3.90E-07 3.90E-09 3.90E-10 1.95E-11 SR-91 3.30E-04 7.28E-07 7.28E-08 3.64E-09 3.30E-05 7.28E-08 7.28E-09 3.64E-10 TC-99 0.00E+00 1.86E-13 1.86E-14 9.29E-16 0.00E+00 1.86E-14 1.86E-15 9.29E-17 TC-99M 1.65E-03 1.47E-05 1.47E-06 7.33E-08 1.65E-04 1.47E-06 1.47E-07 7.33E-09 TE-129 8.70E-03 9.81E-07 9.81E-08 4.90E-09 8.70E-04 9.81E-08 9.81E-09 4.90E-10 TE-129M 6.00E-05 5.86E-07 5.86E-08 2.93E-09 6.00E-06 5.86E-08 5.86E-09 2.93E-10 TE-131 2.85E-03 7.47E-07 7.47E-08 3.74E-09 2.85E-04 7.47E-08 7.47E-09 3.74E-10 TE-131M 5.10E-04 2.85E-06 2.85E-07 1.43E-08 5.10E-05 2.85E-07 2.85E-08 1.43E-09 TE-132 5.40E-04 4.27E-06 4.27E-07 2.13E-08 5.40E-05 4.27E-07 4.27E-08 2.13E-09 W-187 8.40E-04 4.07E-06 4.07E-07 2.04E-08 8.40E-05 4.07E-07 4.07E-08 2.04E-09 Y-90 4.50E-07 3.38E-09 3.38E-10 1.69E-11 4.50E-08 3.38E-10 3.38E-11 1.69E-12 Y-91 1.65E-06 3.41E-08 3.41E-09 1.71E-10 1.65E-07 3.41E-09 3.41E-10 1.71E-11 Y-91M 1.68E-04 5.01E-07 5.01E-08 2.51E-09 1.68E-05 5.01E-08 5.01E-09 2.51E-10 Y-93 1.44E-03 3.45E-06 3.45E-07 1.73E-08 1.44E-04 3.45E-07 3.45E-08 1.73E-09 ZN-65 1.62E-04 1.61E-06 1.61E-07 8.07E-09 1.62E-05 1.61E-07 1.61E-08 8.07E-10 ZR-95 1.26E-04 1.24E-06 1.24E-07 6.22E-09 1.26E-05 1.24E-07 1.24E-08 6.22E-10 RN 03-038

11.2-26 Reformatted November 2011 Table 11.2-1 (Continued) ISOTOPIC VALUES IN THE LIQUID WASTE PROCESSING SYSTEM(1) Isotope Waste Holdup Tank Floor Drain Tank Charcoal Bed (2) Mixed Bed (2) Cation (2) Waste Monitor Tank (1) Ci Ci Ci Ci Ci Ci AG-110M 1.27E-02 1.27E-03 2.05E-02 1.86E-04 1.86E-05 1.90E-07 BA-137M 8.54E-02 8.54E-03 1.32E-25 1.20E-27 0.00E+00 0.00E+00 BA-140 1.27E-01 1.27E-02 1.62E-01 1.48E-03 1.48E-04 1.79E-06 CE-141 1.45E-03 1.45E-04 2.15E-03 1.95E-05 1.95E-06 2.13E-08 CE-143 2.82E-02 2.82E-03 1.02E-02 9.28E-05 9.28E-06 2.48E-07 CO-58 3.91E-02 3.91E-03 6.94E-02 6.31E-04 6.31E-05 0.00E+00 CO-60 4.45E-02 4.45E-03 8.30E-03 7.54E-05 7.54E-06 6.60E-07 CR-51 5.09E-03 5.09E-04 4.36E-02 3.96E-04 3.96E-05 7.63E-08 CS-134 3.00E-02 3.00E-03 1.01E-01 1.01E-02 1.01E-03 4.37E-07 CS-136 6.81E-02 6.81E-03 9.85E-03 9.85E-04 9.85E-05 1.02E-05 CS-137 8.45E-03 8.45E-04 1.34E-01 1.34E-02 1.34E-03 1.19E-06 FE-55 8.99E-02 8.99E-03 1.92E-02 1.75E-04 1.75E-05 1.35E-05 FE-59 1.18E-02 1.18E-03 4.41E-03 4.01E-05 4.01E-06 1.77E-07 I-131 2.91E-03 2.91E-04 5.10E-01 4.64E-03 6.71E-07 4.28E-08 I-132 9.08E+00 9.08E-01 3.08E-02 2.80E-04 7.67E-06 1.36E+00 I-133 4.54E-01 4.54E-02 3.46E-01 3.15E-03 0.00E+00 6.20E-05 I-134 2.36E+00 2.36E-01 1.52E-03 1.38E-05 0.00E+00 1.24E-05 I-135 1.45E+00 1.45E-01 1.70E-01 1.55E-03 0.00E+00 9.78E-05 LA-140 3.82E+00 3.82E-01 2.30E-01 2.09E-03 2.09E-04 1.79E-06 MN-54 2.82E+00 2.82E-01 2.50E-02 2.27E-04 2.27E-05 6.11E-05 MO-99 2.45E-01 2.45E-02 4.03E-02 3.67E-04 3.67E-05 3.02E-06 NA-24 1.54E-02 1.54E-03 8.53E-08 7.75E-10 7.75E-11 2.31E-07 NB-95 6.36E-02 6.36E-03 4.59E-03 4.17E-05 4.17E-06 7.23E-07 NB-95m 4.82E-01 4.82E-02 2.87E-05 2.61E-07 2.61E-08 3.47E-11 NP-239 2.73E-03 2.73E-04 1.23E-02 1.12E-04 1.12E-05 4.12E-08 PR-143 0.00E+00 0.00E+00 4.35E-02 3.96E-04 3.96E-05 9.82E-11 RN 03-038

11.2-27 Reformatted November 2011 Table 11.2-1 (Continued) ISOTOPIC VALUES IN THE LIQUID WASTE PROCESSING SYSTEM(1) Isotope Waste Holdup Tank Floor Drain Tank Charcoal Bed (2) Mixed Bed (2) Cation (2) Waste Monitor Tank (1) Ci Ci Ci Ci Ci Ci RH-103M 2.18E-02 2.18E-03 1.09E-01 9.95E-04 9.94E-05 2.37E-07 RU-103 3.36E-02 3.36E-03 1.09E-01 9.93E-04 9.93E-05 4.76E-07 RU-106 4.45E-02 4.45E-03 1.41E+00 1.28E-02 1.28E-03 0.00E+00 SR-89 8.36E-02 8.36E-03 2.09E-03 1.90E-05 1.90E-06 1.07E-06 SR-90 9.99E-01 9.99E-02 1.93E-04 1.75E-06 1.75E-07 0.00E+00 SR-91 7.27E-02 7.27E-03 9.97E-04 9.06E-06 9.06E-07 1.07E-06 TC-99 8.72E-01 8.72E-02 2.38E-09 2.17E-11 2.17E-12 1.31E-05 TC-99M 1.36E-03 1.36E-04 3.78E-02 3.43E-04 3.43E-05 2.01E-08 TE-129 1.18E-04 1.18E-05 9.14E-04 8.31E-06 8.31E-07 1.77E-09 TE-129M 9.99E-03 9.99E-04 2.69E-03 2.45E-05 2.45E-06 3.31E-08 TE-131 0.00E+00 0.00E+00 1.15E-03 1.05E-05 1.05E-06 8.44E-15 TE-131M 5.00E-02 5.00E-03 5.14E-03 4.67E-05 4.67E-06 6.66E-07 TE-132 2.63E-01 2.63E-02 1.17E-02 1.06E-04 1.06E-05 4.45E-08 W-187 1.82E-03 1.82E-04 6.78E-03 6.16E-05 6.16E-06 2.66E-08 Y-90 8.63E-02 8.63E-03 8.48E-06 7.71E-08 7.71E-09 3.39E-08 Y-91 1.54E-02 1.54E-03 1.80E-04 1.63E-06 1.63E-07 1.29E-07 Y-91M 1.64E-02 1.64E-03 6.87E-04 6.25E-06 6.25E-07 1.94E-07 Y-93 2.54E-02 2.54E-03 4.79E-03 4.36E-05 4.36E-06 1.85E-07 ZN-65 1.36E-05 1.36E-06 7.91E-03 7.19E-05 7.19E-06 1.54E-10 ZR-95 5.00E-05 5.00E-06 5.92E-03 5.38E-05 5.38E-06 1.55E-09 (1) System modeled with first three components in Duratek processing system (Charcoal processing, mixed bed resin demineralizer and cation demineralizer). No credit taken for additional removal prior to waste monitor tank. (2) Activity values based on 50% waste holdup tank processing and 50% floor drain tank processing. RN 03-038

11.2-28 Reformatted November 2011 TABLE 11.2-2 EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component PUMPS

1.

Reactor Coolant Drain Tank Pumps Number 2 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 100 (140) (1) 2 100 (140) (1) Design head, ft 1 300 (250) (1) 2 300 (250) (1) Material Stainless Steel

2.

Waste Evaporator Feed Pump (Used As Waste Holdup Tank Pump) Number 1 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200 Material Stainless Steel

3.

Waste Evaporator Condensate Pump Number 1 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200 Material Stainless Steel (1) Also serves as spent resin sluice pump 02-01 02-01 02-01 RN 03-038

11.2-29 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

4.

Chemical Drain Tank Pump Number 1 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200 Material Stainless Steel

5.

Spent Resin Sluice Pump Number 1 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 140 (100) (2) 2 140 (100) (2) Design head, ft 1 250 (300) (2) 2 250 (300) (2) Material Stainless Steel

6.

Laundry and Hot Shower Tank Pump Number 1 Type Mechanical Seal Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200 Material Stainless Steel (2) Also serves as reactor coolant drain tank pump. 02-01

11.2-30 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

7.

Floor Drain Tank Pump Number 1 Type Mechanical Seal Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200

8.

Waste Monitor Tank Pumps Number 2 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 1 35 2 100 Design head, ft 1 250 2 200 Material Stainless Steel

9.

Excess Liquid Waste Pumps Number 2 Type Canned Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 Design Head, ft 315 02-01 02-01 02-01

11.2-31 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component HEAT EXCHANGERS

1.

Reactor Coolant Drain Tank Heat Exchanger Number 1 Type U-tube Estimated UA, BTU/hr°F 70,000 Design pressure, psi, Shell 150 Tube 150 Design temperature, °F, Shell 250 Tube 200 Design flow, lb/hr, Shell 112,000 Tube 44,600 Temperature in, °F, Shell 105 Tube 180 Temperature out, °F Shell 125 Tube 130 Material Shell Carbon Steel Tube Stainless Steel TANKS

1.

Reactor Coolant Drain Tank Number 1 Usable volume, gallons 350 Type Horizontal Design pressure, psig (3) 100 Design temperature, °F 250 Material Stainless Steel Diaphragm No

2.

Waste Holdup Tank Number 1 Usable volume, gallons 10,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No (3) External design pressure - 60 psig. 02-01 02-01

11.2-32 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

3.

Waste Evaporator Condensate Tank Number 1 Usable volume, gallons 5,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm Yes

4.

Chemical Drain Tank Number 1 Usable volume, gallons 600 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No

5.

Spent Resin Storage Tank Number 1 Usable volume, ft3 (4) 350 Type Vertical Design pressure, psig 150 Design temperature, °F 200 Radiation level inside compartment, R/hr 1000 Material Stainless Steel Diaphragm No

6.

Laundry and Hot Shower Tank Number 1 Usable volume, gallons 10,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No (4) Total for resin and liquid.

11.2-33 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

7.

Floor Drain Tank Number 1 Usable volume, gallons 10,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No

8.

Waste Monitor Tanks Number 2 Usable volume, gallons 5,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No

9.

Waste Evaporator Reagent Tank (No Longer In Service) Number 1/evaporator Usable volume, gallons 5 Type Vertical Design pressure, psig 150 Design temperature, °F 200 Material Stainless Steel Diaphragm No

10.

Excess Waste Holdup Tank Number 1 Usable volume, gallons 10,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No RN 03-038

11.2-34 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

11.

Decontamination Pit Collection Tank Number 1 Usable volume, gallons 10,000 Type Vertical Design pressure, psig Atmosphere Design temperature, °F 200 Material Stainless Steel Diaphragm No DEMINERALIZERS

1.

Waste Evaporator Condensate Demineralizer Number 1 Type Flushable Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 Resin volume, ft3 NA Material Stainless Steel Purification media None Design process decontamination factor NA

2.

Waste Monitor Tank Demineralizer Number 1 Type Flushable Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 Resin volume, ft3 30 Material Stainless Steel Purification media Anion and/or cation exchange resins as required Design process decontamination factor 10 RN 03-038

11.2-35 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

3.

Excess Liquid Waste Demineralizers Number 2 Type Flushable Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 Resin volume, ft3 30 Material Stainless Steel Purification media Anion and/or cation exchange resins as required

4.

Duratek Demineralizer Vessels Number 5 Type Flushable Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 25 Resin volume, ft3 19.5 to 45 Material Stainless Steel Purification media Charcoal Bed, Mixed Bed and/or cation exchange resins as required

5.

Polishing Vessel Number 1 Type Flushable Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 25 Resin volume, ft3 30 Material Stainless Steel Purification media Mixed Bed and Cation RN 03-038

11.2-36 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component FILTERS

1.

Waste Evaporator Condensate Filter Number 1 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 5 Size of particles, 98% ret., microns (nominal) 25 Materials Housing Filter element Stainless Steel EICF (5) Component

2.

Spent Resin Sluice Filter Number 1 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 250 P at design flow, psi 5 Size of particles, 98% ret., microns (nominal) 25 Surface radiation level, R/hr 100 Materials Housing Filter element Stainless Steel EICF (5)

3.

Laundry and Hot Shower Tank Filter Number 1 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 5 Size of particles, 98% ret., microns (nominal) 25 Surface radiation level, R/hr 100 Materials Housing Filter element Stainless Steel EICF (5) Epoxy Impregnated Cellulose Fiber or glass fiber. RN 03-038 RN 03-038 RN 03-038

11.2-37 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component

4.

Floor Drain Tank Filter Number 1 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 5 Size of particles, 98% ret., microns (nominal) 25 Surface of radiation level, R/hr 100 Materials Housing Filter element Stainless Steel EICF Component

5.

Waste Monitor Tank Filter Number 1 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 5 Size of particles, 98% ret., microns (nominal) 25 Surface radiation level, R/hr 90 Materials Housing Filter element Stainless Steel EICF

6.

Excess Liquid Waste Filters Number 2 Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi <5 Size of particles, 98% ret., microns (nominal) 20 Surface radiation level, mR/hr <100 Materials Housing Filter element Stainless Steel Stainless Steel

11.2-38 Reformatted November 2011 TABLE 11.2-2 (Continued) EQUIPMENT PRINCIPAL DESIGN PARAMETERS Component STRAINERS

1.

Laundry and Hot Shower Tank Strainer Number 1 Type Basket Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 0.2 (Estimated) Nominal rating, inch 0.0625 Surface radiation level Negligible Materials Stainless Steel

2.

Floor Drain Tank Strainer Number 1 Type Basket Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 35 P at design flow, psi 0.2 (Estimated) Nominal rating, inch 0.0625 Surface radiation level Negligible Materials Stainless Steel

3.

Decontamination Pit Collection Tank Strainer Number 1 Type Duplex Basket Design pressure, psig 150 Design temperature, °F 200 Design flow, gpm 350 Nominal rating, inch 0.015 Material Stainless Steel EVAPORATORS

1.

Waste Evaporator (No Longer In Service) Number 1 Design flow, gpm 15 Feed conc., ppm boron 10 - 2500 Bottoms conc., ppm boron 7000 - 21,000 Design process decontamination factor 1000 Steam design pressure, psig 50 02-01 02-01 RN 03-038

11.2-39 Reformatted November 2011 TABLE 11.2-3 LIQUID WASTE PROCESSING SYSTEM MAJOR COMPONENT INVENTORIES Reactor Waste Waste Duratek Duratek Duratek Reactor Excess Coolant Monitor Monitor Charcoal Mixed Bed Cation Coolant Waste Floor Waste Decontamination Concentration Tank Tank Demineralizer Demineralizer Demineralizer Drain Holdup Drain Holdup Pit Collection at 1% Fuel Number 1 Number 2 Tank Tank Tank Tank Tank Defect (4000 gal) (4000 gal) (45 cu ft) (28 cu ft resin) (28 cu ft) (280 gal) (8000 gal) (8000 gal) (10,000 gal) (10,000 gal) Isotope Ci/cc Curies Curies Curies Curies Curies Curies Curies Curies Curies Curies H-3 3.50E+00 5.30E+01 3.71E+00 1.06E+02 1.06E+02 AG-110M 3.00E-03 4.54E-02 2.66E-01 2.42E-03 2.42E-04 3.18E-03 9.08E-02 9.08E-02 BA-137M 2.00E+00 3.03E+01 1.08E+02 1.07E+01 1.07E+00 2.12E+00 6.06E+01 6.06E+01 BA-140 4.40E-03 6.66E-02 3.10E-01 2.81E-03 2.81E-04 4.66E-03 1.33E-01 1.33E-01 BR-83 8.90E-02 1.35E+00 4.38E-02 3.98E-04 0.00E+00 9.43E-02 2.70E+00 2.70E+00 BR-85 5.20E-03 7.87E-02 6.48E-23 5.89E-25 0.00E+00 5.51E-03 1.57E-01 1.57E-01 CE-141 6.90E-04 1.04E-02 5.62E-02 5.11E-04 5.11E-05 7.31E-04 2.09E-02 2.09E-02 CE-143 5.20E-04 7.87E-03 1.04E-02 9.44E-05 9.44E-06 5.51E-04 1.57E-02 1.57E-02 CE-144 4.70E-04 7.12E-03 0.00E+00 0.00E+00 0.00E+00 4.98E-04 1.42E-02 1.42E-02 CO-58 1.40E-02 2.12E-01 1.20E+00 1.09E-02 1.09E-03 1.48E-02 4.24E-01 4.24E-01 CO-60 1.30E-03 1.97E-02 1.17E-01 1.06E-03 1.06E-04 1.38E-03 3.94E-02 3.94E-02 CR-51 5.50E-03 8.33E-02 4.40E-01 4.00E-03 4.00E-04 5.83E-03 1.67E-01 1.67E-01 CS-134 4.40E+00 6.66E+01 3.58E+02 3.58E+01 3.58E+00 4.66E+00 1.33E+02 1.33E+02 CS-136 4.50E+00 6.81E+01 2.89E+02 2.89E+01 2.89E+00 4.77E+00 1.36E+02 1.36E+02 CS-137 2.10E+00 3.18E+01 1.72E+02 1.72E+01 1.72E+00 2.23E+00 6.36E+01 6.36E+01 FE-55 2.40E-03 3.63E-02 2.15E-01 1.96E-03 1.96E-04 2.54E-03 7.27E-02 7.27E-02 FE-59 5.20E-04 7.87E-03 4.35E-02 3.95E-04 3.95E-05 5.51E-04 1.57E-02 1.57E-02 I-130 3.30E-02 5.00E-01 2.50E-01 2.27E-03 0.00E+00 3.50E-02 9.99E-01 9.99E-01 I-131 3.00E+00 4.54E+01 1.85E+02 1.68E+00 6.93E-05 3.18E+00 9.08E+01 9.08E+01 I-132 3.10E+00 4.69E+01 1.30E+01 1.18E-01 7.49E-03 3.29E+00 9.39E+01 9.39E+01 I-133 4.60E+00 6.97E+01 6.04E+01 5.49E-01 0.00E+00 4.88E+00 1.39E+02 1.39E+02 I-134 6.00E-01 9.08E+00 1.32E-02 1.20E-04 0.00E+00 6.36E-01 1.82E+01 1.82E+01 I-135 2.40E+00 3.63E+01 7.98E+00 7.25E-02 0.00E+00 2.54E+00 7.27E+01 7.27E+01 RN 03-038

11.2-40 Reformatted November 2011 TABLE 11.2-3 (Continued) LIQUID WASTE PROCESSING SYSTEM MAJOR COMPONENT INVENTORIES Reactor Waste Waste Duratek Duratek Duratek Reactor Excess Coolant Monitor Monitor Charcoal Mixed Bed Cation Coolant Waste Floor Waste Decontamination Concentration Tank Tank Demineralizer Demineralizer Demineralizer Drain Holdup Drain Holdup Pit Collection at 1% Fuel Number 1 Number 2 Tank Tank Tank Tank Tank Defect (4000 gal) (4000 gal) (45 cu ft) (28 cu ft resin) (28 cu ft) (280 gal) (8000 gal) (8000 gal) (10,000 gal) (10,000 gal) Isotope Ci/cc Curies Curies Curies Curies Curies Curies Curies Curies Curies Curies LA-140 1.40E-03 2.12E-02 2.69E-01 2.45E-03 2.45E-04 1.48E-03 4.24E-02 4.24E-02 MN-54 4.10E-04 6.21E-03 3.65E-02 3.32E-04 3.32E-05 4.35E-04 1.24E-02 1.24E-02 MN-56 2.20E-02 3.33E-01 1.29E-02 1.17E-04 1.17E-05 2.33E-02 6.66E-01 6.66E-01 MO-99 7.90E-01 1.20E+01 2.76E+01 2.51E-01 2.51E-02 8.37E-01 2.39E+01 2.39E+01 NB-95 6.70E-04 1.01E-02 6.01E-02 5.46E-04 5.46E-05 7.10E-04 2.03E-02 2.03E-02 NB-95m 0.00E+00 0.00E+00 2.78E-04 2.53E-06 2.53E-07 0.00E+00 0.00E+00 0.00E+00 PR-143 6.20E-04 9.39E-03 4.72E-02 4.29E-04 4.29E-05 6.57E-04 1.88E-02 1.88E-02 RB-86 3.60E-02 5.45E-01 2.48E+00 2.48E-01 2.48E-02 3.82E-02 1.09E+00 1.09E+00 RB-89 1.80E-01 2.73E+00 1.76E-06 1.76E-07 1.76E-08 1.91E-01 5.45E+00 5.45E+00 RH-103M 6.90E-04 1.04E-02 5.30E-02 4.82E-04 4.82E-05 7.31E-04 2.09E-02 2.09E-02 RH-106 2.10E-04 3.18E-03 1.87E-02 1.70E-04 1.70E-05 2.23E-04 6.36E-03 6.36E-03 RU-103 6.40E-04 9.69E-03 5.30E-02 4.82E-04 4.82E-05 6.78E-04 1.94E-02 1.94E-02 RU-106 2.10E-04 3.18E-03 1.87E-02 1.70E-04 1.70E-05 2.23E-04 6.36E-03 6.36E-03 SR-89 4.00E-03 6.06E-02 3.40E-01 3.09E-03 3.09E-04 4.24E-03 1.21E-01 1.21E-01 SR-90 2.00E-04 3.03E-03 1.80E-02 1.63E-04 1.63E-05 2.12E-04 6.06E-03 6.06E-03 SR-91 5.30E-03 8.03E-02 3.02E-02 2.75E-04 2.75E-05 5.62E-03 1.61E-01 1.61E-01 SR-92 1.20E-03 1.82E-02 7.86E-04 7.15E-06 7.15E-07 1.27E-03 3.63E-02 3.63E-02 TC-99 0.00E+00 0.00E+00 1.69E-06 1.54E-08 1.54E-09 0.00E+00 0.00E+00 0.00E+00 TC-99M 8.40E-01 1.27E+01 2.65E+01 2.41E-01 2.41E-02 8.90E-01 2.54E+01 2.54E+01 TE-125M 4.70E-04 7.12E-03 4.00E-02 3.63E-04 3.63E-05 4.98E-04 1.42E-02 1.42E-02 TE-127 1.50E-02 2.27E-01 8.08E-02 7.35E-04 7.35E-05 1.59E-02 4.54E-01 4.54E-01 RN 03-038

11.2-41 Reformatted November 2011 TABLE 11.2-3 (Continued) LIQUID WASTE PROCESSING SYSTEM MAJOR COMPONENT INVENTORIES Reactor Waste Waste Duratek Duratek Duratek Reactor Excess Coolant Monitor Monitor Charcoal Mixed Bed Cation Coolant Waste Floor Waste Decontamination Concentration Tank Tank Demineralizer Demineralizer Demineralizer Drain Holdup Drain Holdup Pit Collection at 1% Fuel Number 1 Number 2 Tank Tank Tank Tank Tank Defect (4000 gal) (4000 gal) (45 cu ft) (28 cu ft resin) (28 cu ft) (280 gal) (8000 gal) (8000 gal) (10,000 gal) (10,000 gal) Isotope Ci/cc Curies Curies Curies Curies Curies Curies Curies Curies Curies Curies TE-127M 3.60E-03 5.45E-02 3.14E-01 2.86E-03 2.86E-04 3.82E-03 1.09E-01 1.09E-01 TE-129 2.00E-02 3.03E-01 1.19E+00 1.08E-02 1.08E-03 2.12E-02 6.06E-01 6.06E-01 TE-129M 2.10E-02 3.18E-01 1.73E+00 1.57E-02 1.57E-03 2.23E-02 6.36E-01 6.36E-01 TE-131 1.60E-02 2.42E-01 1.19E-01 1.08E-03 1.08E-04 1.70E-02 4.85E-01 4.85E-01 TE-131M 2.90E-02 4.39E-01 5.31E-01 4.83E-03 4.83E-04 3.07E-02 8.78E-01 8.78E-01 TE-132 2.90E-01 4.39E+00 1.14E+01 1.04E-01 1.04E-02 3.07E-01 8.78E+00 8.78E+00 TE-134 2.80E-02 4.24E-01 1.15E+01 1.05E-01 1.05E-02 2.97E-02 8.48E-01 8.48E-01 Y-90 5.70E-05 8.63E-04 1.31E-02 1.19E-04 1.19E-05 6.04E-05 1.73E-03 1.73E-03 Y-91 5.40E-04 8.18E-03 4.90E-02 4.46E-04 4.46E-05 5.72E-04 1.64E-02 1.64E-02 Y-91M 2.90E-03 4.39E-02 2.08E-02 1.89E-04 1.89E-05 3.07E-03 8.78E-02 8.78E-02 Y-92 1.10E-03 1.67E-02 3.15E-03 2.86E-05 2.86E-06 1.17E-03 3.33E-02 3.33E-02 ZR-95 6.70E-04 1.01E-02 5.73E-02 5.14E-04 5.21E-05 7.10E-04 2.03E-02 2.03E-02 Normal discharge path for Laundry and Hot Shower Tank - No significant activity. Not calculated Excess Waste Holdup is not normally used. If used the estimated inventories would be similar to the Waste Holdup Tank. Estimated inventories would be similar to the Floor Drain Tank. RN 03-038

11.2-42 Reformatted November 2011 TABLE 11.2-4 RANGE OF MEASURED DEMINERALIZER DECONTAMINATION FACTORS FOR SELECTED ISOTOPES Isotope Minimum Maximum I-131 1.1.x 101 1.6 x 104 I-133 1.1 x 101 1.8 x 104 I-135 1.4 x 101 2.0 x 104 Cs-137 2.4 1.3 x 103 F-18 1.73 x 101 1.5 x 103 Co-58 3.2 x 101 8.2 x 103 Mn-54 > 2.5 x 101 > 1.3 x 102 Note: Mixed bed demineralizers cation resin in lithium-7 form, anion resin in borated form.

11.2-43 Reformatted November 2011 TABLE 11.2-5 LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS F-FLOW R - Radiation Q - Flow Integrator I - Indication P - Pressure C - Control L - Level A - Alarm T - Temperature Flow Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout FI-1007 Waste Evaporator Feed Pump Discharge 150 200 0-30 gpm Local FIC-1008 Reactor Coolant Drain Tank Pump Discharge 150 250 0-250 gpm WPS Panel FIA-1009 Reactor Coolant Drain Tank Recirculation 150 250 0-250 gpm WPS Panel FICA-1011 Spent Resin Sluice Pump Discharge 150 200 1-150 gpm WPS Panel FI-1085A Waste Monitor Tank Pump Number 1 Discharge 150 200 0-100 gpm WPS Panel FI-1085B Waste Monitor Tank Pump Number 2 Discharge 150 200 0-100 gpm WPS Panel FICA-4845 Nuclear Blowdown Sluice Pump Discharge 150 200 0-250 gpm NB Panel Pressure Instrumentation PIA-1004 Reactor Coolant Drain Tank 150 250 0-30 psig WPS Panel PIA-1006A Spent Resin Storage Tank 150 200 0-100 psig WPS Panel PIA-1006B Spent Resin Storage Tank 150 200 0-100 psig Solid Waste Disposal Panel 02-01

11.2-44 Reformatted November 2011 TABLE 11.2-5 (Continued) LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS Pressure Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout PIA-4842A Nuclear Blowdown Spent Resin Storage Tank 150 200 0-160 psig Local PIA-4842B Nuclear Blowdown Spent Resin Storage Tank 150 200 0-150 psig Solid Waste Disposal Panel PI-1016 Waste Evaporator Feed Pump Filter Inlet 150 200 0-160 psig Local PI-1017 Waste Evaporator Feed Filter Outlet 150 200 0-160 psig Local PI-1018A Reactor Coolant Drain Tank Pump Number 1 Discharge 150 250 0-160 psig Local PI-1018B Reactor Coolant Drain Tank Pump Number 2 Discharge 150 250 0-160 psig Local PI-1018C Laundry and Hot Shower Tank Pump Discharge 150 200 0-160 psig Local PI-1018D Chemical Drain Tank Pump Discharge 150 200 0-160 psig Local PI-1018G Waste Evaporator Condensate Pump Discharge 150 200 0-160 psig Local PI-1074 Waste Evaporator Outlet - NO LONGER IN SERVICE -

150 200 0-160 psig Local PI-1075 Waste Evaporator Condensate Demineralizer Outlet 150 200 0-160 psig Local 02-01 02-01 RN 03-038

11.2-45 Reformatted November 2011 TABLE 11.2-5 (Continued) LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS Pressure Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout PI-1076 Waste Evaporator Condensate Filter Outlet 150 200 0-160 psig Local PI-1078 Floor Drain Tank Filter Inlet 150 200 0-160 psig Local PI-1079 Floor Drain Tank Filter Outlet 150 200 0-160 psig Local PI-1080 Laundry and Hot Shower Tank Filter Inlet 150 200 0-160 psig Local PI-1081 Laundry and Hot Shower Tank Filter Outlet 150 200 0-160 psig Local PI-1084A Waste Monitor Tank Pump Number 1 Discharge 150 200 0-160 psig Local PI-1084B Waste Monitor Tank Pump Number 2 Discharge 150 200 0-160 psig Local PI-1086 Resin Sluice Filter Inlet 150 200 0-160 psig Local PI-1087 Resin Sluice Filter Outlet 150 200 0-160 psig Local PI-1088 Waste Monitor Tank Filter Inlet 150 200 0-160 psig Local 02-01 02-01

11.2-46 Reformatted November 2011 TABLE 11.2-5 (Continued) LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS Pressure Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout PI-1089 Waste Monitor Tank Filter Outlet 150 200 0-160 psig Local PI-1090 Floor Drain Tank Pump Discharge 150 200 0-160 psig Local PI-7657 Waste Evaporator Concentrates Tank Pump Discharge 150 200 0-100 psig Local PI-7903 Excess Liquid Waste Filter Inlet 110 100 0-160 psig Local PI-7907 Excess Liquid Waste Filter Outlet 110 100 0-160 psig Local PI-7911 Excess Liquid Waste Demineralizer Filter Inlet 110 100 0-160 psig Local PI-7913 Excess Liquid Waste Demineralizer Filter Outlet 110 100 0-160 psig Local PI-7921 Excess Waste Sump Pump A Discharge 25 100 0-60 psig Local PI-7923 Excess Waste Sump Pump B Discharge 25 100 0-60 psig Local PI-7925 Excess Liquid Waste System Area Sump Pump Discharge 25 100 0-60 psig Local 02-01 02-01

11.2-47 Reformatted November 2011 TABLE 11.2-5 (Continued) LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS Level Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout LICA-1001A and 1001B Waste Holdup Tank 150 200 0-100%

Local & WPS Panels LICA-1002A and 1002B Chemical Drain Tank 150 200 0-100% Local & WPS Panels & Solid Waste Disposal Panel LICA-1003 Reactor Coolant Drain Tank 150 250 0-100% WPS Panel LICA-1005A and 1005B Spent Resin Storage Tank 150 200 0-100% WPS Panel & Solid Waste Disposal Panel LICA-1010A and 1010B Laundry and Hot Shower Tank 150 200 0-100% Local & WPS Panels LICA-1012A and 1012B Waste Evaporator Condensate Tank 150 200 0-100% Local & WPS Panels LICA-1077A and 1077B Floor Drain Tank 150 200 0-100% Local & WPS Panels LICA-1082A and 1082B Waste Monitor Tank Number 1 150 200 0-100% Local & WPS Panels LICA-1083A and B Waste Monitor Tank Number 2 150 200 0-100% Local & WPS Panels LICA-4843A Nuclear Blowdown Spent Resin Storage Tank 150 200 0-100% Solid Waste Disposal Panel LICA-4843 Nuclear Blowdown Spent Resin Storage Tank 150 200 0-72" Nuclear Blowdown Panel LICA-7651A&B Waste Evaporator Concentrates Tank 150 200 0-100% Local & Solid Waste Disposal Panel LICA-7901 Excess Waste Holdup Tank Atmosphere 100 0-100% Local 02-01 RN 04-037

11.2-48 Reformatted November 2011 TABLE 11.2-5 (Continued) LIQUID WASTE PROCESSING SYSTEM INSTRUMENTATION PRINCIPAL DESIGN PARAMETERS Level Instrumentation Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range

  • Location of Readout LICA-7905 Decontamination Pit Collection Tank Atmosphere 100 0-100%

Local Temperature Instrumentation TIA-1058 Reactor Coolant Drain Tank 150 250 50-250°F WPS Panel TI-7655 Waste Evaporator Concentrates Tank 150 200 30-180°F Local Radiation Instrumentation RML-5 Waste Discharge Line 150 100 101-106 cpm WPS & Radiation Monitor Panels & Local RML-9 Waste Discharge Line 150 100 101-106 cpm Radiation Monitor Panel & Local

  • Note:

Use of wider range instrumentation than specified is allowed provided that the required instrument accuracy in the measured range is maintained. 02-01

11.2-49 Reformatted November 2011 TABLE 11.2-6 TANK OVERFLOW PROTECTION Tank System Monitoring Annunciation Overflow of Collection Provisions Processing of Overflow Volume Control Tank CVCS Level Indication High Level Alarm Flow diverted to Recyle Holdup Tank on high level By Recycle Holdup Tank Reactor Makeup Water Storage Tank MU Level Indication High Level Alarm Overflow to Drain Via Waste Holdup Tank Boric Acid Tank CVCS Level Indication High Level Alarm Overflow to Drain Via Waste Holdup Tank Boric Acid Batching Tank CVCS Level Indication Overflow to Drain Via Waste Holdup Tank Recycle Holdup Tank BRS Level Indication High Level Alarm Overflow on Floor to Drain Via Waste Holdup Tank via Floor Sump Pump Reactor Coolant Drain Tank LWPS Level Indication High Level Alarm Overflow on Floor to Drain Via Waste Holdup Tank via Reactor Building Sump Pump Waste Holdup Tank LWPS Level Indication High Level Alarm Overflow on Floor to Drain Floor Drain Tank via Floor Sump Pump Floor Drain Tank LWPS Level Indication High Level Alarm Overflow on Floor to Drain Floor Drain Tank via Floor Sump Pump Miscellaneous Waste Drain Tank ND Level Indication High Level Alarm Overflow on Floor to Drain Floor Drain Tank via Floor Sump Pump Waste Evaporator Concentrates Tank WD Level Indication High Level Alarm Overflow on Floor to Drain Via Waste Holdup Tank 02-01 RN 03-038 02-01 RN 03-038

11.2-50 Reformatted November 2011 TABLE 11.2-6 (Continued) TANK OVERFLOW PROTECTION Tank System Monitoring Annunciation Overflow of Collection Provisions Processing of Overflow Laundry and Hot Shower Tank LWPS Level Indication High Level Alarm Overflow on Floor to Drain Flow Drain Tank via Floor Sump Pump Chemical Drain Tank LWPS Level Indication High Level Alarm Overflow on Floor to Drain Flow Drain Tank via Floor Sump Pump Waste Evaporator Condensate Tank LWPS Level Indication High Level Alarm Overflow to Drain Waste Holdup Tank via Miscellaneous Waste Holdup Tank Waste Monitor Tank LWPS Level Indication High Level Alarm Overflow to Drain Via Floor Drain Tank LWPS Spent Resin Storage Tank LWPS Level Indication High Level Alarm Relief to Drain Via Waste Holdup Tank Nuclear Blowdown Spent Resin Storage Tank NB Level Indication High Level Alarm Relief to Secondary Side Drain Channel Nuclear Blowdown Holdup Tank Refueling Water Storage Tank SF Level Indication High Level Alarm Overflow to Drain Via Waste Holdup Tank Excess Waste Holdup Tank ELWS Level Indication High Level Alarm Overflow to Area Sump Via ELWS Decontamination Pit Collection Tank ELWS Level Indication High Level Alarm Overflow to Area Sump Via ELWS Nuclear Blowdown Holdup Tank NB Level Indication High Level Alarm Overflow to Secondary Side Drain Nuclear Blowdown Holdup Tank Nuclear Blowdown Monitor Tank NB Level Indication High Level Alarm Overflow to Secondary Side Drain Nuclear Blowdown Holdup Tank 02-01 02-01 RN 03-038 RN 03-038 RN 03-038

11.2-51 Reformatted November 2011 TABLE 11.2-6a COMPARISON OF TANKS OUTSIDE CONTAINMENT WITH PROVISIONS OF BRANCH TECHNICAL POSITION ETSB 11-1 (Rev. 1), PARAGRAPH B.1.b ETSB 11-1 (Rev. 1), Paragraph B.1.b Item No. (2) Item (1) Item (2) Items (3) and (4) Tank Location (1) Elevation Monitor Alarm Overflow Drain Curb, etc. Routed to LWS Miscellaneous Waste Drain Tank AB; 374 C C C C NC C Laundry and Hot Shower Drain Tank AB; 374 C C C C NC C Floor Drain Tank AB; 374 C C C C NC C Waste Holdup Tank AB; 374 C C C C C C Chemical Drain Tank AB; 374 C C C C NC C Recycle Holdup Tank AB; 388 C C C C C C Waste Evaporator Condensate Tank AB; 388 C C C C NC C Waste Evaporator Concentrates Tank AB; 412 C C C C NC C Primary Spent Resin Tank AB; 412 C C C C C C Nuclear Blowdown Spent Resin Tank AB; 412 C C C C C C 02-01 02-01

11.2-52 Reformatted November 2011 TABLE 11.2-6a (Continued) COMPARISON OF TANKS OUTSIDE CONTAINMENT WITH PROVISIONS OF BRANCH TECHNICAL POSITION ETSB 11-1 (Rev. 1), PARAGRAPH B.1.b ETSB 11-1 (Rev. 1), Paragraph B.1.b Item No. (2) Item (1) Item (2) Items (3) and (4) Tank Location (1) Elevation Monitor Alarm Overflow Drain Curb, etc. Routed to LWS Nuclear Blowdown Holdup Tank AB; 436 C C C C C C Nuclear Blowdown Monitor Tank AB; 436 C C C C C C Boric Acid Tank AB; 463 C C C C C C Volume Control Tank AB; 463 C C C C C C Waste Monitor Tank AB; 463 C C C C NC C Excess Waste Holdup Tank FB; 412 -9 C C C C C C Decontamination Pit Collection Tank FB; 412 -9 C C C C C C Refueling Water Storage Tank YD; 412 C C C C C C Reactor Makeup Water Tank YD; 412 C C C C C C Condensate Storage Tank YD; Grade C C NC NC NC NC 02-01 02-01

11.2-53 Reformatted November 2011 TABLE 11.2-6a (Continued) COMPARISON OF TANKS OUTSIDE CONTAINMENT WITH PROVISIONS OF BRANCH TECHNICAL POSITION ETSB 11-1 (Rev. 1), PARAGRAPH B.1.b Notes:

1.

Location AB - Auxiliary Building FB - Fuel Handling Building YD - Yard; refueling water storage and reactor makeup water tanks are in an outdoor extension of the Auxiliary Building

2.

C - Conforms to ETSB 11-1 (Rev. 1) NC - Does not conform to ETSB 11-1 (Rev. 1) 02-01 02-01

11.2-54 Reformatted November 2011 Table 11.2-7 PARAMETERS USED IN THE CALCULATION OF ESTIMATED ACTIVITY IN LIQUID WASTES (1) Collector Tank With Sources Volume of Liquid Waste Collection Period Assumed Before Processing Comments Reactor Coolant Drain Tank Waste Holdup Tank (1)

1. Shim Bleeds 500 gal/day
2. Equipment drains 100 gal/day
3. WHT inputs 75 gal/day Total 675 gal/day 6.0 days Discharged Floor Drain Tank (1)
1. FDT inputs 450 gal/day 8.8 days Discharged Total 410,100 gal/yr (1)

This table does not represent system capacity. The estimated waste processing system capacity based on indefinite operation for liquids collected in the floor drain tank and waste holdup tank is 25 gpm or approximately 12,500,000 gal/yr. RN 03-038

11.2-55 Reformatted November 2011 TABLE 11.2-8 (Sheet 1) PWR-GALE CODE INPUT PARAMETERS USED IN CALCULATING RELEASES OF RADIOACTIVE MATERIALS IN LIQUID EFFLUENTS Reactor Power Level, MWt 2900 Primary System Mass of Primary Coolant, thousand lbs 377.9 Average Letdown Rate, gpm 105 Average Letdown Cation Demineralizer Flow Rate, gpm 10.5 Secondary System Steam Flow Rate, million lbs/hr 12.85 Mass of Steam/Steam Generator, thousand lbs 7.407 Mass of Liquid/Steam Generator, thousand lbs 109 Mass of Secondary Coolant, thousand lbs 2220 Steam Generator Blowdown Rate, thousand lbs/hr 45 Steam Generator Blowdown Treatment Option See Section 11.2.6.4.3 02-01 02-01 RN 03-038

11.2-56 Reformatted November 2011 TABLE 11.2-8 (Continued) (Sheet 2) PWR-GALE CODE INPUT PARAMETERS USED IN CALCULATING RELEASES OF RADIOACTIVE MATERIALS IN LIQUID EFFLUENTS Stream Flow Rate (gal/day) Fraction of Primar Coolant Activity Fraction Discharged Collection Time (days) Decay Time (days) Decontamination Factors I Cs Others Shim Bleed 1.45 x 103 NA 1.0 2.08 0.11 1 x 102 2 x 101 1 x 103 Equipment Drains 3.00 x 102 1.00 1.0 2.08 0.11 1 x 102 2 x 101 1 x 103 Clean Wastes 2.00 x 102 1.00 1.0 2.08 0.11 1 x 102 2 x 101 1 x 103 Dirty Wastes 1.34 x 103 0.05 1.0 2.97 0.11 1 x 103 2 x 102 1 x 103 Blowdown _ (2) _ (1) 1.0 0.01 0 1 x 103 1 x 102 1 x 103 Detergent Wastes _ (1) _ (1) 1.0 NA (3) NA NA NA NA (1) Calculated by PWR-GALE Code. (2) Value given on Sheet l, Table 11.2-8. (3) Not Applicable. 02-01 02-01 RN 03-038

TABLE 11.2-12 LIQUID EFFLUENTS ANNUAL RELEASES 11.2-57 Reformatted November 2011 COOLANT CONCENTRATIONS--------------------------------------------------------- ADJUSTED DETERGENT TOTAL NUCLIDE HALF-LIFE PRIMARY SECONDARY EQUIP. DRAIN MISC. WASTES SECONDARY TURB BLDG TOTAL LWS TOTAL WASTES (DAYS) (MICRO CI/ML)(MICROCI/ML) (CURIES) (CURIES) (CURIES) (CURIES) (CURIES) (CI/YR) (CI/YR) (CI/YR) CORROSION AND ACTIVATION PRODUCTS CR 51 2.78E+01 1.15E-03 1.01E-07.00069.00041.00000.00000.00110.00112.00000.00110 MN 54 3.03E+02 1.86E-04 2.41E-08.00011.00007.00000.00000.00018.00019.00000.00019 FE 55 9.50E+02 9.61E-04 8.41E-08.00059.00035.00000.00000.00095.00097.00000.00097 FE 59 4.50E+01 6.04E-04 6.19E-08.00036.00022.00000.00000.00058.00060.00000.00060 CO 58 7.13E+01 9.64E-03 8.57E-07.00586.00351.00000.00001.00938.00959.00000.00960 CO 60 1.92E+03 1.20E-03 1.08E-07.00074.00044.00000.00000.00118.00121.00000.00120 NP239 2.35E+00 7.88E-04 5.84E-08.00035.00020.00000.00000.00056.00057.00000.00057 FISSION PRODUCTS BR 83 1.00E-01 5.02E-03 1.56E-07.00101.00046.00000.00000.00147.00151.00000.00150 BR 84 2.21E-02 3.08E-03 2.99E-08.00001.00000.00000.00000.00001.00001.00000.00001 RB 86 1.87E+01 5.23E-05 5.28E-09.00355.00072.00000.00000.00426.00436.00000.00440 RB 88 1.24E-02 2.44E-01 1.28E-06.00031.00006.00000.00000.00037.00038.00000.00038 SR 89 5.20E+01 2.11E-04 2.47E-08.00013.00008.00000.00000.00020.00021.00000.00021 SR 91 4.03E-01 5.41E-04 3.00E-08.00007.00004.00000.00000.00012.00012.00000.00012 Y 91M 3.47E-02 4.16E-04 4.19E-08.00005.00003.00000.00000.00008.00008.00000.00008 Y 91 5.88E+01 3.86E-05 3.69E-09.00003.00002.00000.00000.00004.00004.00000.00004 ZR 95 6.50E+01 3.62E-05 3.68E-09.00002.00001.00000.00000.00004.00004.00000.00004 NB 95 3.50E+01 3.02E-05 3.75E-09.00002.00001.00000.00000.00003.00003.00000.00003 MO 99 2.79E+00 5.45E-02 5.49E-06.02547.01492.00000.00005.04044.04136.00000.04100 TC 99M 2.50E-01 4.35E-02 2.00E-05.02367.01409.00000.00012.03788.03874.00000.03900 RU103 3.96E+01 2.72E-05 2.49E-09.00002.00001.00000.00000.00003.00003.00000.00003 RH103M 3.96E-02 5.16E-05 2.72E-08.00002.00001.00000.00000.00003.00003.00000.00003 TE125M 5.80E+01 1.75E-05 1.11E-09.00001.00001.00000.00000.00002.00002.00000.00002 TE127M 1.09E+02 1.68E-04 1.09E-08.00010.00006.00000.00000.00016.00017.00000.00017 TE127 3.92E-01 7.11E-04 1.02E-07.00017.00010.00000.00000.00028.00028.00000.00028 TE129M 3.40E+01 8.46E-04 7.51E-08.00051.00030.00000.00000.00081.00083.00000.00083 TE129 4.79E-02 1.81E-03 7.83E-07.00033.00020.00000.00000.00053.00055.00000.00055 I130 5.17E-01 1.68E-03 1.08E-07.00299.00138.00000.00001.00438.00448.00000.00450 TE131M 1.25E+00 1.75E-03 1.29E-07.00060.00034.00000.00000.00095.00097.00000.00097 TE131 1.74E-02 1.32E-03 7.67E-07.00011.00006.00000.00000.00017.00018.00000.00018 I131 8.05E+00 1.67E-01 1.56E-05.93165.43224.00000.00152 1.36541 1.39624.00000 1.40000 TE132 3.25E+00 1.73E-02 1.39E-06.00842.00494.00000.00001.01337.01367.00000.01400 I132 9.58E-02 1.05E-01 1.16E-05.06297.02969.00000.00020.09286.09496.00000.09500 I133 8.75E-01 2.79E-01 2.04E-05.77052.35541.00000.00166 1.12759 1.15305.00000 1.20000 I134 3.67E-02 5.42E-02 7.88E-07.00106.00049.00000.00000.00155.00158.00000.00160 CS134 7.49E+02 1.51E-02 1.46E-06 1.06596.21496.00000.00001 1.28093 1.30985.00000 1.30000 I135 2.79E-01 1.69E-01 8.77E-06.15241.07011.00000.00047.22299.22803.00000.23000 CS136 1.30E+01 8.06E-03 6.84E-07.53751.10831.00000.00001.64583.66041.00000.66000 CS137 1.10E+04 1.08E-02 9.71E-07.76787.15485.00000.00001.92272.94356.00000.94000 BA137M 1.77E-03 1.98E-02 1.35E-05.71796.14478.00000.00001.86275.88223.00000.88000 BA140 1.28E+01 1.35E-04 1.20E-08.00008.00005.00000.00000.00012.00013.00000.00013 LA140 1.67E+00 1.02E-04 1.79E-08.00007.00004.00000.00000.00011.00011.00000.00011 CE141 3.24E+01 4.23E-05 3.76E-09.00003.00002.00000.00000.00004.00004.00000.00004 CE143 1.38E+00 2.77E-05 2.21E-09.00001.00001.00000.00000.00002.00002.00000.00002 PR143 1.37E+01 3.05E-05 2.66E-09.00002.00001.00000.00000.00003.00003.00000.00003 CE144 2.84E+02 1.98E-05 2.41E-09.00001.00001.00000.00000.00002.00002.00000.00002 PR144 1.20E-02 3.99E-05 3.17E-08.00001.00001.00000.00000.00002.00002.00000.00002 ALL OTHERS 4.24E-04 1.24E-08.00002.00001.00000.00000.00003.00003.00000.00003 TOTAL (EXCEPT TRITIUM) 1.22E+00 1.06E-04 5.08446 1.55406.00000.00411 6.64263 6.79263.00000 6.80000 TRITIUM RELEASE 580 CURIES PER YEAR RN 03-038

11.2-58 Reformatted November 2011 TABLE 11.2-13 COMPARISON OF RADIONUCLIDE CONCENTRATIONS IN LIQUID EFFLUENTS TO THE LIMITS OF 10 CFR 20 Isotope Annual Release Ci/yr Expected Site Boundary Concentration µCi/cc Effluent Concentration Limit - ECL (1) µCi/cc Ratio of Expected Concentration to ECL CR 51 1.1E-03 2.2E-13 5.0E-04 4.4E-10 MN 54 1.9E-04 3.8E-14 3.0E-05 1.3E-09 FE 55 9.7E-04 1.9E-13 1.0E-04 1.9E-09 FE 59 6.0E-04 1.2E-13 1.0E-05 1.2E-08 CO 58 9.6E-03 1.9E-12 2.0E-05 9.6E-08 CO 60 1.2E-03 2.4E-13 3.0E-06 8.0E-08 NP239 5.7E-04 1.1E-13 2.0E-05 5.7E-09 BR 83 1.5E-03 3.0E-13 9.0E-04 3.3E-10 BR 84 1.0E-05 2.0E-15 4.0E-04 5.0E-12 RB 86 4.4E-03 8.8E-13 7.0E-06 1.3E-07 RB 88 3.8E-04 7.6E-14 4.0E-04 1.9E-10 SR 89 2.1E-04 4.2E-14 8.0E-06 5.3E-09 SR 91 1.2E-04 2.4E-14 2.0E-05 1.2E-09 Y 91M 8.0E-05 1.6E-14 2.0E-03 8.0E-12 Y 91 4.0E-05 8.0E-15 8.0E-06 1.0E-09 ZR 95 4.0E-05 8.0E-15 2.0E-05 4.0E-10 NB 95 3.0E-05 6.0E-15 3.0E-05 2.0E-10 MO 99 4.1E-02 8.2E-12 2.0E-05 4.1E-07 TC 99M 3.9E-02 7.8E-12 1.0E-03 7.8E-09 RU103 3.0E-05 6.0E-15 3.0E-05 2.0E-10 RH103M 3.0E-05 6.0E-15 6.0E-03 1.0E-12 TE125M 2.0E-05 4.0E-15 2.0E-05 2.0E-10 TE127M 1.7E-04 3.4E-14 9.0E-06 3.8E-09 TE127 2.8E-04 5.6E-14 1.0E-04 5.6E-10 TE129M 8.3E-04 1.7E-13 7.0E-06 2.4E-08 TE129 5.5E-04 1.1E-13 4.0E-04 2.8E-10 I130 4.5E-03 9.0E-13 2.0E-05 4.5E-08 TE131M 9.7E-04 1.9E-13 8.0E-06 2.4E-08 TE131 1.8E-04 3.6E-14 8.0E-05 4.5E-10 I131 1.4E+00 2.8E-10 1.0E-05 2.8E-05 TE132 1.4E-02 2.8E-12 9.0E-06 3.1E-07 I132 9.5E-02 1.9E-11 1.0E-04 1.9E-07 I133 1.2E+00 2.4E-10 7.0E-06 3.4E-05 I134 1.6E-03 3.2E-13 4.0E-04 8.0E-10 CS134 1.3E+00 2.6E-10 9.0E-07 2.9E-04 I135 2.3E-01 4.6E-11 3.0E-05 1.5E-06 CS136 6.6E-01 1.3E-10 6.0E-06 2.2E-05 CS137 9.4E-01 1.9E-10 1.0E-06 1.9E-04 BA140 1.3E-04 2.6E-14 8.0E-06 3.3E-09 LA140 1.1E-04 2.2E-14 9.0E-06 2.4E-09 CE141 4.0E-05 8.0E-15 3.0E-05 2.7E-10 CE143 2.0E-05 4.0E-15 2.0E-05 2.0E-10 PR143 3.0E-05 6.0E-15 7.0E-05 8.6E-11 CE144 2.0E-05 4.0E-15 3.0E-06 1.3E-09 PR144 2.0E-05 4.0E-15 2.0E-05 2.0E-10 ALL OTH 3.0E-05 6.0E-15 1.0E-08 6.0E-07 H-3 5.8E+02 1.2E-07 1.0E-03 1.2E-04 TOTAL 6.8E-04 (1) FROM 10CFR20, APPENDIX B, TABLE 2, COLUMN 2. RN 03-038

11.2-59 Reformatted November 2011 TABLE 11.2-14

SUMMARY

OF CALCULATED LIQUID PATHWAY DOSES VIRGIL C. SUMMER NUCLEAR STATION Organ Receiving Maximum Dose Pathway Location Age Group Dose Organ Dose (mrem/yr) Total Body Dose (mrem/yr) Drinking Water Columbia Water Supply System Adult Teen Child Infant Liver Liver Liver Liver 6.1E-2 5.6E-2 1.1E-1 1.4E-1 4.6E-2 2.5E-2 2.6E-2 1.8E-2 Fish Ingestion Parr/Monticello System Adult Teen Child Liver Liver Liver 3.2E+0 3.3E+0 2.9E+0 2.4E+0 1.4E+0 5.2E-1 Shoreline Activity Parr/Monticello System Adult Teen Child Skin Skin Skin 1.5E-1 2.0E-2 4.1E-3 1.3E-1 1.7E-2 3.5E-3 Swimming Parr/Monticello System Adult Total Body 1.2E-4 1.2E-4 Boating Parr/Monticello System Adult Total Body 3.0E-4 3.0E-4 RN 03-038

11.2-60 Reformatted November 2011 TABLE 11.2-15 APPENDIX I CONFORMANCE

SUMMARY

TABLE VIRGIL C. SUMMER NUCLEAR STATION LIQUID EFFLUENTS Appendix I Criteria Virgil C. Summer Nuclear Station Type of Dose Design Objective (1) Point of Dose Evaluation Calculated Dose Point of Dose Evaluation (6) Liquid Effluents Dose to total body from all pathways 5 mrem/yr per site Location of the highest dose offsite (2) 2,54 mrem/yr (3) Parr/Monticello (4) Reservoir System Dose to any organ from all pathways 5 mrem/yr per site Same as above 3.41 mrem/yr (5) Parr/Monticello (4) Reservoir System (1) Design objectives as specified in the Commissions Appendix I Conformance Option, 40 FR 40816, September 4, 1975. (2) Evaluated at a location that is anticipated to be occupied during plant lifetime or evaluated with respect to such potential land and water usage and food pathways as could actually exist during the term of plant operation. (3) Dose to adult. (4) Fish were assumed to be exposed to average radionuclide concentrations in the Parr/Monticello Reservoir System. (5) Dose to teen liver. (6) Points given correspond to points of does evaluation under Appendix I heading. 02-01 RN 03-038 RN 03-038

 REFlfUNCi C#Nt.. DRAINS f£ACTIJl Coo..ANT SYSTEM PR(SStRE TAN( F£ACT~ Coo...ANT SYSTEM lOO' DRAINS 1161LE: C(HTAIPKNT ooT5U CONlAlI+£NT WPS_ WASTE PROCESSING SYSTEM CVCS<< CHEMICAL AND VOLUME CONTROl SYSTEM BRS - BORON RECYCLE SYSTEM STRS - BORON THERMAL REGENERATION SYSTEM SFPCS - SPENT FUEL PIT COOLING SYSTEM S - SAMPLE POINT m STRAINER IT] - FILTER REPROCESS I-----...-~----" ........ 1--1 P\\.ANT YENT SfP<S SPENT FLU PIT OCM<N SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUClEAR STATION liquid Waste Processing System Process flow Diagram Figure 11.2-1 REV. 1 Amendment 0 August 1984

SITE AREA: 2200 ACRES DAM DAM MONT ICELLO RESERVO IR AUXILIARY eUILDlfliG DA \\ CANAL

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LIQUID RELEASES: Figure 11.2-3 Amendment 0 August 1984

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A 45 ft3 Carbon B 28 ft3 Cation D 19.5 ft3 Carbon C 28 ft3 Mixed E 28 ft3 Cation PV 21 ft3 MB 7ft3 Cation (top) Typical Line Up A B D C E PV Charcoal for cleanup of oil and grease (organics), cobalt 58 and 60, initial iodines, and cesium & for service as a prefilter to protect the other demin beds. Cation demins to remove metals and transition metals. Cation demins to remove metals and transition metals. Mixed bed resin that reduces undesired positive and negative ions. Polishing Vessel SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLER STATION Duratek System Typical Lineup FSAR Figure 11.2-5 RN 03-038

11.3-1 Reformatted Per Amendment 00-01 11.3 GASEOUS WASTE SYSTEM 11.3.1 DESIGN OBJECTIVES The Gaseous Waste Processing System (GWPS) is designed to remove fission product gases from the reactor coolant in the volume control tank. The system is also designed to collect gases from the boron recycle and waste evaporators, reactor coolant drain tank, recycle holdup tanks, and reactor vessel. The system has the capacity for long term storage. Under normal operation the annual releases due to leakage and routine releases from the GWPS will be sufficiently low such that site boundary doses will be a small fraction of regulation requirements. The system is capable of operating under conditions of fuel defects in combination with equipment faults of moderate frequency. The system is designed to preclude the possibility of an internal explosion. However, the system volume is distributed so that the dose in the unlikely event of an explosion is approximately the same as the dose due to a gas decay tank rupture as analyzed in Section 15.3.5. 11.3.2 SYSTEM DESCRIPTION The GWPS consists mainly of a closed loop comprised of 2 waste gas compressors, 2 catalytic hydrogen recombiners, and gas decay tanks to accumulate the fission product gases. The routing of piping containing radioactive gases is either through shielded cubicles or behind shield slabs. NOTE: The following paragraph is being retained for historical purposes only. Components of a similar design to those used in the GWPS have been operating for several years with excellent performance. Systems constructed from carbon steel have been in service for more than 3 years and no failure due to corrosion damage has been reported. The major input to the GWPS during normal operation is taken from the gas space in the volume control tank. Table 11.3-1, based on the Reactor Coolant System activities given in Table 11.1-2, shows the maximum fission product inventory in the GWPS over the 40 year plant life. Table 11.3-2, based on the Reactor Coolant System activities given in Table 11.1-5, shows the expected fission product inventory in the GWPS over the 40 year life. 00-01

11.3-2 Reformatted Per Amendment 00-01 Figure 11.3-1 based on the reactor coolant activities given in Table 11.1-2 shows that for a given power rating with 1% fuel defects, the quantity of fission gas activity accumulated after 40 years continuous operation is about twice the activity accumulated after short term operation. Figure 11.3-2 based on realistic reactor coolant activities given in Table 11.1-5 shows that the quantity of fission gas activity accumulated after 40 years continuous operation is essentially Krypton-85 with the short lived isotopes contributing approximately 12% of the total. This is because the accumulated activity other than Krypton-85 arises from short lived isotopes which reach equilibrium after a short operating period. This accumulation of Krypton-85 is not a hazard to the plant operator because:

1.

Radiation background levels in the plant are not noticeably affected by the accumulation of Krypton-85 which is a beta emitter, for which the tanks themselves provide adequate shielding.

2.

The system activity inventory is distributed in several tanks so that the maximum permissible inventory in any single tank is actually less than that of earlier GWPS designs. Since this system permits fission gas removal from the reactor coolant during normal operation, it is expected to reduce plant activity levels caused by a leakage of reactor coolant. With operation of this system, it is possible to collect virtually all of the Krypton-85 released to the Reactor coolant and to achieve a reduction in the fission product gas inventory in the Reactor Coolant System as shown in Table 11.3-3. Table 11.3-3 is based on the Reactor Coolant System activities given in Table 11.1-1. Provisions are made to collect any residual gases stripped out of solution by the boron recycle and waste evaporators, gases from the reactor coolant drain tank, gases from the recycle holdup tanks, and gases from the reactor vessel. Process flow diagrams and piping and instrumentation diagrams are shown on Figures 11.3-3 and 11.3-4, respectively. Table 11.3-4 gives process parameters for key locations in the system, with reference to locations on Figure 11.3-3 and based on the Reactor Coolant System activities given in Table 11.1-5. The process parameters are derived assuming the system to operate as described in Section 11.3.4. The stripping efficiency used in the analysis is 0.4, and the volume control tank purge rate is 0.7 scfm of hydrogen. RN 02-025

11.3-3 Reformatted Per Amendment 00-01 11.3.3 SYSTEM DESIGN 11.3.3.1 Component Design Gaseous waste processing equipment parameters are given in Table 11.3-5. Component safety classes and the corresponding code and code class are shown in Table 3.2-1. All materials used for pressure retaining components are allowed by Section III of the ASME Code, and no malleable wrought or cast iron or plastic pipe is used. Quality assurance requirements of Westinghouse Administrative Specifications for the Procurement of Nuclear Steam Supply System Components, Revision 5, March 1975 are applied to all NNS components within Westinghouse scope. Hence, all components within Westinghouse scope meet the design guidance as outlined in Branch Technical Position ETSB 11-1. 11.3.3.1.1 Waste Gas Compressor Packages Two (2) waste gas compressor packages are provided to circulate gases around the system loop. One (1) unit is normally used with the other on a standby basis. The units are water-sealed centrifugal displacement machines which are skid-mounted in a self-contained package. Construction is primarily of carbon steel. Mechanical seals are provided to minimize the out-leakage of seal water. 11.3.3.1.2 Catalytic Hydrogen Recombiner Packages Two (2) catalytic hydrogen recombiners are provided. One (1) of the 2 recombiners is normally used to remove hydrogen from the hydrogen-nitrogen-fission gas mixtures by oxidation to water vapor, which is removed by condensation. The other recombiner is available on a standby basis. Both units are self-contained and designed for continuous operation. The recombiner is located in the system where the hydrogen concentration and pressure are optimum with respect to hydrogen removal. 11.3.3.1.3 Waste Gas Decay Tanks Waste gas decay tanks are provided as described in Table 11.3-5. The tanks are of vertical-cylindrical type and are constructed of carbon steel. There are 8 waste gas decay tanks, 6 are used during normal operation while the remaining 2 are used for shutdown and startup.

11.3-4 Reformatted Per Amendment 00-01 11.3.3.1.4 Valves Each valve in the recombiner system is designed to meet the temperature, pressure, and code requirements for the specific application in which it is used. The recombiner circuits contain manual valves provided with a metal diaphragm to prevent stem leakage and control valves with gaseous leakoffs returned to the GWPS. Other parts of the GWPS use control valves with bellows seal. Relief valves have soft seats and operate at pressures which are normally less than 2/3 of the relief valve set pressure. The relief valves of the major components discharge to the shutdown tanks. This permits decay and controlled disposal of all discharges less than about 3000 scf. It also provides a means for containing and detecting seat leakage across the relief valves. 11.3.3.2 Instrumentation and Control Design The main system instrumentation is described in Table 11.3-6 and shown on the piping and instrumentation diagrams, Figure 11.3-4. The instrumentation readout is located mainly on the Waste Processing System (WPS) panel in the Auxiliary Building. Some instruments are read near the equipment location. All alarms are shown separately on the WPS panel and further relayed to 1 common WPS annunciator on the main control board. Where suitable, instrument lines are provided with diaphragm seals to prevent fission gas outleakage through the instrument. Figure 11.3-5 shows the location of the instruments on the compressor package. The compressors are interlocked with the seal water inventory in the moisture separators and trip off on either a high or a low moisture separator level. During normal operation the proper seal water inventory is maintained automatically. Figure 11.3-6 indicates the location of the instruments on the recombiner installation. The catalytic hydrogen recombiner packages are designed for automatic operation with a minimum of operator attention. Each package includes 4 online gas analyzers, 1 each to measure hydrogen in, oxygen in, hydrogen out, and oxygen out, which are the primary means of recombiner control. A multipoint temperature recorder monitors temperatures at several locations in the packages. Process gas flowrate is measured by an orifice located upstream of the recombiner preheater. Local pressure gauges indicate pressure at the recombiner inlet and the oxygen supply pressure.

11.3-5 Reformatted Per Amendment 00-01 The following controls and alarms are incorporated to maintain the gas composition outside the range of flammable and explosive mixtures:

1.

A high flow alarm actuates when the volume control tank purge flow exceeds a predetermined value. This high flow alarm is set below the flow which corresponds to the maximum inlet concentration (6% hydrogen by volume) the recombiner can process in one pass.

2.

If the hydrogen concentration in the recombiner feed exceeds 4% by volume, a high hydrogen and high-high hydrogen/oxygen shutdown alarm sounds, the oxygen feed is terminated through TCV01114, and the volume control tank hydrogen purge flow is terminated. The control and alarm setpoints were lowered to a identical setpoint of 4% by volume to limit the possible accumulation of hydrogen in the system to 4% by volume for compliance with Technical Specifications.

3.

If the oxygen concentration in the recombiner feed reaches 2% by volume, a high oxygen and high-high oxygen/shutdown alarm sounds, oxygen feed flow is limited through HCV01118, and the oxygen feed is terminated through TCV01114. The control and alarm setpoints were lowered to a identical setpoint of 2% by volume which is below the flammable limit for hydrogen-oxygen mixtures for Technical Specifications compliance.

4.

If hydrogen in the recombiner discharge exceeds 0.15% by volume, an alarm sounds. This alarm warns of high hydrogen feed, possible hydrogen-oxygen catalytic reactor malfunction, or loss of oxygen feed.

5.

If oxygen in the recombiner discharge exceeds 60 ppm an alarm sounds and oxygen feed is terminated. This control prevents any accumulation of oxygen in the system in case of catalytic reactor malfunction.

6.

On low flow through the recombiner, oxygen feed is terminated. This control prevents an accumulation of oxygen following system malfunction.

7.

High discharge temperature from the cooler-condenser (downstream from the catalytic reactor) will terminate oxygen feed. This protects against loss of cooling water flow in the cooler-condenser.

8.

High temperature indication by any 1 of 6 thermocouples in the catalyst bed will limit oxygen feed so that no further increase is possible.

9.

High temperature indication at the recombiner catalytic reactor discharge will terminate oxygen feed to the recombiner. 98-01 02-01

11.3-6 Reformatted Per Amendment 00-01 11.3.4 OPERATING PROCEDURES 11.3.4.1 General Description The GWPS is a closed loop comprised of 2 waste gas compressors, 2 catalytic hydrogen recombiners, 6 gas decay tanks for normal power service, 2 gas decay tanks for service at shutdown and startup, 1 gas decay tank drain pump, 1 waste gas drain filter and 4 gas traps. All of the equipment is located in the Auxiliary Building. 11.3.4.2 Startup Operation Startup commences with the system flushed free of air by purging with nitrogen which is discharged to the atmosphere. One (1) compressor, 1 recombiner, and 1 shutdown decay tank are in service. The reactor is at cold shutdown and the volume control tank contains nitrogen in the gas space. Reactor coolant contains neither hydrogen nor fission gases, but it may be saturated with air. When the reactor startup procedure requires that a hydrogen blanket be established in the volume control tank gas space, fresh hydrogen is charged into the tank. The hydrogen-nitrogen mixture vented from the tank enters the circulating nitrogen stream at the compressor suction. Since the pressure downstream remains constant by use of a pressure regulating valve, nitrogen added to the loop will accumulate in the shutdown decay tank causing the tank pressure to rise. Initially, the volume control tank vent gas will be very lean in hydrogen, and almost all the influent gas will accumulate in the tank. As the operation continues, however, the vent gas hydrogen content will gradually increase until it is almost totally hydrogen at the point when all of the nitrogen has been removed from the coolant. At that time, hydrogen gas is entering the volume control tank at 0.7 scfm and mixing with the 40 scfm circulating nitrogen stream to give a 1.8 volume % mixture of hydrogen in nitrogen at the recombiner inlet. Approximately 0.35 scfm of oxygen is added in the recombiner and reacted with the hydrogen to yield a discharge stream of 0.1 volume % hydrogen in nitrogen after water vapor is condensed. When the reactor coolant nitrogen concentration is within operating specifications, the shutdown tank is isolated and flow is routed to 1 of the decay tanks provided for normal power service. Gas accumulated in the shutdown tank will be retained for use during operations to strip hydrogen from the reactor coolant when the plant is shut down. 02-01

11.3-7 Reformatted Per Amendment 00-01 11.3.4.3 Normal Operations During normal power operation, nitrogen gas is continuously circulated around the loop by 1 of the 2 compressors. Fresh hydrogen gas is charged to the volume control tank where it is mixed with fission gases which are stripped from the reactor coolant into the tank gas space. The contaminated hydrogen gas is then vented from the tank into the circulating nitrogen stream to transport the fission gases into the GWPS. The resulting mixture of nitrogen-hydrogen-fission gas is pumped by the compressor to the recombiner where enough oxygen is added to reduce the hydrogen to a low residual level by oxidation to water vapor on a catalytic surface. After the water vapor is removed, the resulting gas stream is circulated to the waste gas decay tanks and back to the compressor suction to complete the loop circuit. Each waste gas decay tank is capable of being isolated and the number of tanks valved into operation at any time is restricted to diminish the amount of radioactive gases which could be released as a consequence of any single failure, such as the rupture of any single tank or connected piping. By alternating use of these tanks, the accumulated activity is distributed among the tanks. 11.3.4.4 Shutdown When the hydrogen contained in the reactor coolant must be removed in preparation for a cold shutdown, the normal gas decay tanks are valved out of service and 1 of the 2 shutdown tanks is placed in service. Additional nitrogen may have to be added to raise the shutdown tank to an acceptable pressure for this operation. In addition, the flow of hydrogen to the volume control tank is stopped, and the tank pressure is maintained with nitrogen. The volume control tank level may be raised and lowered to aid in hydrogen removal. Once the hydrogen concentration has been lowered to acceptable levels the volume control tank purge to the waste gas system may be secured. 11.3.5 PERFORMANCE TESTS NOTE The following paragraph is being retained for historical purposes only. Compressor and recombiner packages are subjected to helium leak test after assembly. Initial performance tests are performed to verify the operability of the components, instrumentation, and control equipment. During reactor operation the system is in use and hence is under continuous surveillance. The system design permits the use of industry standard leak detection methods for leak testing and subsequent elimination of leaks. 02-01 02-01 RN 03-023

11.3-8 Reformatted Per Amendment 00-01 11.3.6 ESTIMATED RELEASES Gaseous releases from the Virgil C. Summer Nuclear Station were calculated using the PWR-GALE Code[1] as specified in Regulatory Guide 1.112 (see Appendix 3A). The input parameters used to calculate gaseous releases are listed in Table 11.3-7 and are discussed in more detail in Sections 11.3.6.2 through 11.3.6.4. Calculated releases using the parameters listed in Table 11.3-7 are presented in Table 11.3-8. A comparison of effluent concentrations with 10 CFR 20, Appendix B, Table 2, Column 1 is presented in Section 11.3.8. 11.3.6.1 Gaseous Waste Processing System The GWPS collects and stores gases stripped from the primary coolant in a continuously recirculating loop which includes pressurized storage tanks. Release calculations for the GWPS are based upon the options allowed by the PWR-GALE Code[1] for such a system (continuous purging of Volume Control Tank, 90 days decay time in storage tanks, 0 day fill time). Using these options and the input parameters given in Table 11.3-7, the GWPS releases are calculated to be 214 Ci/yr of noble gases, 4 x 10-4 Ci/yr of airborne particulates, and 7 Ci/yr of Carbon-14. The isotopic distribution of these releases is given in Table 11.3-8 11.3.6.2 Reactor Building Purge Radioactive gases are released inside the Reactor Building when primary system components are opened or if leakage from the primary system occurs. The gaseous activity inside the Reactor Building may be purged up to 1000 hours per year in Modes 1-4 by the 6 inch low volume purge system. The low volume purge rate is 600 cfm. Activity is also released periodically when the 36 inch Reactor Building Purge System is used during Modes 5-6. The Reactor Building Charcoal Cleanup System is operated intermittently to reduce airborne iodine concentrations prior to Reactor Building access or purge system operation. The Reactor Building 36 inch and 6 inch purge flow is exhausted to the atmosphere through HEPA filters and charcoal adsorbers. Release calculations are based upon the PWR-GALE Code[1] parameters for leakage rate (1%/day of primary coolant noble gas inventory, 0.001%/day of primary coolant iodine inventory), recirculation cleanup time (16 hours), mixing efficiency (70%), decontamination factors (100 for HEPA filters, 10 for charcoal adsorbers) and number of high volume purges (36 inch during cold shutdown) per year (4) and a conservatively assumed continuous low volume purge (6 inch) rate of 1000 cfm. Using these parameters and the input parameters given in Table 11.3-7, the Reactor Building purge releases are calculated to be 2638 Ci/yr of noble gases, 2.5 x 10-2 Ci/yr of iodine, 1.9 x 10-3 Ci/yr of airborne particulates, and 1 Ci/yr of Carbon-14. The isotopic distribution of these releases is given in Table 11.3-8. RN 02-028 RN 02-028 RN 02-028 RN 02-028

11.3-9 Reformatted Per Amendment 00-01 11.3.6.3 Auxiliary Building Ventilation The Auxiliary Building Charcoal Exhaust System continuously exhausts air drawn from Auxiliary Building areas with moderate potential for radioactive contamination (demineralizers, storage tanks, gas decay tanks, evaporators, pump rooms, etc.). The supply and exhaust ducts are arranged so that air flow is always in the direction of progressively greater potential contamination. Exhaust air from these areas is drawn through the roughing/HEPA/charcoal filter plenums continuously and is ducted to the main exhaust fans and the main plant vent. There is no bypass around this filter plenum. The release calculations are based upon the assumption that reactor coolant leakage in the Auxiliary Building occurs primarily in the areas exhausted by the charcoal exhaust system. PWR-GALE Code [1] parameters for Auxiliary Building leakage (160 lbs/day), iodine partition factor (0.0075), and decontamination factors (100 for HEPA filters, 10 for charcoal adsorbers) have been used in the calculations. Using these parameters and the input parameters given in Table 11.3-7, the Auxiliary Building ventilation release is calculated to be 128 Ci/yr of noble gases, 1.1 x 10-2 Ci/yr of iodine, and 1.6 x 10-3 Ci/yr of airborne particulates. The isotopic distribution of this release is given in Table 11.3-8. 11.3.6.4 Secondary System 11.3.6.4.1 Turbine Building Vents Turbine Building steam leakage may release radioactive gas to the Turbine Building atmosphere if primary to secondary leakage occurs. Turbine Building Ventilation System exhausts are not treated prior to release. Release calculations were based on the PWR-GALE Code[1] parameters for steam leakage (1700 lbs/hr), primary to secondary leakage (100 lbs/day), and fraction of iodine that remains airborne (1). Using these parameters and the input parameters given in Table 11.3-7, the Turbine Building vent release is calculated to be 2.4 x 10-3 Ci/yr of iodine. The isotopic distribution of this release is given in Table 11.3-8. 11.3.6.4.2 Condenser Air Removal System Offgas from the Condenser Air Removal System may contain radioactive gases, if primary to secondary leakage occurs. When condenser offgas contains any significant amount of radioactivity, it is exhausted through HEPA filters and charcoal adsorbers in the Auxiliary Building Charcoal Exhaust System from particulate and iodine removal. Release calculations are based upon taking credit for the charcoal adsorbers and the PWR-GALE Code[1] parameters for primary to secondary leakage (100 lbs/day), steam generator partition factors (0.01 for iodine and 0.001 for nonvolatiles), and Main Condenser/Condenser Air Removal System partition factors (0.15 for volatile iodine species and zero for nonvolatile species). Using these parameters and the input parameters given in Table 11.3-7, the Condenser Air Removal System release is calculated to be 81 Ci/year of noble gases and 6.9 x 10-3 Ci/yr of iodine. The isotopic distribution of this release is given in Table 11.3-8. RN 02-028

11.3-10 Reformatted Per Amendment 00-01 11.3.6.4.3 Steam Generator Blowdown The Steam Generator Blowdown Processing System provides for cooling the blowdown in heat exchangers to prevent flashing. Consequently, no gaseous release is expected to result from steam generator blowdown. 11.3.6.5 Release Criteria It is the intent of the Applicant to operate this system by periodically discharging gases stored by the GWPS. This method of operation minimizes disposal of the accumulated inventory at the end of plant life and reduces plant personnel exposure. Planned discharges during periods of favorable meteorology are made after a sample of decayed gaseous effluent is analyzed and are continuously monitored during release. Radiation monitor (RM-A10) automatically terminates the discharge upon detection of high activity by closing the appropriate tank outlet valve. This method of operation provides both operational flexibility and assures that the release of radioactive material in gaseous effluents is within the limits of 10 CFR 20, Appendix B, Table 2, Column 1 and the limits of Appendix I to 10 CFR 50. 11.3.7 RELEASE POINTS Release points for potentially radioactive gaseous wastes are shown schematically by Figure 11.3-7. Figure 11.3-8 shows the physical locations of these, and other nonradioactive exhausts. Table 11.3-9 presents data for the numbered vents shown by Figure 11.3-8. The data include base and exit elevations of the stacks, cross Section dimensions, volumetric flow rate, exit velocity, and comments. Table 11.3-9a compares exhaust system equipment to Branch Technical Position ETSB 11-2. 11.3.8 DILUTION FACTORS Dilution factors (/Q's) utilized in evaluating the releases of gaseous effluents were calculated according to the methods set forth in Regulatory Guide 1.111, based on 1 year of onsite meteorological data. A detailed discussion of the applicable methodology appears in Section 2.3.5.2; the results of the calculation of annual average (/Q's) values are listed in Table 2.3-133. Examination of Table 2.3-133 reveals that the highest concentration of gaseous effluents at the exclusion zone boundary is expected to occur in the southeastern sector, where relative concentration of 5.3 x 10-6 sec/m3 was calculated. RN 02-028

11.3-11 Reformatted Per Amendment 00-01 Expected annual gaseous release rates presented in Table 11.3-8 were used in conjunction with a ( /Q's) value of 5.3 x 10-6 sec/m3 to estimate maximum expected radioisotope concentrations, in air outside the restricted area. The release rates, the expected concentrations, and the effluent concentration limits from 10 CFR 20, Appendix B, Table 2, are listed in Table 11.3-10. As prescribed in 10 CFR 20, these concentrations are those expected as an average at the exclusion zone boundary over a 1-year period. It can be seen that the expected concentration level of each isotope is well below the individual limit specified. In addition to the limits for each isotope, the requirements of 10 CFR 20, Appendix B state that, for a mixture of radionuclides, the following relationship must hold:

=

N 1 i i i 1 ECL C Where: Ci = concentration of radionuclide i. ECLi = effluent concentration limit of radionuclide i from 10 CFR 20, Appendix B, Table 2, Column 1. N = number of radionuclides in the mixture. The sum of the ratios of expected radionuclide concentrations to their effluent concentration limits for the mixture defined by the second column of Table 11.3-10 is 3.5 x 10-3, which is less than unity, as required. 11.3.9 ESTIMATED DOSES (2) Potential pathways of exposure (1) of man to radioactive materials in gaseous effluents from the Virgil C. Summer Nuclear Station are identified and discussed in Section 11.6.2. Doses to individuals in the environs of the plant from each of the potentially significant pathways were calculated; methodology for the results of the calculations are discussed in the following paragraphs. (1) The term "exposure" as used in this Section refers only to the disposition of radioactive materials in the environment in such a way that persons could receive a dose from them. (2) Current values are being maintained in the ODCM. RN 02-028 RN 02-028 RN 02-028

11.3-12 Reformatted Per Amendment 00-01 Dilution factors and relative deposition were calculated according to the methods of Regulatory Guide 1.111, as discussed in Section 2.3. All results presented in these Sections were obtained using the calculational techniques prescribed in Regulatory Guide 1.109. Except where noted in discussion of doses for specific pathways, all usage and consumption values, transport times, bioaccumulation factors, dose conversion factors, and other constants utilized were those suggested in Regulatory Guide 1.109. Maximum doses to individuals were calculated for cloud submersion, ground plane contamination, inhalation, and vegetable, milk, and meat ingestion pathways. Assumptions, including point of exposure, are described for each pathway in the following paragraphs; the calculated gaseous pathway doses are summarized in Table 11.3-11. All estimates were based on the predicted gaseous releases given in Table 11.3-8. Each dose was calculated at the location of the highest dose offsite at which the pathway could be assumed to exist. Exposure to an individual from submersion in a cloud containing radioactive effluents was evaluated at the nearest residence, located 1.1 miles to the east-southeast of the plant. The total body dose was calculated to be 6.4 x 10-2 mrem/yr, while the skin dose was 1.8 x 10-1 mrem/yr. External irradiation from activity deposited on the ground surfaces was also evaluated at the nearest residence. These analyses indicate that a dose of 5.0 x 10-3 mrem/yr to the skin and 4.3 x 10-3 mrem/yr to the total body can be expected from this pathway. In addition, the nearest residence is the location for estimating the maximum individual dose to be received from the air inhalation pathway. The maximum dose to an organ of an individual at this location inhaling radioiodine and radioparticulates in the plant effluent was calculated to be 6.9 x 10-2 mrem/yr to an adults thyroid. The predicted dose to an individual obtaining 100% of his vegetable consumption from a garden adjacent to the nearest residence was also determined. Maximum calculated exposure from this pathway was 7.7 x 10-1 mrem/yr to a childs thyroid. Maximum total body dose was 7.1 x 10-1 mrem/yr to a child. Predicted doses from ingestion of milk from animals grazing year-round on land contaminated by radioparticulates deposited from the effluent plume were evaluated at the location of the nearest cow, in the west-southwest sector at 1.5 miles from the plant. Although the cow at this location is not currently being milked, the suitability of the location for raising dairy cattle and the increasing popularity of dairying in the region were considered sufficient reason to assume that the pathway could reasonably be expected to exist at this location during the life of the plant. The maximum organ dose from ingestion of milk from a cow grazing year-round at this location was 1.1 x 100 mrem/yr to an infants thyroid. The infant is also expected to receive the maximum total body dose of 3.2 x 10-1 mrem/yr.

11.3-13 Reformatted Per Amendment 00-01 Exposure from consumption of meat was evaluated at the same location as that for cow milk. The maximum organ dose to an individual from ingestion of meat from a cow grazing year-round at this location was 1.4 x 10-1 mrem/yr to the bones of an adult. The maximum total body dose from the meat ingestion pathway was 4.6 x 10-2 mrem/yr to a child. Maximum individual doses calculated as described above were used to evaluate the status of conformance of predicted gaseous effluents from the Virgil C. Summer Nuclear Station with the requirements of Appendix I to 10 CFR 50. The assumptions and results of this evaluation are summarized in Table 11.3-12. Beta and gamma doses in air were calculated according to the methods of Regulatory Guide 1.109. It will be noted that the calculated doses indicate the plant design conforms to the "as low as reasonably achievable" criteria established in Appendix I. Conformance with 10 CFR 50, Appendix I was demonstrated using meteorological data observed during 1975 and preoperational land use census data. During plant operation, conformance with Appendix I will be demonstrated in the USNRC Regulatory Guide 1.21 Annual Radioactive Effluent Release Report. Meteorological data used in preparation of the annual effluent report may consist of meteorological data averaged over multiple years to provide a better estimate of dispersion values. The maximum exposed individual location used for gaseous release dose calculation will be based on current census data 11.3.10 REFERENCES

1.

U.S. Nuclear Regulatory Commission, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors," NUREG-0017, April, 1976. 99-01

11.3-14 Reformatted Per Amendment 02-01 TABLE 11.3-1 DESIGN BASIS ACCUMULATED RADIOACTIVITY PER UNIT IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATION Activity (Curies) Following Plant Shutdown Isotope Zero Decay 30 Days 50 Days Kr-85 53,000 52,700 52,500 All other noble gases Kr-83m 4.3 ~ 0 ~ 0 Kr-85m 47.0 ~ 0 ~ 0 Kr-87 4.6 ~ 0 ~ 0 Kr-88 45.0 ~ 0 ~ 0 Xe-131m 530 91 28.1 Xe-133 56,000 1090 78.8 Xe-133m 2900 ~ 0 ~ 0 Xe-135 500 ~ 0 ~ 0 Xe-135m 2.1 ~ 0 ~ 0 Xe-138 0.13 ~ 0 ~ 0 The table is based on 40 years continuous operation with 1 % fuel defect and 60 gpm letdown. Power assumed to be 2958 MWt. The data are based on a volume control tank purge rate of 0.7 scfm, a 40 % stripping efficiency and the stripping fractions listed in Table 11.1-1. 02-01 RN 02-025 RN 02-025

11.3-15 Reformatted Per Amendment 02-01 TABLE 11.3-2 EXPECTED ACCUMULATED RADIOACTIVITY PER UNIT IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATION Activity (Curies) Following Plant Shutdown Isotope Zero Decay 30 Days 50 Days Kr-85 6400 6370 6340 All other noble gases Kr-85m 5.0 ~ 0 ~ 0 Kr-87 0.88 ~ 0 ~ 0 Kr-88 5.1 ~ 0 ~ 0 Xe-131m 200 34.3 10.6 Xe-133 550 10.7 0.8 Xe-133m 11 ~ 0 ~ 0 Xe-135 47 ~ 0 ~ 0 Xe-135m 0.047 ~ 0 ~ 0 Xe-138 0.037 ~ 0 ~ 0 Inventories are based on reactor coolant concentrations given in Table 11.1-5. The table is based on 40 years continous operation with 60 gpm letdown. Power assumed to be 2958 MWt. The data are based on a volume control tank purge rate of 0.7 scfm, a 40 % stripping efficiency and the stripping fractions listed in Table 11.1-6. 02-01 RN 02-025 RN 02-025 02-01

11.3-16 Reformatted Per Amendment 02-01 TABLE 11.3-3 REDUCTION IN REACTOR COOLANT SYSTEM GASEOUS FISSION PRODUCTS RESULTING FROM NORMAL OPERATION OF THE GASEOUS WASTE PROCESSING SYSTEM (1) Reactor Coolant Gaseous Fission Product Activities - µc/gm Isotope GWPS Operating (2) GWPS Not Operating Kr-83m 0.42 0.43 Kr-85 0.052 7.6 Kr-85m 1.7 1.8 Kr-87 1.1 1.1 Kr-88 3.1 3.2 Kr-89 0.089 0.089 Xe-131m 0.23 2.3 Xe-133 58 290 Xe-133m 7.2 19 Xe-135 6.9 8.6 Xe-135m 0.52 0.52 Xe-137 0.18 0.18 Xe-138 0.64 0.64 (1) Based on operating with cladding defects in fuel generating 1 % of the rated core thermal power (2958 MWt) and a purification letdown rate of 60 gpm. (2) Volume control tank purge rate is 0.7 scfm. Stripping efficiency is 40 %. 02-01 02-01 RN 02-025

11.3-17 Reformatted Per Amendment 02-01 TABLE 11.3-4 PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM* BASIS: Power Level - 2900 MWt No. of Units - 1 Gas Decay Tanks (Note 4) - 8 Operating Interval - 1 day Stripping Efficiency - 0.4 ITEM DESCRIPTION TEMP PRESS FLOW N2 H2 ISOTOPIC CONCENTRATION, C/cc (NOTE 1) GAS STREAMS F PSIG SCFM KR 85 (NOTE 3) KR85M KR87 KR88 XE-133 XE-133M XE-135 1. VOLUME CONTROL TANK PURGE 130 15 0.7 0 100 3.07 X 10-2 1.81 X 10-1 6.51 X 10-2 1.67 X 10-1 1.95 X 101 3.81 X 10-1 8.74 X 10-1 2. GAS DECAY TANK DISCH. TO COMP. AMB 0.5 40 99.9 0.1 1.31 X 101 9.95 X 10-2 5.61 X 10-3 1.45 X 10-1 3.79 X 101 1.40 5.50 X 10-1 3. COMPRESSOR SUCTION AMB 0.5 40.7 98.3 1.7 1.29 X 101 1.01 X 10-1 6.63 X 10-3 1.45 X 10-1 3.76 X 101 1.39 5.56 X 10-1 4. COMP. DISCH. TO RECOMBINER 140 45 40.7 98.3 1.7 1.29 X 101 1.01 X 10-1 6.63 X 10-3 1.45 X 10-1 3.76 X 101 1.39 5.56 X 10-1 5. RECOMBINER DISCH. TO GAS DECAY TANKS 140 30 40 99.9 0.1 1.31 X 101 1.03 X 10-1 6.75 X 10-3 1.48 X 10-1 3.82 X 101 1.41 5.66 X 10-1 6. MISC. VENTS-EVAPS. RCDT. RECYCLE HOLDUP TANK EDUCTOR 140 0.5 NEG 0 100 0 0 0 0 0 0 0 7. RECOMBINER OXYGEN SUPPLY AMB 50 0.35 0 0 0 0 0 0 0 0 0 8. RECOMBINER CALIBRATING GAS AMB 15 0.004 100 4 0 0 0 0 0 0 0 9. RECOMBINER CALIBRATING GAS AMB ATM 0.004 100 4 0 0 0 0 0 0 0 98-01 02-01 02-01 02-01

11.3-18 Reformatted Per Amendment 02-01 TABLE 11.3-4 (Continued) PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM* ITEM DESCRIPTION TEMP PRESS FLOW N2 H2 ISOTOPIC CONCENTRATION, C/cc (NOTE 1) GAS STREAMS F PSIG SCFM KR 85 (NOTE 3) KR85M KR87 KR88 XE-133 XE-133M XE-135 10. WASTE GAS SYSTEM NITROGEN SUPPLY AMB 100 0 100 0 0 0 0 0 0 0 0 11. NSSS NITROGEN SUPPLY AMB 100 0 100 0 0 0 0 0 0 0 0 12. NITROGEN RELIEF TO PLANT VENT AMB 100 0 100 0 0 0 0 0 0 0 0 13. NSS HYDROGEN SUPPLY AMB 100 0.7 0 100 0 0 0 0 0 0 0 14. VOLUME CONTROL TANK HYDROGEN AMB 100 0.7 0 100 0 0 0 0 0 0 0 15. HYDROGEN RELIEF TO PLANT VENT AMB 100 0 0 100 0 0 0 0 0 0 0 16. WASTE GAS DISCH. TO PLANT VENT AMB ATM 0 100 0 1.31 x 101 0 0 0 0 0 0 17. RECYCLE GAS TO VOLUME CONTROL TANK AMB 100 0 100 0 0 0 0 0 0 0 0 18. PRESSURIZER RELIEF TANK VENT AND RETURN 120 3 0 100 0 0 0 0 0 0 0 0 19. SHUTDOWN TANK RELIEF AMB ATM 0 100 0 0 0 0 0 0 0 0 02-01 02-01 02-01 02-01

11.3-19 Reformatted Per Amendment 02-01 TABLE 11.3-4 (Continued) PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM* ITEM DESCRIPTION TEMP PRESS FLOW ISOTOPIC CONCENTRATION C/cc (NOTE 2) LIQUID STREAMS F PSIG GPD KR85 (NOTE 3) KR85M KR87 KR88 XE-133 XE-133M XE-135 1. WASTE GAS COMPRESSOR DRAIN 140 45 0 3.43 2.69 X 10-2 1.77 X 10-3 3.87 X 10-2 8.27 3.05 X 10-1 1.22 X 10-1 2. RECOMBINER DRAIN 140 30 6 2.61 2.05 X 10-2 1.35 X 10-3 2.95 X 10-2 6.30 2.32 X 10-1 9.32 X 10-2 3. GAS DECAY TANK DRAINS AMB 40 36 9.18 X 10-1 6.98 X 10-3 3.94 X 10-4 1.02 X 10-2 2.20 8.13 X 10-2 3.19 X 10-2 4. SYSTEM DRAINS TO VOL CONTROL TANK 140 30-45 42 1.16 8.91 X 10-3 5.30 X 10-4 1.29 X 10-2 2.78 1.02 X 10-1 4.07 X 10-2 5. RECOMBINER REACTOR MAKEUP WATER AMB 0 0 0 0 0 0 0 0 6. COMPRESSOR MAKEUP WATER AMB 36 0 0 0 0 0 0 0 02-01 02-01 02-01

11.3-20 Reformatted Per Amendment 02-01 TABLE 11.3-4 (Continued) PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM* ITEM COMPONENT TEMP PRESS VOL N2 H2 COMPONENT INVENTORY, CURIES F PSIG FT3 KR 85 (NOTE 3) KR85M KR87 KR88 XE-133 XE-133M XE-135 A COMPRESSOR 140 45 4 98.3 1.7 6.02 4.58 X 10-2 2.58 X 10-3 6.67 X 10-2 1.74 X 101 6.45 X 10-1 2.53 X 10-1 B. RECOMBINER 140 30 4 99.9 0.1 4.51 3.42 X 10-2 1.93 X 10-3 4.99 X 10-1 1.3 X 101 4.82 X 10-1 1.89 X 10-1 C. GAS DECAY TANK AMB 1.0 600 99.9 0.1 1.04 X 103 1.80 1.02 X 10-1 2.62 6.86 X 102 2.54 X 101 9.96 TOTAL SYSTEM 6.28 X 103 1.88 1.06 X 10-1 2.74 3.18 X 103 2.65 X 101 1.04 X 101

  • based on stripping fractions from Table 11.1-6 and reactor coolant activities from Table 11.1-5.

NOTES: 1. Concentration in c per cc of gas at atmospheric pressure and 140F. 2. Concentrations in c per cc liquid at room temperature. 3. Kr - 85 concentrations are maximum values, but do not occur simultaneously with other isotope maximum concentrations. 4. Includes two shutdown tanks. 5. AMB - Ambient 6. NEG - Negligible 7. ATM - Atmospheric 02-01 02-01 02-01

TABLE 11.3-5 GASEOUS WASTE PROCESSING SYSTEM COMPONENT DATA Waste Gas Compressor Packages Number 2 Design pressure, psig 150 Design temperature, °F 180 Normal operating temperature, °F 70-140 Normal operating pressure, psig Suction 0.5-2.0 Discharge 0-110 Design flowrate (N2 at 60°F, 0 psig), scfm 40 Waste Gas Decay Tanks Number 8 Design pressure, psig 150 Design temperature, °F 180 Volume (Each). ft3 600 Normal operating pressure, psig 0-110 Normal operating temperature, °F 50-140 Material of construction Carbon Steel Catalytic Hydrogen Recombiner Packages Number 2 Design inlet pressure, psig 110 Design inlet temperature, °F 140 Design flowrate, scfm 40 Design hydrogen recombiner rate, scfm 2.4 Design discharge pressure, psig 15 Design discharge temperature, °F 140 Material of construction Stainless Steel 02-01 11.3-21 Reformatted Per Amendment 02-01

TABLE 11.3-6 GASEOUS WASTE PROCESSING SYSTEM INSTRUMENTATION DESIGN PARAMETERS A - Alarm C - Control F - Flow I - Indication L - Level P - Pressure Q - Water Integrator R - Radiation T - Temperature Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range Alarm Setpoint Control Setpoint Location of Readout FLOW INSTRUMENTATION FIA - 1094 Volume Control Tank Discharge Flow 150 250 0.3-1.2 scfm 1.2 scfm WPS panel QAI - 1091 Gas Decay Tank Water Flush 150 180 0-6000 gal 3000-6000 gal (adjustable) Local PRESSURE INSTRUMENTATION PI - 1031 Moisture Separator 150 180 0-160 psig Local PI - 1033 Moisture Separator 150 180 0-160 psig Local PIA - 1036 Gas Decay Tank Number 1 150 180 0-150 psig 0-30 psig 100 psig 20 psig PIA - 1037 Gas Decay Tank Number 2 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1038 Gas Decay Tank Number 3 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1039 Gas Decay Tank Number 4 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel RN 07-010 02-01 11.3-22 Reformatted Per Amendment 02-01

TABLE 11.3-6 (Continued) GASEOUS WASTE PROCESSING SYSTEM INSTRUMENTATION DESIGN PARAMETERS Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range Alarm Setpoint Control Setpoint Location of Readout PRESSURE INSTRUMENTATION (Cont) PIA - 1052 Gas Decay Tank Number 5 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1053 Gas Decay Tank Number 6 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1054 Gas Decay Tank Number 7 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1055 Gas Decay Tank Number 8 150 180 0-150 psig 0-30 psig 100 psig 20 psig WPS panel PIA - 1065 Hydrogen Supply Header 150 180 0-150 psig 90 psig WPS panel PIA - 1066 Nitrogen Supply Header 150 180 0-150 psig 90 psig WPS panel PICA - 1092 Compressor Suction Header 150 180 2 psi vac.- 2 psig 0.5 psi 0.5 psi vac. WPS panel PI - 1093 Gas Decay Tank Makeup Water 150 180 0-150 psig 2 psi Local PA - 1094 Volume Control Tank Discharge Pressure 150 250 0-20 psig Local 02-01 02-01 11.3-23 Reformatted Per Amendment 02-01

11.3-24 Reformatted Per Amendment 02-01 TABLE 11.3-6 (Continued) GASEOUS WASTE PROCESSING SYSTEM INSTRUMENTATION DESIGN PARAMETERS Channel Number Location of Primary Sensor Design Pressure (psig) Design Temperature (°F) Range Alarm Setpoint Control Setpoint Location of Readout LEVEL INSTRUMENTATION LICA - 1030 Compressor Moisture Separator 150 180 0-30 Inches H20 15 inches H20 15 to -10 inches 8 to -5 inch -1 inches H20 WPS panel and local LICA - 1032 Compressor Moisture Separator 150 180 0-30 inches H20 15 inches H20 15 to -10 inches -8 to -5 inches -1 inches H20 WPS panel and local RADIATION INSTRUMENTATION RM - A10 Gas Discharge Monitor 15 100 Adjustable WPS panel and control room 02-01

11.3-25 Reformatted Per Amendment 02-01 TABLE 11.3-7 PWR-GALE CODE INPUT PARAMETERS USED IN CALCULATING RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS Reactor Power Level, MWt 2914 Holdup Time for Xenon Stripped from Primary Coolant, days 90 Holdup Time for Krypton Stripped from Primary Coolant, days 90 Reactor Building Free Volume, million ft3 1.84 Flow Rate through Reactor Building Charcoal Cleanup System, thousand cfm 24 Continuous Reactor Building Ventilation Rate, cfm 1000 (1) Primary System Mass of Primary Coolant, thousand lbs 404 Letdown Rate, gpm 60 Letdown Cation Demineralizer Flow Rate, gpm 6 Secondary System Steam Flow Rate, million lbs/hr 12.2 Mass of Steam/Steam Generator, thousand lbs 8.56 Mass of Liquid/Steam Generator, thousand lbs 94 Mass of Secondary Coolant, thousand lbs 2260 Steam Generator Blowdown Rate, thousand lbs/hr 61 Steam Generator Blowdown Tank Vent Option Not applicable because of cooling by heat exchangers. HEPA/Charcoal Treatment of Releases Gaseous Waste Processing System See Section 11.3.6.1 Reactor Building Purge See Section 11.3.6.2 Auxiliary Building Ventilation See Section 11.3.6.3 Condenser Air Removal System See Section 11.3.6.4.2 Gas Stripping of Letdown Flow Option Continuous purging of volume control tank. (1) Conservative scenario that maximizes the estimated annual gaseous effluents via Reactor Building purge. The current Technical Specifications limit the Reactor Building purge at power to no more than 1000 hours per year at a design flow of 600 cfm. RN 02-028 RN 02-028 RN 02-028 02-01 02-01 02-01

TABLE 11.3-8 11.3-26 Reformatted Per CALCULATED RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM THE PLANT (1) Releases (Ci/yr) Nuclide Gaseous Waste Processing System Reactor Building Purge Exhaust Auxiliary Building Leakage Turbine Building Leakage 02-01 Condenser Air Removal System Total Kr-83m (2) 1 (2) (2) (2) 1 Kr-85m (2) 11 2 (2) 2 15 Kr-85 210 4 (2) (2) (2) 220 Kr-87 (2) 2 1 (2) (2) 3 Kr-88 (2) 14 5 (2) 3 22 Kr-89 (2) (2) (2) (2) (2) (2) Xe-131m 3 10 (2) (2) (2) 13 Xe-133m (2) 42 2 (2) 1 45 Xe-133 1 2500 110 (2) 71 2700 Xe-135m (2) (2) (2) (2) (2) (2) Xe-135 (2) 54 7 (2) 4 65 Xe-137 (2) (2) (2) (2) (2) (2) Xe-138 (2) (2) 1 (2) (2) 1 I-131 (2) 1.4 x 10-2 4.6 x 10-3 1.1 x 10-3 2.8 x 10-3 2.3 x 10-2 I-133 (2) 1.1 x 10-2 6.6 x 10-3 1.3 x 10-3 4.1 x 10-3 2.3 x 10-2 Mn-54 4.5 x 10-5 2.1 x 10-4 1.8 x 10-4 (3) (3) 4.3 x 10-4 Fe-59 1.5 x 10-5 7.3 x 10-5 6.0 x 10-5 (3) (3) 1.5 x 10-4 Co-58 1.5 x 10-4 7.3 x 10-4 6.0 x 10-4 (3) (3) 1.5 x 10-3 Co-60 7.0 x 10-5 3.3 x 10-4 2.7 x 10-4 (3) (3) 6.7 x 10-4 Sr-89 3.3 x 10-6 1.7 x 10-5 1.3 x 10-5 (3) (3) 3.3 x 10-5 Sr-90 6.0 x 10-7 2.9 x 10-6 2.4 x 10-6 (3) (3) 5.9 x 10-6 Cs-134 4.5 x 10-5 2.1 x 10-4 1.8 x 10-4 (3) (3) 4.3 x 10-4 Cs-137 7.5 x 10-5 3.7 x 10-4 3.0 x 10-4 (3) (3) 7.4 x 10-4 C-14 7 1 8 H-3 580 Ar-41 25 25 (1) Based upon the parameters given in Table 11.3-7. (2) Less than 1 Ci/yr noble gases, less than 10-4 Ci/yr for iodine. (3) Less than 1% of total for nuclide. Amendment 02-01

11.3-27 Reformatted Per Amendment 02-01 TABLE 11.3-9 STACK RELEASE INFORMATION Item No. (1) Item Location (Building) Base Elevation Exit Elevation Exit Area Cross Section Volume Flow Rate Estimated Exit Velocity

1.

Main Plant Vent Auxiliary 511-0 524-0 72 by 96 in 172,000 cfm 3,583 fpm

2.

Purge Exhaust Auxiliary 511-0 524-0 36 by 36 in 20,000 cfm 2,220 fpm

3.

Condensate Return Unit Vent Auxiliary 418-0 455-3 28.9 in2 2,700 lb/hr (2) 6,000 fpm

4.

Air Exhaust Control 505-0 512-0 72 by 48 in 9,000 cfm 375 fpm

5.

Air Exhaust Control 505-0 512-0 72 by 48 in 9,000 cfm 375 fpm 6 Air Exhaust (3) Intermediate 463-0 492-0 38 by 14 in 10,200 cfm 2,760 fpm

7.

Air Exhaust (3) Intermediate 485-0 463-0 84 by 24 in 40, 000 cfm 2,860 fpm

8.

Condenser Exhaust (4) Turbine 415-0 454-6 113.1 in2 800 lb/hr 750 fpm

9.

Main Steam Dump (5) (3 points) Intermediate 463-0 471-0 101.6 in2 each 740,000 lb/hr each (6)

10.

Main Steam Safety and Relief Valves(7) (10 points) Intermediate 463-0 475-0 233.7 in2 each 930,000 lb/hr each 40,000 fpm Main Steam Safety and Relief Valves(7) (5 points) Auxiliary 485-0 497-0 233.7 in2 each 930,000 lb/hr each 40,000 fpm 02-01 02-01 02-01

11.3-28 Reformatted Per Amendment 02-01 TABLE 11.3-9 (Continued) STACK RELEASE INFORMATION Item No. (1) Item Location (Building) Base Elevation Exit Elevation Exit Area Cross Section Volume Flow Rate Estimated Exit Velocity

11.

Main Steam Power Relief Valves(5) (2 points) Intermediate 463-0 475-0 113.1 in2 each 740,000 lb/hr each (6) Main Steam Power Relief Valve(5) (1 point) Auxiliary 485-0 497-0 113.1 in2 each 740,000 lb/hr each (6)

12.

Roof Vent (3 points) Turbine 503-0 48 in dia each 34,087 cfm each 1,115 fpm

13.

Roof Vent (7 points) Turbine 533-0 120 in dia each 199,958 cfm each 1,000 fpm

14.

Reheat Steam Safety Relief Valves(8) (4 points) Turbine 474-0 531-0 975.8 in2 2.33 x 106 lb/hr 40,000 fpm

15.

EFW Pump Exhaust Intermediate 420-0 475-0 135 in2 15,000 cfm 15,000 fpm 02-01 02-01 02-01

11.3-29 Reformatted Per Amendment 02-01 TABLE 11.3-9 (Continued) STACK RELEASE INFORMATION NOTES (1) See Figure 11.3-8 for location. (2) Condensate return unit vent volume flow rate is maximum theoretically possible. (3) Location of intermediate building air exhaust relative to air intakes has not been finalized. (4) Condenser exhaust flow is continuous. (5) Main steam dump and power relief estimated occurrences are as follows:

a.

Actual, 22 times per year for 10 minutes each time.

b.

Test, 12 times per year for 1 minute each time. (6) Main steam dump and power relief exit velocities are not available. Each vent incorporates a valve and exit diffuser of a proprietary design which diffuses and disperses the flow in a horizontal pattern, 360 degrees around the vent vertical axis. The volume flow rate is the flow through the vent stack prior to diffuser release. (7) Main steam safety and relief estimated occurrences are as follows:

a.

Actual, 2 times per year for 2 minutes each time.

b.

Test, 3 times per year for less than 1 minute each time. (8) Not expected to occur during life of plant. Test monthly for 10 minutes each time. 02-01

11.3-30 Reformatted Per Amendment 00-01 TABLE 11.3-9a COMPARISON OF NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS WITH BRANCH TECHNICAL POSITION ETSB 11-2 Branch Technical Position ETSB 11-2 Item Reactor Building Purge Exhaust System Reactor Building Charcoal Cleanup System Auxiliary Building HEPA Exhaust System Auxiliary Building Charcoal Exhaust System Fuel Handling Building Charcoal Exhaust System Controlled Access Area Exhaust System Hot Machine Shop Ventilation System B-1-a System complies, except design is for intermittent operation. System complies, except design is for intermittent operation. System complies. System complies. System complies. System complies. System complies. B-1-b System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-1-c System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-1-d System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-2-a System complies, except that heaters or cooling coils have not been used because relative humidity will not exceed 70 percent. System complies, except that heaters or cooling coils have not been used because relative humidity will not exceed 70 percent. System complies, except charcoal filters are not used. System complies, except that heaters or cooling coils have not been used because relative humidity will not exceed 70 percent. System complies, except that heaters or cooling coils have not been used because relative humidity will not exceed 70 percent. System complies, except that heaters or cooling coils have not been used because relative humidity will not exceed 70 percent. System complies, except charcoal filters are not used. B-2-b System complies. System complies. System plenum capacity is 50,000 cfm, 2 banks each, 5 filters wide by 5 filters high. System plenum capacity is 45,000 cfm, 9 filters wide by 5 filters high. System complies. System complies. System complies. B-2-c System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. System has local differential pressure devices across filter banks - no alarm. B-2-d Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. Wiring was purchased and qualified to IPCEA and IEEE Standards. B-2-e System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-2-f System complies. System complies. System complies. System complies. System complies. System complies. System complies. 02-01

11.3-31 Reformatted Per Amendment 00-01 TABLE 11.3-9a (Continued) COMPARISON OF NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS WITH BRANCH TECHNICAL POSITION ETSB 11-2 Branch Technical Position ETSB 11-2 Item Reactor Building Purge Exhaust System Reactor Building Charcoal Cleanup System Auxiliary Building HEPA Exhaust System Auxiliary Building Charcoal Exhaust System Fuel Handling Building Charcoal Exhaust System Controlled Access Area Exhaust System Hot Machine Shop Ventilation System B-2-g Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Plenums were specified to indicate no leakage through soap bubble testing with ducts at 2 psig. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. Ducts were specified to satisfy leakage testing requirements of SMACNA, Chapter 8, 1967. B-3-a Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. Heaters not used in exhaust system plenums because relative humidity is not expected to exceed 70 percent. B-3-b System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement B-3-c System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement B-3-d System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement System complies. with this requirement B-3-e System generally complies, except access has been provided on one side of plenum. System generally complies, except access has been provided on one side of plenum. System generally complies, except access doors have been provided on one side of plenum and capacity is 51,000 cfm. System generally complies, except access doors have been provided on one side of plenum and capacity is 45,000 cfm. System generally complies, except access doors have been provided on one side of plenum. System generally complies, except access doors have been provided on one side of plenum. System generally complies, except access doors have been provided on one side of plenum. 02-01 02-01 02-01

11.3-32 Reformatted Per Amendment 00-01 TABLE 11.3-9a (Continued) COMPARISON OF NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS WITH BRANCH TECHNICAL POSITION ETSB 11-2 Branch Technical Position ETSB 11-2 Item Reactor Building Purge Exhaust System Reactor Building Charcoal Cleanup System Auxiliary Building HEPA Exhaust System Auxiliary Building Charcoal Exhaust System Fuel Handling Building Charcoal Exhaust System Controlled Access Area Exhaust System Hot Machine Shop Ventilation System B-3-f System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. System generally complies, except testing is in accordance with SMACNA Standards. B-3-g System complies. System complies. Not applicable. System complies. System complies. System complies. Not applicable B-3-h System complies. System complies. Not applicable System complies. System complies. System complies. Not applicable B-3-i System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. B-3-j System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-3-k System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-3-l System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. System generally complies. B-4-a System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-4-b System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. System complies, except door sizes are 30 by 60. B-4-c System complies. System complies. System complies. System complies. System complies. System complies. System complies. B-4-d Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. Openings have been provided for insertion of test probes. B-4-e System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. System startup and operating procedures comply with this requirement. 02-01

11.3-33 Reformatted Per Amendment 00-01 TABLE 11.3-9a (Continued) COMPARISON OF NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS WITH BRANCH TECHNICAL POSITION ETSB 11-2 Branch Technical Position ETSB 11-2 Item Reactor Building Purge Exhaust System Reactor Building Charcoal Cleanup System Auxiliary Building HEPA Exhaust System Auxiliary Building Charcoal Exhaust System Fuel Handling Building Charcoal Exhaust System Controlled Access Area Exhaust System Hot Machine Shop Ventilation System B-5-a Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. Filter plenum tested in place initially and at frequency not to exceed 18 months. Tests include visual inspection. B-5-b Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. Filter system total flow rate will be checked for +/-10 percent. B-5-c Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. Filter plenums will be DOP tested at frequency noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with ANSI 101.1 Section 3. B-5-d Filter plen. will be tested with refrig. at the freq. noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with App. B of DP-1082. Filter plen. will be tested with refrig. at the freq. noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with App. B of DP-1082. Not Applicable, Filter plen. will be tested with refrig. at the freq. noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with App. B of DP-1082. Filter plen. will be tested with refrig. at the freq. noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with App. B of DP-1082. Filter plen. will be tested with refrig. at the freq. noted in B-5-a above & following painting, fire or chem. release near the filter plenum. Filters will be tested in accor. with App. B of DP-1082. Not Applicable. 02-01 02-01 02-01

11.3-34 Reformatted Per Amendment 00-01 TABLE 11.3-9a (Continued) COMPARISON OF NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS WITH BRANCH TECHNICAL POSITION ETSB 11-2 Branch Technical Position ETSB 11-2 Item Reactor Building Purge Exhaust System Reactor Building Charcoal Cleanup System Auxiliary Building HEPA Exhaust System Auxiliary Building Charcoal Exhaust System Fuel Handling Building Charcoal Exhaust System Controlled Access Area Exhaust System Hot Machine Shop Ventilation System B-6-a, b Char. media from those plen. shall be lab tested at a freq. not to exceed 18 mo. Sample shall be obtained by removing 1 entire bed from an adsorber tray & with drawing a 2 thick x 2 dia. sample. Lab test shall verify an iodine removal eff. of 90 percent for radioactive methyl iodine and it shall be in accor. w/RDT M16-IT para. 4.5.3. Char. media from those plen. shall be lab tested at a freq. not to exceed 18 mo. Sample shall be obtained by removing 1 entire bed from an adsorber tray & with drawing a 2 thick x 2 dia. sample. Lab test shall verify an iodine removal eff. of 90 percent for radioactive methyl iodine and it shall be in accor. w/RDT M16-IT para. 4.5.3. Not Applicable. Char. media from those plen. shall be lab tested at a freq. not to exceed 18 mo. Sample shall be obtained by removing 1 entire bed from an adsorber tray & with drawing a 2 thick x 2 dia. sample. Lab test shall verify an iodine removal eff. of 90 percent for radioactive methyl iodine and it shall be in accor. w/RDT M16-IT para. 4.5.3. Char. media from those plen. shall be lab tested at a freq. not to exceed 18 mo. Sample shall be obtained by drawing a 2 thick x 2 dia. sample. Lab test shall verify an iodine removal eff. of 95 percent for radioactive methyl iodine and it shall be in accor. w/ ASTM D3803-1989 at a test media temperature of 30°C. Char. media from those plen. shall be lab tested at a freq. not to exceed 18 mo. Sample shall be obtained by removing 1 entire bed from an adsorber tray & with drawing a 2 thick x 2 dia. sample. Lab test shall verify an iodine removal eff. of 90 percent for radioactive methyl iodine and it shall be in accor. w/RDT M16-IT para. 4.5.3. Not Applicable. 02-01 02-01 RN 02-034 00-01 00-01 00-01

11.3-35 Reformatted Per Amendment 02-01 TABLE 11.3-10 COMPARISON OF RADIONUCLIDE CONCENTRATIONS IN GASEOUS EFFLUENTS TO THE LIMITS OF 10 CFR 20 Isotope Annual Release (Ci/yr) (Expected) (1) Site Boundary Concentration (µCi/ml) Effluent (2) Concentration Limit In Air (µCi/ml) Ratio of Expected Concentration to Concentration Limit Kr-85m 15 2.6E-12 1.0E-7 2.6E-5 Kr-85 220 3.7E-11 7.0E-7 5.3E-5 Kr-87 3 5.1E-13 2.0E-8 2.6E-5 Kr-88 22 3.7E-12 9.0E-9 4.1E-4 Xe-131m 13 2.2E-12 2.0E-6 1.1E-6 Xe-133m 45 7.6E-12 6.0E-7 1.3E-5 Xe-133 2700 4.6E-10 5.0E-7 9.2E-4 Xe-135 65 1.1E-11 7.0E-8 1.6E-4 Xe-138 1 1.7E-13 2.0E-8 8.5E-6 I-131 2.3E-2 3.7E-15 2.0E-10 1.9E-5 I-133 2.3E-2 3.7E-15 1.0E-9 3.7E-6 Mn-54 4.3E-4 7.3E-17 1.0E-9 7.3E-8 Fe-59 1.5E-4 2.6E-17 5.0E-10 5.2E-8 Co-58 1.5E-3 2.6E-16 1.0E-9 2.6E-7 Co-60 6.7E-4 1.1E-16 5.0E-11 2.2E-6 Sr-89 3.3E-5 5.6E-18 2.0E-10 2.8E-8 Sr-90 5.9E-6 1.0E-18 6.0E-12 1.7E-7 Cs-134 4.3E-4 7.3E-17 2.0E-10 3.7E-7 Cs-137 7.4E-4 1.3E-16 2.0E-10 6.5E-7 C-14 8 1.3E-12 3.0E-9 4.3E-4 H-3 580 9.7E-11 1.0E-7 9.7E-4 Ar-41 25 4.2E-12 1.0E-8 4.2E-4 TOTAL 3.5E-3 (1) Expected concentration in worst sector averaged over a one-year period. (2) From 10 CFR 20, Appendix B, Table 2, Column 1. RN 02-028

11.3-36 Reformatted Per Amendment 02-01 TABLE 11.3-11

SUMMARY

OF CALCULATED GASEOUS PATHWAY DOSES VIRGIL C. SUMMER NUCLEAR STATION Organ Receiving Maximum Dose Pathway Location Age Group Organ Dose (mrem/yr) Total Body Dose (mrem/yr) Cloud Submersion Nearest Residence (1.1 Miles ESE) All Skin 2.8E-1 1.1E-1 Ground Plane Contamination Nearest Residence (1.1 Miles ESE) All Skin 9.8E-3 8.4E-3 Air Inhalation Nearest Residence (1.1 Miles ESE) Adult Teen Child Infant Thyroid Thyroid Thyroid Thyroid 8.6E-2 5.5E-2 6.3E-2 8.7E-2 5.8E-2 3.2E-2 3.2E-2 3.4E-2 Vegetable Ingestion Nearest Residence (1.1 Miles ESE) Adult Teen Child Bone Thyroid Thyroid 7.4E-1 4.4E-1 8.7E-1 2.7E-1 3.4E-1 7.1E-1 Cow Milk Ingestion Nearest Cow (1.5 Miles WSW) (Not now milked) Adult Teen Child Infant Thyroid Thyroid Thyroid Thyroid 3.5E-1 5.2E-1 1.0E-0 2.4E-0 5.3E-2 7.5E-2 1.6E-1 3.2E-1 Meat Ingestion Nearest Cow (1.5 Miles WSW) Adult Teen Child Bone Thyroid Thyroid 1.4E-1 3.3E-2 5.7E-2 3.7E-2 2.6E-2 4.6E-2

11.3-37 Reformatted Per Amendment 02-01 TABLE 11.3-12 APPENDIX I CONFORMANCE

SUMMARY

TABLE VIRGIL C. SUMMER NUCLEAR STATION GASEOUS EFFLUENTS Appendix I Criteria Virgil C. Summer Nuclear Station Type of Dose Design Objective (1) Point of Dose Evaluation Calculated Dose Point of Dose Evaluation (9) Gaseous Effluents (3) Gamma dose in air 10 mrad/yr per site Location of the highest dose offsite (2) 0.29 mrad/yr Location of highest annual average concentration at the site boundary (SE at 1 mile) Beta dose in air 20 mrad/yr per site Same as above 0.62 mrad/yr Same as above Dose to total body 5 mrem/yr per site Location of the (2) highest dose offsite 0.11 mrem/yr Nearest residence (ESE at 1.1 miles) Dose to skin of an individual 15 mrem/yr per site Same as above 0.28 mrem/yr Same as above Radioiodines and Particulates (5) Released to the Atmosphere Dose to any organ from all pathways 15 mrem/yr per site Location of the (6) highest dose offsite 2.49 mrem/yr (8) Nearest cow (7) (WSW at 1.5 miles) 02-01 02-01

11.3-38 Reformatted Per Amendment 02-01 TABLE 11.3-12 (Continued) APPENDIX I CONFORMANCE

SUMMARY

TABLE VIRGIL C. SUMMER NUCLEAR STATION GASEOUS EFFLUENTS (1) Design objectives as specified in the Commissions Appendix I Conformance Option, 40 FR 40816, September 4, 1975. (2) Evaluated at a location that is anticipated to be occupied during plant lifetime or evaluated with respect to such potential land and water usage and food pathways as could actually exist during the term of plant operation. (3) Calculated only for noble gases. (4) Evaluated at a location that could be occupied during the term of plant operation. (5) Doses due to carbon-14 and tritium intake from terrestrial food chains are included in this category. (6) Evaluated at a location where an exposure pathway actually exists at time of licensing. However, if the applicant determines design objectives with respect to radioactive iodine on the basis of existing conditions and if potential changes in land and water usage and food pathways could result in exposures in excess of the guideline values given above, the applicant should provide reasonable assurance that a monitoring and surveillance program will be performed to determine:

1) the quantities of radioactive iodine actually released to the atmosphere and deposited relative to those estimated in the determination of design objectives; 2) whether changes in land and water usage and food pathways which would result in individual exposures greater than originally estimated have occurred; and 3) the content of radioactive iodine and foods involved in the changes, if and when they occur.

(7) Cows are not currently milked at this location. Doses evaluated were based on the assumption that pathways could reasonably be expected to exist during plant life. (8) Dose to an infant thyroid from air inhalation and cow milk ingestion. (9) Points given correspond to points of dose evaluation under Appendix I heading.

70 Equilibrium Inventory Of All Other Gaseous Isotopes 60 (i) Q) 50

l Co) 0

~ 40 KR-85 Buildup 0......c Q)>c 30 (j) co CJ Gaseous Inventory At 40 Years c 0 20 (j) KR-85: 53 Kilocuries (j) LL Other Isotopes: 60 Kilocuries RN Total: 113 K ilocu ries 02-025 10 o o 10 20 30 40 50 Time (Years) RN 02-025 July 2002 SOUTH CAROLINA ELECTRIC &GAS CO VIRGIL C. SUMMER NUCLEAR STATION Gaseous Waste Processing System Fission Gas Accumulation Based on Continuous Core Operation at 2958 Mwt with 1% Fuel Defects and 60 gpm Continuous Letdown Figure 11.3-1

3 7 6 5 ~ 4 o.....c CD> C en ctS C9 c .~ 2 en u:: 1 o KR-85 Buildup Gaseous Inventory At 40 Years KR-85: 6.4 Kilocuries Other Isotopes: 0.8 Kilocuries Total: 7.2 Kilocuries Equilibrium Inventory Of All Other Gaseous Isotopes RN 02-025 o 10 20 30 40 50 Tim e (Years) RN 02-025 July 2002 SOUTH CAROLINA ELECTRIC & GAS CO VIRGIL C. SUMMER NUCLEAR STATION Estimated Gaseous Waste Processing System Fission Gas Accumulation Based on Table 11.1-5 and Full Power Operation at 2958 Mwt and 60 gpm Continuous Letdown Figure 11.3-2

!.q£N:;:c,.;wQ:,'-'**'1 ~l' J_~ (J,lAl..YTj(llElXNllIE:RS 'N f'NW.J,.,tl <ii'.i'"'" )>-_tJ-'--....~f-----------------I SOUTH CAROLINA ELECTRIC & GAS CO. VIRGllC. SUMMER NUCLEAR STATION Gaseous Waste Processing System Process flow Diagram figure 11.3-3 Amendment 0 August 1984

RN 04-020 February 2005

RN 05-041 October 2005

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lI£Y-143 17827-WG ESSENTIAL THIS IS A NUCLEAR SAFETY RELATED JlIlCIAIENT. NO DEVIATION SHALL BE tIITt"TEO OR PERFORNEIl WITHOUT PRIOR OOCUNENTATION.oN[) WRITTEN _OVAL FSAR Figure 11.3-4, SH. 3 A B c o G H K

TO RECOMBINER COMPRESSOR FROM VOLUME CONTROL TANK MOISTURE SEPARATOR COOLER SEAL WATER RETURN T - TEMPERATURE MEASUREMENT P - PRESSURE MEASUREMENT L - LEVEL MEASUREMENT SOUTH CAROLINA ELECTRIC & GAS CO. VIRGILC. SUMMER NUCLEAR STATION Waste Gas Compressor Package Figure 11.3-5 Amendment 0 August 1984

FROM GAS COMPRESSOR p HEATER OXYGEN CATALYT IC REACTOR p TO GAS DECAY TANK TO GAS ANALYZER p PHASE SEPARAI uR T T T T TO GAS ANALYZER COOLER CONDENSOR t---~ T TO GAS ANALYZER TO GAS ANALYZER T - TEMPERATURE MEASUREMENT P.. PRESSURE MEASUREMENT F - FLOW MEASUREMENT SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLEAR STATION Catalytic Hydrogen Recombiner Package Figure 11.3-6 Amendment 0 August 1984

Gaseous Waste Release Points SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLEAR STATION o WEST PENETRATlON ACCESS AREA VENTll. VENTS ON AUXILIARY BUILDING ~ INTERMITTENT & CONTROLLED SOURCES o HEPA FILTER [£] CHARCOAL FILTER ---... CONTINUOUS SOURCES ELEV.524' "- ElEV. 524' I PLANT EXHAUST 11' 1 MAIN PLANT VENT HOT MACHINE SHOP EXHAUST REACTOR I BUILDING I FUEL HANDLING BUILDING ,,....~, ~-...~_.1 ~: -IHlcIH P -J~EXCESS WASTE HOLDUP TANK ( ) ""- ____..... __..J ~DECONTAMINATIONPIT HOLDUP TANK HOT MACHINE SHOP L-...JHlp: DECONTAMINATION ~lpIH: VENTS FROM NON* RADIOACTIVE SOURCES AREA EXHAUST 011( WASTE EVAPORATOR CONCENTRATES TANK 011( RECYCLE HOLDUP TANKS (2) CONTROLLED plHlclHI-WASTE HOLDUP TANK ACCESS EXHAUST Ill( 8 ..... _..r - - WASTE EVAPORATOR 8 -fH C I Hlp: L __ RECYCLE EVAPORATOR III( FLOOR DRAIN TANK 7 VENTS 1 CHEMICAL DRAIN TANK 3 VENTS ." t III( ElEV.303' ElEV.533' 011( SHUTDOWN (GAS DECAY) TANKS (2) 11' MAIN TURBINE BUILDING ~ __ PRIMARY SPENT RESIN TANK 011( SECONDARY STEAM LEAKAGE VENTS TO LOCAL CUBICLES ~STEAM PACKING EXHAUSTER AUXILIARY BUILDING SUMP .-- LAUNDRY & HOT SHOWER TANK DEAERATOR CONDENSER AIR REMOVAL ~ WASTE EVAPORATOR CONDENSATE TANK BAY f-WASTE MONITOR TANKS SYSTEM __ SECONDARY SPENT RESIN TANK -~---~---~---- AUXILIARY BUILDING LEGEND: [!] PRE - FILTER OR ROUGHING FILTER PENETRATION ACCESS AREAS ~EAST PENETRATlON ACCESS AREA VENTll. O CIRCLED NUMBERS CORRESPOND TO VENT NUMBERS IN TABLE 11.3.7-1 AND FIGURE 11.3.7-2 Figure 11.3-7 Amendment 0 August 1984

DIESEL GENERATOR \\ BUILDING ~ EXHAUST STAt" _P"ER I1J> OPERATOR RELIEF E_ERGENCY FEEDWUER PUMP EXHAUST ROOFVE7 -ill-INTERMEDIATE BUILDING 1UlNJ STEAM OUMP i.......-.-... REHEAT SHAM-.....-...l..-: RELIEFS I -{1)- TURBINE GENERATOR o I I FUEL HANDLING BUILDING MAIN CD PLANT VENTI C8J /ill E~~:~~TJ (j) POWER OPERATOR RELIEF \\ AUXILIARY BUILDING n rl n4 -ItT 0 Lb**** o CONTROL BU ILO I NG <Do SOUTH CAROLINA ELECTRIC & GAS CO. VIRGILC. SUMMER NUCLEAR STATION Potentially Radioactive Gaseous Waste Release Points Figure 11.3-8 Amendment 0 August 1984

11.4-1 Reformatted November 2011 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS Radiation monitors and analysis of samples are used to monitor process and effluent streams in order to record and control releases of radioactive materials generated as a result of normal operations, including anticipated operational occurrences, and during postulated accidents. The general description and approximate ranges of radiation monitors that measure, indicate, and record the activity of process and effluent streams in the plant are tabulated in Table 11.4-1. Additional radiation monitors are discussed in Sections 12.1.4 and 12.2.4. More detailed information on instrument ranges may be found in Health Physics procedures. 11.4.1 DESIGN OBJECTIVES Major and potentially significant effluent discharge paths for release of radioactive material during normal reactor operation, including anticipated operational occurrences, are monitored as required by General Design Criterion 64. These effluent discharge paths are as follows:

1.

Major Effluent Discharge Paths

a.

Reactor Building purge exhaust

b.

Plant vent exhaust (including the gaseous waste discharge)

c.

Liquid waste discharge

2.

Potentially Significant Effluent Discharge Paths

a.

Condenser exhaust

b.

Nuclear Blowdown waste effluent

c.

Steam Generator Blowdown

d.

Turbine room sump

e.

Condensate Polish Backwash Tank effluent Additionally, samples of these effluents are required for laboratory analysis to allow evaluation of the measured releases and compliance with the reporting requirements of Regulatory Guide 1.21 (see Appendix 3A). Monitoring and analysis of samples provide information useful in complying with the objectives of Appendix I to 10 CFR 50. 02-01 02-01

11.4-2 Reformatted November 2011 Process systems which do not discharge to the environment are monitored to permit identification of possible system or equipment malfunction through detection of activity levels in the system. The systems monitored are the following:

1.

Component Cooling Water

2.

Primary Coolant Letdown

3.

Spent Fuel Cooling Water The Radiation Monitoring System continuously monitors the normal plant effluent discharge paths under steady-state, transient, or accident condition. After an accident the system provides information to aid in determining the magnitude of the accident. Locations where permanently mounted monitors are not provided are monitored periodically with portable instruments. Radiation monitoring, effluent flow measurements and meteorological instrumentation provide data which, under emergency procedures, determine the direction and rate of movement of radioactive effluents. The monitors located in the normal plant effluent discharge paths, area gamma monitors, and manual sampling provide information on the rate and location of releases. Radiation monitoring, analysis of samples, and Health Physics surveillance provide data used for assuring compliance with the requirements of 10 CFR 20. Utilization of off-line process monitors provides improved access for maintenance and inspection. Remotely actuated check sources for the radiation monitor detectors facilitate routine functional verification. Detector assemblies are removable in the event that extensive maintenance is required. Decontamination of process monitors is facilitated by use of appropriate taps or valving to allow flushing or cleaning. Expected process flow, composition, and contamination are a function of the specific process system being monitored and are discussed in the Section where the system is discussed. 11.4.2 CONTINUOUS MONITORING Measurement data from the effluent radiation monitors and the analysis of samples by laboratory equipment (see Section 11.4.3) provide information used to comply with Regulatory Guide 1.21 (see Appendix 3A) and for evaluation of the releases to the environment. RN 07-037

11.4-3 Reformatted November 2011 Measured activity levels are indicated and recorded on the Radiation Monitoring control panel located in the Control Room (except RM-L8 and RM-L11). Local indication at or near the detector location is provided for each channel. The turbine room sump radiation monitor, RM-L8, is equipped for local indication, recording, and high level alarm and also actuates an alarm in the Control Room. If radiation discharge limits are exceeded, the sump pumps are automatically tripped and the flow path may be aligned to discharge to the Excess Liquid Waste System. The condensate polish backwash effluent radiation monitor RM-L11 is equipped for local indication, recording, and high level alarm. Alarm is also provided through the local control panel for the condensate polishing system. A high radiation alarm will trip the backwash tank discharge pumps. Detectors have remotely actuated check sources to provide functional verification. In addition, the monitoring channels are routinely calibrated by exposure to a calibrated test source for verification against initial calibration curves. Radiation monitor detection range is based upon the type of fluid being monitored for activity, the capability of commercial instrumentation, and the function of the monitor. Each rate meter is equipped with two adjustable alarm levels (alert and high) and a channel failure/loss of power alarm. These alarms, except for movable monitors, are annunciated on the Radiation Monitoring System control panel in the Control Room. Channels which have an interlock function with other systems (see Figure 11.4-1) are provided with a bypass switch for use during maintenance or testing. Use of the bypass switch is annunciated. Reliable power for the fixed, process, and effluent radiation monitoring instrumentation is obtained from the diesel backed, 120 volt instrument bus. Associated sample pumps obtain power from the 480 volt diesel backed bus. Radiation monitors RM-L8 and RM-L11 are powered by local 120 volt a-c power. Individual ratemeter power supplies are independent of each other. The reliable a-c power source is divided between two trains (see Table 11.4-1) to provide a high degree of availability. Specific instrumentation channels used for process and effluent monitoring are as follows:

1.

Primary Coolant Letdown Monitor, Channel RM-L1 A sample of the primary coolant letdown from the Nuclear Sampling System is monitored by this channel utilizing an off-line monitor located outside the Reactor Building (see Figure 9.3-4). The sample is delayed to allow decay of N16 activity. Two detectors, overlapping in ranges, as shown by Table 11.4-1, are used to obtain a wide measuring range, from approximately 1 x 10-3 µCi/cc to 103 µCi/cc, based upon Cs-137. 02-01

11.4-4 Reformatted November 2011 The reactor coolant gross gamma activity resulting from instantaneous mixing of fission products from a small fraction of one fuel rod at equilibrium causes a net signal increase approximately equal to the signal produced by 1 percent failed fuel within four minutes. The contribution to gross gamma activity from corrosion products and tramp uranium is minor compared to that from 1 percent failed fuel. The alarm setpoint is established at a level such that any sudden increase above the normal or expected operational reactor coolant activity is suspected as being due to gross fuel failure. Analyses of a manual sample provide verification of compliance with Technical Specifications and indicates any action to be taken. No automatic interlock functions are served by this channel. If the detector is not available, the sampling frequency is increased. Loss of sample flow is annunciated in the Control Room. Use of two detectors, scintillation and Geiger-Mueller (G-M), and two alarm modules provides a high level of availability.

2.

Component Cooling Water Monitors, Channels RM-L2A and RM-L2B Each of the two Component Cooling Water loops (see Figure 9.2-4) is equipped with an off-line, shielded sampler having a gamma sensitive scintillation detector for measurement of gross gamma activity in a sample of the cooling water. These monitors detect leakage of radioactivity into the Component Cooling Water and cause an alarm to sound upon detection of excessive activity levels. Sensitivity and range of these monitors is approximately 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137. High activity, approximately one decade above normal background, causes an alarm and initiates closure of the Component Cooling Water surge tank vent valve. Laboratory analysis of a Component Cooling Water sample is used to verify the cause of an alarm and the extent of leakage into the Component Cooling Water loop.

3.

Steam Generator Blowdown Monitor, Channel RM-L3 A fluid sample from the combined Steam Generator Blowdown (see Figure 10.4-13) is passed through a shielded, off-line sampler having a gamma sensitive scintillation detector for measurement of gross gamma activity. This monitor is used to detect primary to secondary leakage and sound an alarm upon detection. Design, sensitivity and range are similar to RM-L2A. High activity causes an alarm and initiates valve operations to divert blowdown flow to the Nuclear Blowdown holdup tank. Laboratory analysis of Steam Generator Blowdown samples is used to verify the alarm condition. 02-01 02-01 02-01 02-01

11.4-5 Reformatted November 2011

4.

Spent Fuel Cooling Water, Channel RM-L4 A sample of Spent Fuel Cooling Water (see Figure 9.1-3) is monitored by a shielded, off-line sampler using a gamma sensitive scintillation detector for measurement of gross gamma activity. Alarms are sounded upon detection of excessive activity. Design, sensitivity, and range are similar to RM-L2A. The monitor alarms alert the operator to an increase of activity. Such an increase is then verified by laboratory analysis of a manual grab sample.

5.

Liquid Waste Effluent Monitor, Channel RM-L5 An off-line monitor is used to sample the Liquid Waste effluent. A lead shielded sampler with a gamma sensitive scintillation detector is located upstream of the discharge dilution point for measurement of effluent gross gamma activity (see Figure 11.2-2, Sheet 4). Sensitivity and range are similar to RM-L2A. High effluent activity causes an alarm and initiates closure of the waste processing discharge valve. The activity level at which this interlock is actuated is adjustable based upon the activity level of an analyzed sample of the waste batch and/or dilution flow available. The activity and dilution flow are evaluated prior to discharge. A high level alarm is also provided at the local Liquid Waste System control panel.

6.

Boron Recycle System Monitor, Channel RM-L6 --NO LONGER IN SERVICE-- A sample of the discharge (see Figure 9.3-18) from the Boron Recycle System recycle evaporator condensate filter to the reactor makeup storage tank is monitored by an off-line, shielded sampler. This sampler uses a gamma scintillation detector to measure gross gamma activity. Sensitivity and range are similar to RM-L2A. High activity in the recycle evaporator condensate filter discharge to the reactor makeup storage tank causes an alarm and initiates diversion of the discharge flow from the reactor makeup storage tank to the recycle evaporator. Laboratory analysis of a sample is used to verify the alarm condition.

7.

Nuclear Blowdown Waste Effluent Monitor, Channel RM-L7 A sample of the discharge (see Figure 10.4-14) from the Nuclear Blowdown Processing System is monitored for gross gamma activity by a shielded, gamma sensitive detector. This off-line monitor is used to detect excessive activity releases from the Nuclear Blowdown Processing System. Alarms are sounded upon detection of such releases. Design, sensitivity, and range are similar to RM-L2A. RN 07-037

11.4-6 Reformatted November 2011 High activity causes alarms and initiates valve operations to terminate discharge and divert the flow to the Nuclear Blowdown Processing System monitor tank. Laboratory analysis of the process discharge fluid is used to verify alarm conditions. The alarm setpoint is based upon the maximum release activity and dilution of the discharge stream.

8.

Turbine Room Sump Monitor, Channel RM-L8 A sample of the turbine room sump discharge is monitored by a shielded gamma scintillation detector to provide local measurement, indication, and alarm of gross gamma activity. The monitor sample valve is opened and the 120 volt a-c sample pump is started when the turbine room sump pumps are operated. A high activity alarm annunciated in the control room, stops the sump pumps. Sensitivity and design are similar to those of the other fixed liquid monitors.

9.

Main Plant Vent Exhaust Monitor, Channel RM-A3 This channel monitors particulate, iodine, and gaseous activity released through the main plant vent from the Auxiliary Building. The particulate detector, a beta scintillator, has a sensitivity and range of approximately 10-11 to 10-7 µCi/cc, Cs-137. The iodine detector, a gamma scintillator, has a sensitivity and range of approximately 2 x 10-11 to 10-7 µCi/cc,, I-131. The gas detector, a beta scintillator, has a sensitivity and range of approximately 2.6 x 10-6 to 2 x 10-2 µCi/cc, Kr-85. A particulate collection filter with an efficiency of 90 percent for particles 0.3 microns or larger is provided and is removable for laboratory analysis. A charcoal filter cartridge, removable for laboratory analysis and having an efficiency of 95 percent for elemental iodine, is also provided. The plant vent discharge sample is taken downstream of the filtering units by a nozzle located near the point of discharge to the environment (see Figure 9.4-9). Sample nozzle flow is adjusted to be isokinetic at a nominal discharge flow rate (227,588 cfm) from the vent stack. The pumping station and the three shielded detection channels are located on the floor below the sampling point to minimize sample line losses. This pumping station is equipped with two pumps to permit continued operation when pump maintenance is required. Should the gaseous activity released through the main plant vent reach or exceed the Hi-Rad setpoint, the gaseous activity detection channel initiates closure of the Waste Gas decay tank discharge valve. The RM-A3 setpoints are estimated by plant procedures to ensure that the site boundary dose limits specified in the ODCM are not exceeded. RN 02-019 02-01 RN 03-011

11.4-7 Reformatted November 2011 Radiation monitor RM-A3 utilizes detectors which are 4-Pi shielded with 3-inches of lead to minimize the effect of the background radiation in normal operation in an environment as listed in Table 3.11-3.

10. Reactor Building Purge Exhaust Monitor, RM-A4 This monitor measures particulate, iodine, and gaseous activity in the discharge of the Reactor Building Purge Exhaust System during purging operations. This monitor is similar to RM-A3 with respect to its description, range, and sensitivity.

The monitored sample is obtained through an isokinetic nozzle located in the purge exhaust duct upstream of the discharge point (see Figure 9.4-28). The monitor is located on the floor below the duct to minimize sample line loses. The monitor obtains its sample from either of two sample probes in the discharge duct. One probe is isokinetic at the design discharge flow rate (20,000 CFM), the other probe is isokinetic at the design low flow rate. The duct flow rate is measured and indicated in the Control Room. Under operating conditions where this flow is more than 10% below design conditions, the data collected by RM-A4 will be corrected for an isokinetic sampling in accordance with ANSI 13.1 1969 Appendix C. Appendix C implies that an isokinetic correction for particle sizes less than 4 microns does not need to be considered. Particles penetrating the containment HEPA filters are normally less than 0.3 microns. Should an unplanned gaseous activity release rate of approximately 4 x 101 µCi/sec, Kr-85, be exceeded, this monitor initiates closure of the Reactor Building purge valves. This activity release rate is based upon a flow rate of 20,000 cfm through the Reactor Building purge vent and a gaseous activity level setpoint of approximately two times the expected background at the monitor location. Such a setpoint provides for rapid detection of any unplanned release. The alarm setpoints for particulate and iodine activity are estimated by plant procedures to ensure that the site boundary dose limits specified in the ODCM are not exceeded. Alarm setpoints may be readjusted prior to commencement of Reactor Building purging based upon existing activity levels within the Reactor Building. Radiation monitor RM-A4 utilizes detectors which are 4-Pi shielded with 3-inches of lead to minimize the effect of the background radiation in normal operation in an environment as listed in Table 3.11-3. The gamma dose rate from the charcoal cartridge located inside the atmospheric monitor (RM-A4) shield assembly at a distance of one foot is approximately 2.8 mr/hr, assuming a sample activity of 102 µCi/cc, sample flow of 1 cfm, 95% collection efficiency, average E of 0.5 Mev, and 30 minutes collection time. 02-01 02-01 RN 02-019 02-01

11.4-8 Reformatted November 2011

11. Condenser Exhaust Monitor, RM-A9 This monitor is a shielded, off-line unit with a gamma scintillation detector which monitors gaseous samples taken from the condenser exhaust discharge line (see Figure 10.4-1). Its purpose is detection of primary to secondary system leakage.

Range of this monitor is approximately 2 x 10-6 to 2 x 10-2 µCi/cc, Xe-133. This range combined with a design discharge flow rate of 40 cfm permits monitoring of an activity discharge rate of 7.5 x 10-2 to 7.5 x 102 µCi/sec.

12. Waste Gas Discharge Monitor, Channel RM-A10 This monitor is a shielded beta scintillation detector which monitors and controls the gaseous activity released from the Waste Gas decay tanks. Sensitivity and range of this monitor are approximately 2 x 10-4 to 2 µCi/cc, Xe-133.

Detection of high gaseous activity initiates closure of the Waste Gas decay tank discharge valve. Prior to a planned Waste Gas decay tank release, the alarm setpoint may be readjusted according to analyzed activity level existing in the Waste Gas decay tanks.

13. Main Plant Vent Exhaust High Range Gas Discharge Monitor, Channel RM-A13 This monitor (see Figure 11.4-2) provides extended range backup to channel RM-A3. The monitor utilizes an ion chamber gamma sensitive detector with a sensitivity and range of approximately 10-2 to 10+5 µCi/cc, XE-133. Detection of high activity provides an alarm, indication, and a record of excessive gas activity released through the main plant vent exhaust. The energy dependence is +/- 10%

from 80 kev to 3 mev. Background correction is achieved by a 2-inch lead shield which reduces background and provides collimation of the detector to respond to the gamma radiation from the exhaust duct. Graphs will be available to operating personnel to reflect the readout of RM-A13 versus the curie concentration for a given release and an assumed isotopic mixture of the effluent. RN 03-011 RN 03-011 RN 03-011

11.4-9 Reformatted November 2011

14. Liquid Waste Effluent Monitor, Channel RM-L9 An off-line monitor is used to sample the Liquid Waste effluent to the penstocks. A lead shielded sampler with a gamma sensitive scintillation detector is located downstream of radiation monitor RM-L5 to monitor the discharge from the Liquid Waste Processing System or the discharge from the Nuclear Blowdown Processing System. Sensitivity and range of this monitor are similar to RM-L2A.

High effluent activity actuates an alarm and initiates closure of the waste discharge valve, thus terminating flow to the penstocks. The activity level at which this interlock is actuated is adjusted to limit the maximum permitted activity of the discharge to the penstocks and to provide backup for radiation monitors RM-L5 and RM-L7.

15. Steam Generator Blowdown Discharge Monitor, Channel RM-L10 A fluid sample from the Steam Generator Blowdown discharge to the Circulating Water System is passed through a shielded, off-line sampler having a gamma sensitive scintillation detector for measurement of gross gamma activity. This monitor is used to terminate the discharge flow to the Circulating Water System in the event of primary to secondary leakage in the steam generators. Design, sensitivity, and range of this monitor are similar to RM-L2A.

High activity actuates an alarm and initiates valve closure to terminate Steam Generator Blowdown flow to the Circulating Water System. This monitor backs up radiation monitor RM-L3. Laboratory analysis of Steam Generator Blowdown samples is used to verify the alarm condition.

16. Purge Exhaust Effluent High Range Radiation Monitor, Channel RM-A14 This monitor (see Figure 11.4-2) provides extended range backup to channel RM-A4. The monitor utilizes an ion chamber gamma sensitive detector with a sensitivity and range of approximately 10-2 to 10+5 µCi/cc, Xe-133. Detection of high activity provides an alarm, indication, and a record of excessive gas activity released through the purge vent exhaust. The energy dependence is +/- 10% from 80 kev to 3 mev. Background correction is achieved by a 2-inch lead shield which reduces background and provides collimation of the detector to respond to the gamma radiation from the exhaust duct.

Graphs will be available to operating personnel to reflect the readout of RM-A14 versus the curie concentration for a given release and an assumed isotopic mixture of the effluent. RN 03-011

11.4-10 Reformatted November 2011

17. Steam Line High Range Monitors, Channels RM-G19A, B and C Each Main Steam line header, upstream of the relief valves (2806-MS) is provided with a high range gamma sensitive monitor to provide indication of the steam activity in the event of a release equivalent to 2 curies/sec/MWth of Xe-133.

These monitors are responsive to 0.1 mr/hr up to 107 mr/hr gamma with a flat response of +/- 10% from 80 KeV to 3 MeV. The anticipated dose response, calculated for the detector due to an equivalent concentration of Xe-133 in the discharge effluent from the steam line is 6.1 x 10-1 mr/hr/µCi/cc. Background correction for each monitor is achieved by 2-inch lead shielding to reduce background and provide collimation of the detector to respond to the gamma radiation from the steam line.

18. Condensate Polish Backwash Effluent Monitor, Channel RM-L11 A sample of the recirculated or batch discharge fluid from the Condensate Polish System backwash receiver tank is monitored for gross gamma activity by an off-line liquid radiation monitor, RM-L11 (see Figure 10.4-7a). Sensitivity is approximately 1 x 10-6 µCi/cc referenced to Cs-137. Indication, alarms, and recording are provided locally.

Interlocks are provided to trip the backwash pumps upon a high activity alarm. The activity of the fluid is evaluated prior to discharge. Alarm is also provided at the local Condensate Polish System control panel. As agreed with the NRC Staff, this monitor is required to be installed prior to startup after the first refueling. 11.4.3 SAMPLING The requirements of General Design Criterion 64 are satisfied with respect to effluent discharge paths as stated below. Radiation monitoring is accomplished using the Radiation Monitoring System and laboratory analysis of samples in accordance with Regulatory Guide 1.21 (see Appendix 3A). Table 11.4-2 presents normal and potential discharge paths from the plant and lists the monitoring and analyses performed. The frequency of sampling is increased should abnormal levels of activity be detected. Effluent monitors RM-L5, RM-L11 and RM-A10 provide backup control of the administrative release of liquid and gaseous waste. These batch releases are analyzed and evaluated prior to discharge. Main plant vent exhaust monitor RM-A3 provides additional backup to the Waste Gas discharge monitor RM-A10. 02-01 02-01

11.4-11 Reformatted November 2011 Ventilation discharge effluent monitors RM-A3 (main plant vent exhaust) and RM-A4 (Reactor Building purge exhaust) provide removable particulate filters and charcoal cartridges for laboratory analyses which allow evaluation of normal and/or abnormal releases. Continuous effluent discharge radiation monitor (RM-A3, RM-A9 and RM-L7) readings of gas or liquid activity would be affected by abnormal background directly following a LOCA and would, at that time, indicate a higher gas or liquid activity than was actually being discharged. This would result only in the possibility of premature actuation of interlocks or alarms. Use of portable instrumentation, sample analysis, and appropriate emergency procedures, as well as the radiation monitors, permits evaluation of the plant discharge to the environment under accident conditions. Liquid effluent monitor RM-L7 (Nuclear Blowdown waste) diverts the process discharge upon detection of excessive gross gamma activity. A manual sample of the diverted discharge fluid is analyzed in the laboratory for evaluation of the isotopic content. Primary to secondary leakage is detected by the condenser exhaust monitor, RM-A9. Additional backup information is provided by the steam generator liquid blowdown monitor, RM-L3. Subsequent analysis of liquid samples in the laboratory provide verification. In the event of simultaneous primary to secondary and turbine leaks, the Turbine Building Ventilation System and Floor Drain System become potential discharge points. Portable monitors and manual samples are used to determine the rate of release. Liquid radiation monitor, RM-L8, monitors turbine room sump discharge and stops the sump pumps upon occurrence of a high radiation alarm. The ventilation exhaust from the Intermediate Building is not considered a potential discharge point due to a lack of potential sources within this building. Routine Health Physics surveillance with portable monitors and analysis of particulate portable sampler filters are used for verification. General Design Criterion 60 is satisfied by use of interlock circuits which are illustrated by Figure 11.4-1. In addition, alarms resulting from detection of high activity alert the operator to the need for appropriate action. Requirements of General Design Criterion 63, with respect to the monitoring of radiation levels in radioactive waste process systems, are satisfied by use of the effluent and process radiation monitoring instrumentation, area gamma monitoring, Health Physics surveillance with portable instruments, and laboratory analysis of process samples.

11.4-12 Reformatted November 2011 11.4.4 INSERVICE INSPECTION, CALIBRATION, AND MAINTENANCE The analyses of radioactive effluents are performed in accordance with requirements set forth in Regulatory Guide 1.21 (see Appendix 3A). Inplant laboratory facilities for these analyses include a separate sample room, radiochemistry laboratory, and count room located in the Control Building. Appropriate equipment for the safe handling of samples and standards is provided. Count room equipment includes the following:

1.

Germanium detectors

2.

Liquid scintillation counter

3.

Gross alpha counter

4.

Appropriate shielding for detection devices

5.

Multichannel analyzers

6.

Data reduction and storage system Detailed written analytical laboratory procedures are provided to assure compliance with sensitivity, accuracy, and reproducibility requirements within the capabilities of commercially available equipment. The schedule for routine calibration of laboratory equipment used for effluent sample analysis is detailed in such a manner as to assure the continuing accuracy of measurements. Calibration performance for the various equipment is closely monitored to determine if malfunction due to procedural difficulties or minimal equipment failure have occurred and, where applicable, whether equipment maintenance is required. Functional checks of equipment performance are performed on a more frequent basis than is calibration in accordance with written laboratory procedures. These checks use materials with the accuracy, stability, and range (though not necessarily traceable to the National Institute of Standards and Technology) necessary to assure a timely indication of equipment operational status. Maintenance or readjustment of the laboratory counting equipment is followed by appropriate recalibration. Laboratory counting systems used in the analysis of effluent samples are calibrated using specific standards consistent with the particular effluent sample geometry. 02-01

11.4-13 Reformatted November 2011 The radioactive material in these standards is either certified by the National Institute of Standards and Technology or has been calibrated against standard reference material certified by the National Institute of Standards and Technology. Calibration standards have the accuracy, stability, and range required for the intended use. Detailed written procedures for calibration of laboratory counting equipment and preparation of standards are provided to assure the continuing adequacy of effluent sample measurements. Intercomparison and interlaboratory spiked samples and analyses of such samples provides further assurance of system measurement accuracy. Equipment ancillary to qualification and quantification of effluent samples is also calibrated using detailed written procedures such as those called for in Regulatory Guides 1.21 and 1.23 (see Appendix 3A) to assure analytical accuracy. Maintenance of laboratory equipment used in the analysis of effluent samples is performed in accordance with procedures recommended by the equipment manufacturers. Laboratory equipment availability is maximized as follows:

1.

The incorporation of modular designs where appropriate

2.

Provision of spare parts

3.

Interchangeability of certain components within different laboratory counting systems when consistent with design considerations. The process sampling system is described in Section 9.3.2. 11.

4.5 REFERENCES

1.

Title 10, Code of Federal Regulations, Part 20, Appendix B, "Concentrations in Air and Water above Natural Background." 02-01

11.4-14 Reformatted November 2011 TABLE 11.4-1 PROCESS AND EFFLUENT RADIOLOGICAL MONITORS Monitor Function Detector Power RM-L1 Primary Coolant Letdown Liquid process monitor, detection of gross activity buildup in reactor coolant. NaI and G-M; Gamma 10-3 to 102 µCi/cc & 10-1 to 103 µCi/cc, Cs-137 Bus A RM-L2A Component Cooling Water Closed Loop A Liquid process monitor, detection of leakage into component cooling water system. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus A RM-L2B Component Cooling Water Closed Loop B Liquid process monitor, detection of leakage into component cooling water system. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B RM-L3 Steam Generator Blowdown Liquid process monitor, detection of leakage into steam generators. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus A RM-L4 Spent Fuel Cooling Water Liquid process monitor, detection of pool water contamination. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B RM-L5 Liquid Waste Effluent Liquid effluent monitor, terminates discharge of waste liquid on detection of excessive activity. Backup to predischarge analysis of liquid waste tank contents. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B -NO LONGER IN SERVICE-RM-L6 Boron Recycle System Liquid process monitor, diverts discharge on detection of excess activity. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B 02-01 RN 07-037

11.4-15 Reformatted November 2011 TABLE 11.4-1 (Continued) PROCESS AND EFFLUENT RADIOLOGICAL MONITORS Monitor Function Detector Power RM-L7 Nuclear Blowdown Waste Effluent Liquid effluent monitor, terminates discharge and diverts flow upon detection of excess activity. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B RM-L8 Turbine Room Sump Monitor Liquid effluent monitor, stops sump pumps upon detection of excessive activity. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 120 Volt RM-L9 Liquid Waste Discharge Effluent liquid monitor, terminates liquid waste discharge to penstocks upon detection of high activity. Provides backup to radiation monitors RM-L5 and RM-L7. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus A RM-L10 Steam Generator Blowdown Discharge Effluent liquid monitor, terminates steam generator blowdown discharge to circulating water system upon detection of high activity. Provides backup to radiation monitor RM-L3. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Cs-137 Bus B RM-L11 Condensate Polish Backwash Effluent Effluent liquid monitor, terminates backwash discharge pump operation upon detection of high activity. Provides backup to analysis of batch activity. NaI; Gamma 10-6 to 10-2 µCi/cc, Cs-137 120v-ac Local 02-01

11.4-16 Reformatted November 2011 TABLE 11.4-1 (Continued) PROCESS AND EFFLUENT RADIOLOGICAL MONITORS Monitor Function Detector Power RM-A3 Main Plant Vent Exhaust Effluent atmospheric monitor, detects excessive release from auxiliary building exhaust. Particulate - Beta 4.7 x 10-11 to 10-7 µCi/cc, Cs-137 Iodine - NaI; Gamma 2 x 10-11 to 2 x 10-7 µCi/cc, I-131 Gas - Beta 2.6 x 10-6 to 2 x 10-2 µCi/cc, Kr-85 Bus A RM-A4 Reactor Building Purge Exhaust Effluent atmospheric monitor, terminates purge exhaust on detection of excess gas activity. Particulate - Beta 4.7 x 10-11 to 10-7 µCi/cc, Cs-137 Iodine - NaI; Gamma 2 x 10-11 to 2 x 10-7 µCi/cc, I-131 Gas - Beta 2 x 10-6 to 2 x 10-2 µCi/cc, Kr-85 Bus A RM-A9 Condenser Exhaust Effluent gas monitor, detects leakage into steam generator system. NaI; Gamma 2 x 10-6 to 2 x 10-2 µCi/cc, Xe-133 Bus B RM-A10 Waste Gas Discharge Effluent gas monitor, terminates discharge of waste gas on detection of excess activity. Backup to predischarge analysis of waste gas tank contents. Gamma 2 x 10-4 to 2 µCi/cc, Xe-133 Bus B RM-A13 Main Plant Vent Exhaust, High Range Gas Discharge Effluent High Range Gas Monitor Ion Chamber - Gamma Ref. 10-2 to 105 µCi/cc, Xe-133 (0.1 to 107 mr/hr) Bus A RM-A14 Purge Exhaust Effluent, High Range Effluent High Range Gas Monitor Ion Chamber - Gamma Ref. 10-2 to 105 µCi/cc, Xe-133 (0.1 to 107 mr/hr) Bus A RM-G19, A,B,C Steam Line High Range Gamma Monitor Ion Chamber - Gamma 0.1 to 107 mr/hr Bus B 02-01 RN 03-011

11.4-17 Reformatted November 2011 TABLE 11.4-2 DISCHARGE MONITORING AND ANALYSIS Normal Discharge Paths Type of Continuous Monitoring Samples for Isotopic Analysis Frequency of Sampling Main Plant Vent RM-A3 Particulate () RM-A3 Iodine () RM-A3 Gas () Fixed filter Charcoal cartridge Gas sample from gas decay tank (includes tritium) Weekly Reactor Building Purge RM-A4 Particulate () RM-A4 Iodine () RM-A4 Gas () Fixed filter Charcoal cartridge Gas sample from containment atmosphere (includes tritium) Before each purge and weekly during a continuous purge (e.g., refueling period) Liquid Waste Effluent Discharge RM-L5 Liquid () RM-L9 Liquid () Aliquot from holdup tank (includes tritium) Prior to batch release Waste Gas Discharge RM-A10 Gas () Gas sample from gas decay tank Prior to batch release 02-01

11.4-18 Reformatted November 2011 TABLE 11.4-2 (Continued) DISCHARGE MONITORING AND ANALYSIS Normal Discharge Paths Type of Continuous Monitoring Samples for Isotopic Analysis Frequency of Sampling Steam Generator Blowdown RM-L3 Liquid () and/or RM-L10 Liquid () Aliquot from blowdown/discharge Daily for composite and immediately following detection of a high radiation level Nuclear Blowdown Waste Effluent RM-L7 Liquid () Aliquot from Nuclear Blowdown Monitor Tank Prior to batch release and immediately following detection of a high radiation level Turbine Room Sump Discharge RM-L8 Liquid () Aliquot from sump Daily for composite and immediately following detection of a high radiation level Condensate Polish Backwash Effluent RM-L11 Liquid () Aliquot from Backwash tank Prior to discharge and following detection of a high radiation level 02-01

TURBINE ROOM SUMP ON TURBINE ROOM SUMP PUMPS MONITOR RM-L8 OFF XPP-104A,B XPP-105A,B COMPONENT COOLING WATER OP COMPONENT COOLING SURGE MONITOR RM-L2(A,B) CL TANK VENT IVV-7096-CC STEAM GENERATOR BLOWDOWN NOR BLOWDOWN TO HOLDUP TANK MONITOR RM-L3 DIV XVT-524, XVT-525, & XVT-541 LIQUID WASTE EFFLUENT OP LIQUID WASTE DISCHARGE MONITOR RM-L5 CL VALVE RCV-018 BORON RECYCLE SYSTEM -NO LONGER IN SERVICE-NOR BORON RECYCLE LIQUID TO MONITOR RM-L6 DIV EVAPORATOR RCV-016 NUCLEAR BLOWDOWN NOR BD EFFLUENT TO MONITOR TANK WASTE EFFLUENT RM-L7 DIV 6122 6121 LIQUID WASTE EFFLUENT OP LIQUID WASTE DISCHARGE MONITOR RM-L9 CL VALVE 6910 STEAM GENERATOR BLOWDOWN OP BLOWDOWN DISCHARGE EFFLUENT RM-L10 CL VALVE 547-BD CONDENSATE POLISHING BACKWASH EFFLUENT MONITOR RM-L11 CONTROL ROOM AIR SUPPLY MONITOR (GAS CHANNEL) RM-A1 MAIN PLANT VENT MONITOR (GAS CHANNEL) RM-A3 WASTE GAS DISCHARGE MONITOR RM-A10 REACTOR BUILDING MANIPULATOR CRANE AREA RM-G17A REACTOR BUILDING PURGE (GAS CHANNEL) RM-A4 REACTOR BUILDING MANIPULATOR CRANE AREA RM-G17B REACTOR BUILDING AIR SAMPLE (GAS CHANNEL) RM-A2 ON CONDENSATE POLISHING OFF BACKWASH PUMPS XPP-153A,B OFF CONTROL ROOM EMERGENCY FILTER ON FAN XFN-30A, 32A OFF CONTROL ROOM EMERGENCY FILTER ON FAN XFN-30B, 32B OP CONTROL ROOM OUTSIDE AIR CL SHUTOFF DAMPERS XDP-0018A,B, 19A,B OP CONTROL ROOM RELIEF & BYPASS CL DAMPERS XDP-21A,B, 22A,B OP WASTE GAS DISCHARGE VALVE CL NCV-014 OP REACTOR BUILDING PURGE VALVES CL A - CHANNEL XVB-1A, XVB-2A OP REACTOR BUILDING PURGE VALVES CL 6056, 6066 OP REACTOR BUILDING PURGE VALVES CL B - CHANNEL XVB-1B, XVB-2B OP REACTOR BUILDING PURGE VALVES CL 6057, 6067 SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLEAR STATION RN 07-037 November 2011 Radiation Monitoring System Interlocks Figure 11.4-1 Rev. 2 00-01 RN 07-037

ROOF EL. 511'-0" AUX I L1ARY BU ILD ING YENT~ 96" RII-Al3 -- 71 IC" [~ - ji ~ ~ 6" 7--~ " '[ RM-A14 PURGE VENT P-------

7. 05

REFERENCES:

E-921-810 E-922-209 SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLEAR STATION Location of High Range Effluent Monitors RM - A13 and RM* A14 Figure 11.4-2 Amendment 0 August 1984

11.5-1 Reformatted November 2011 11.5 SOLID WASTE SYSTEM 11.5.1 DESIGN OBJECTIVES The Solid Waste System is designed to package and/or solidify radioactive wastes for shipment to an approved offsite burial facility in accordance with applicable Department of Transportation (DOT), NRC, and State regulations. The system conforms to 10 CFR 20 and 10 CFR 50 requirements by providing shielding so that radiation exposure of operating personnel and the public is within acceptable limits. Solid waste packaging is accomplished in an area located on the ground floor (elevation 436) of the Auxiliary Building, a Seismic Category 1 structure. Design, fabrication, and testing of Solid Waste System components and piping is in accordance with ANSI B31.1 and other accepted standards referenced by ANSI B31.1. Additional onsite system tests will be performed using nonradioactive materials prior to commercial operation. Packaging and shipping conform to 49 CFR 171 through 49 CFR 178. Individual container shields and casks are used, when required, to maintain radiation levels within applicable radioactive materials regulations. 11.5.2 SYSTEM INPUTS Radioactive waste packaged includes:

1.

Spent resins.

2.

Used filter cartridges.

3.

Radioactive hardware.

4.

Compacted waste such as rags, paper, clothing, etc. Secondary side condensate polisher resin may also be handled by the Solid Waste System (refer to 10.4.6). Design quantities and activity levels of the various wastes are listed in Tables 11.5-1 through 11.5-4. RN 08-003

11.5-2 Reformatted November 2011 11.5.3 EQUIPMENT DESCRIPTION 11.5.3.1 Processing The input to the Solid Waste System consists of the contents of several radioactive waste storage tanks containing primary spent resins, reactor grade demineralizer spent resins, non-reactor grade demineralizer spent resins, and nuclear blowdown spent resins and the associated valves, piping, and pumps. These components are located at elevation 412 and 447 in the Auxiliary Building. Radwaste solidification when required is accomplished using approved equipment and process control program. Liquid waste contained in the reactor grade and non-reactor grade demineralizer is recirculated using their respective pumps and a sample is taken. This sample is used in the Process Control Program to determine pH adjustment, waste/binder ratio, and for the purpose of test solidification. Liquid waste is transferred to the fill head and into the liner located in the solidification area. Primary and Secondary spent resins are transferred from their respective holdup tanks to either a disposable liner in the solidification area or a liner in the truck bay. A process shield or a DOT cask may be used when activity or exposure dictates. The resins may then be either solidified or dewatered for shipment. Dewater return is routed to the Excess Liquid Waste Hold up Tank, the Decon Pit Collection Tank, or the Floor Drain Tank. Each waste transfer is immediately followed by a flush operation of the waste transfer piping and the internals of the fill head when used. Labyrinth shield walls separate the drumming station control room, the piping process skid cubicle, and the container fill area from one another. Equipment is described in more detail in Section 11.5.3.2. 11.5.3.2 Equipment The equipment comprising the Solid Waste System is described in Sections 11.5.3.2.1 through 11.5.3.2.6. Table 11.5-5 provides equipment design parameters. Table 11.5-7 shows how the utility equipment, components, structures, and services that interface with the vendor-supplied solidification system comply with the applicable criteria of Regulatory Guide 1.143, Rev. 1, October 1979, "Design Guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water-Cooled Nuclear Power Plants" and Branch Technical Position ETSB 11-3, Rev. 2, July 1981, "Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Plants." 02-01 RN 08-003 RN 08-003

11.5-3 Reformatted November 2011 11.5.3.2.1 Waste Storage Tanks and Pumps Tanks containing radioactive waste and wetted parts of pumps are fabricated from stainless steel, type 304, except as noted. The primary spent resin storage tank and pump are described in Section 11.2. Other radwaste tanks and pumps are the Nuclear Blowdown spent resin storage tank and Nuclear Blowdown spent resin storage tank pump. The spent resin storage tank and the associated pump are located in the Auxiliary Building at floor elevation 412. The waste hold-up tank and the associated pump are located at floor elevation 374. The resin transfer lines are sloped to avoid low points in the piping. Also, 5 diameter pipe bends are used from the resin tanks to the solidification area. These long sweeping bends are necessary to avoid plugging that could occur in the inner wall of a bend. Instrument lines are kept to a minimum to avoid dead legs. All lines will be flushed with makeup water prior to any maintenance activities. The reactor grade and non-reactor grade demineralizers located on elevation 447 are flushed directly to liners in the truck bay at elevation 436. The resin and waste transfer lines are 2" nominal diameter. This size is sufficient for the desired flow rates without causing excessively large pressure drops that could result in line plugging. 11.5.3.2.2 Instrumentation and Controls The system uses temperature, flow, pressure, and level instruments to monitor and/or control the process located throughout the system.

1.

Inplant Control Panel The inplant control panel is a standard enclosure of NEMA 12 construction. The inplant control panel provides full operational control for resin and liquid waste transfer operations to the waste processing equipment. It is also used to interface with equipment for dewater return and for flushing operations. This panel contains switches and lights for valves and pumps which facilitate these operations. RN 08-003 RN 08-003 RN 08-003

11.5-4 Reformatted November 2011

2.

Power Panel The power panel provides power for the operation of the various pump motors and valve motors in the system.

3.

Radiation Monitoring Radiation monitoring is provided by portable instrumentation and/or semi-permanent radiation monitoring equipment. 11.5.3.2.3 Waste Containers and Shielding All wastes are packaged in containers which meet DOT/NRC requirements. The containers used for solidification and resin dewatering provide appropriate connections for processing. Higher activity wastes will be processed with the liner already in a cask located in the truck bay access. A double lid cask top will be used to limit exposure in this area. The main part of the lid shields the entire top of the cask except the immediate area required for the fill head. After the process is complete, the fill head is removed and the secondary cask lid is installed. Final closure of all radwaste liners may be accomplished by remote handling equipment, if required. High integrity containers are remotely sealed by a hand manipulated device which screws the closure cap into the liner. 11.5.3.2.4 Contamination Control Facilities An adjacent decontamination area is provided for cleanup of contaminated containers. Exposed surfaces of filled containers or casks are surveyed by the Health Physics Group to identify the presence of removable radioactive contamination prior to transfer to storage or shipment. Containers are decontaminated in the adjacent decontamination area, if required. 02-01 RN 08-003 RN 08-003

11.5-5 Reformatted November 2011 11.5.3.2.5 Handling Equipment Equipment used for handling waste containers and equipment within the radwaste area and for truck loading includes the following:

1.

One ton jib crane.

2.

Three ton jib crane.

3.

Twenty ton hoist and monorail.

4.

Three ton bridge crane.

5.

Ten ton bridge crane. The one ton jib crane is located on a wall above the truck access floor at elevation 455. It is used for hoisting chemicals and equipment from the truck access area to the mezzanine floor. It has a lift of 23 feet at a speed of 22 ft/min. The three ton jib crane is located on a wall above the solidification area. It is used to handle the vendors fill head and other equipment. It has a 23 foot lift at a speed of 11 ft/min. The twenty ton hoist and monorail is used to load the containers on a truck for transport to a burial site. It has a lift of 27 feet at a speed of 10 ft/min. The three ton bridge crane is located over the radioactive filter area at floor elevation 463. It is used in conjunction with a 3-1/2 inch thick lead filter transfer cask to remove spent radioactive filter cartridges from the filter housings located in concrete cubicles on the floor below at elevation 452-6". The trolley has a transfer mechanism which permits the hoist and the cask to engage a monorail which extends over the radwaste fill area. A hatch at floor elevation 463 is removed and the hoist lowers the cask to the radwaste area at floor elevation 436. It has a lift of 52 feet at a speed of 22 ft/min and a trolley speed of 65 ft/min. The ten ton bridge crane is located in the hot machine shop. It is chiefly used to service the machine shop. However, a portion of the floor area in the machine shop is partitioned from the rest of the shop for storage of unused containers, 55 gallon drums, pallets, etc. The storage area is also serviced by this crane. The hoist has a lift of 24 feet at either 7 or 20 ft/min. The trolley has a speed of either 32-1/2 or 65 ft/min. RN 08-003 02-01 RN 08-003 02-01

11.5-6 Reformatted November 2011 11.5.3.2.6 Waste Compactors One waste compactors is available for use and is capable of handling 90 cubic feet of waste in one compaction. 11.5.3.2.6.1 Normal Use Compactor An electromechanical compactor with a compressive force capacity of 100,500 lbs. is used to compact dry wastes into a 90 cubic foot container. During compaction the container is completely enclosed. A self-contained HEPA filter and blower system filters the air released in the compaction process before any air is discharged to the Fuel Handling Building Charcoal Exhaust System. An electrical interlock prevents the operation of the compactor if the door which encloses the container is not completely closed. This prevents injury to the operator and unfiltered air from escaping to the Hot Machine Shop atmosphere. 11.5.3.2.6.2 Standby Compactor -No Longer In Service-11.5.3.2.7 Truck Loading Features A wall penetration is provided between the fill and truck access area to fill directly to containers on a truck. This penetration is located in the shielded cubicle of the solidification area such that exposure in the truck access is limited. 11.5.4 EXPECTED VOLUMES The expected annual volume of solid radioactive wastes together with the associated Curie content of principal nuclides to be processed are described in Sections 11.5.4.1 through 11.5.4.4. 11.5.4.1 Activity Levels The activity level of the wastes generated directly from operation of the Nuclear Steam Supply System is based upon reactor plant operation at a base load factor of 80 percent power with reactor coolant activity levels determined on the basis of fission product diffusion through cladding defects in 0.12 percent of the fuel rods. The system is conservatively designed to accommodate solid wastes generated by plant operations with up to 1 percent fuel defects. Source term data used for system design are presented in Section 11.1. RN 08-003 02-01 02-01 RN 08-003

11.5-7 Reformatted November 2011 Table 11.5-1 lists the demineralizer resin volumes and expected volumes replaced on an average yearly basis. Table 11.5-2 presents a summary of the anticipated total solid radioactive waste generated per year. The expected activity of the solid waste at time of shipment is dependent upon the decay storage time. An isotopic breakdown of spent resin activities is presented in Table 11.5-3. The maximum activity of expended filter cartridges is given in Table 11.5-4. The associated Curie content and volume of waste shipped from a number of Westinghouse designed operating reactors is given in Table 11.5-6 for each year from 1971 through 1974. 11.5.4.2 Processed Wastes In the case of primary spent resins, the Curie content totals approximately 1390 Ci/yr. Nuclear Blowdown System spent resins are estimated, for design purposes, to account for approximately 4.0 Curies per year. 11.5.4.3 Filter Cartridges The volume of expended filter cartridges processed for disposal by the Solid Waste System is based upon the expected filter cartridge change frequency for potentially radioactive filters. The assumption is made that filters processing reactor coolant will require cartridge renewal due to excessive radiation levels or high P. The maximum expected activity of expended filter cartridges shipped from the site is conservatively based upon a shielding criteria of a maximum contact dose rate. 11.5.4.4 Miscellaneous Solid Wastes The annual volume of miscellaneous solid wastes processed by the solid waste compactor is assumed to amount to 85 containers of 90 cubic feet of compacted refuse. The wastes consist of rags, coveralls, ventilation filter cartridges, and various other potentially contaminated refuse. This refuse is normally classified as low specific activity. 11.5.5 PACKAGING 11.5.5.1 Evaporator Bottoms and Chemical Samples -No Longer In Service-11.5.5.2 Spent Resin Resin in a demineralizer is considered spent when its decontamination factor falls below a permissible level. The spent resin, from demineralizers in the primary system is stored in a 350 ft3 storage tank. The spent resin from nuclear blowdown demineralizers in the secondary system is stored in a 600 ft3 Nuclear Blowdown System storage tank. The resin stored in the primary system is normally allowed to decay for a period of up to several months. RN 08-003 RN 08-003 RN 08-003

11.5-8 Reformatted November 2011 The reactor grade water demineralizers normally contain 28 ft3 of resin per demineralizer in each of the three demineralizers. The non-reactor grade water demineralizers contain 28 ft3 of resin per demineralizer in each of the seven demineralizers. These ten demineralizers are located on elevation 447 of the auxiliary building and are sluceable directly to an available liner in the truck bay. These demineralizers are flushed when the decontamination factor falls below a permissible level. When a sufficient quantity of resin has accumulated and decayed, the resin is sampled, analyzed for isotopic constituents and activities, and packaged. Prior to packaging, nitrogen is sparged into the tank to form a slurry which is transferred to the liner by nitrogen cover gas pressure. Dewatering of the resin is accomplished using dewatering equipment with the water being returned either to the Excess Liquid Waste Holdup Tank, the Decon Pit Collection Tank, or the Floor Drain Tank. Spent resin may also be solidified. The radiation level of the primary resin is expected to require use of a 4 inch lead shield on some occasions. The radiation level of the Nuclear Blowdown System resin is expected to require not more than 1-1/2 inch lead shield. The primary spent resin storage tank has a two inch discharge line located along the tank center line, protruding from its top and extending to within 3 inches above the dished bottom. In preparation for packaging, the discharge valve is opened and the center discharge tube cleared by backflush with a burst of flush water from the Reactor Makeup Water System. Pressure to 100 psig is available, if required. Flush water may continue to be added if needed to obtain a reasonable slurry. The discharge valve is then closed. Loosening of the resin is achieved by introducing nitrogen through seven spargers at the tank bottom. Resin sluice water can be recirculated through the spargers to loosen the resin if desired. When the nitrogen pressure increases to that required for resin transfer, the resin discharge valve is opened. Nitrogen continues to bubble through the resin bed to maintain a gas pressure for transfer of the resin until the liner reaches the full level. The liner vent during this operation is directed to the plant vent or to a portable ventilation unit. Similarly, the reactor grade and non-reactor grade demineralizers are flushed directly to a truck bay liner. The Nuclear Blowdown System spent resin storage tank is discharged by use of a procedure similar to that used for the primary spent resin storage tank. The resin slurry is discharged through a 2 inch nozzle located at the tank bottom. Nitrogen gas is bubbled into the tank bottom connection to loosen and mix the resin and pressurize the tank. When the tank gas pressure increases to that required for resin transfer, the resin slurry discharge valve is opened. Operation of both tanks from this point is similar. RN 08-003 RN 08-003 RN 08-003

11.5-9 Reformatted November 2011 When the transfer is complete, the resin discharge and nitrogen supply valves are closed and a tank vent valve is opened to discharge the nitrogen cover gas from the storage tank. In addition, the flush water supply valve is opened to backflush and forward flush and decontaminate the resin transfer line. A flow diagram of the primary resin system is provided by Figure 11.2-2, Sheet 3. Figure 10.4-15 describes the nuclear blowdown resin storage. A flow of approximately 40 gpm is required to transfer the resin slurry to the liner in the radwaste area. It is anticipated that approximately 1300 std ft3 and 2200 std ft3 of nitrogen gas will be the maximum required for each resin transfer operation from the 350 ft3 primary resin storage tank and 600 ft3 nuclear blowdown resin storage tank, respectively. The Nitrogen System is set to supply nitrogen to the resin storage tanks at a pressure of 100 psig, if needed. The resin storage tanks are designed for 150 psig. Relief valves on the primary and nuclear blowdown resin storage tanks are set to relieve at 100 psig and 150 psig, respectively. The primary resin storage tank relieves to the waste holdup tank. The Nuclear Blowdown resin storage tank relieves to the Nuclear Blowdown System reservoir by way of an open drain. 11.5.5.3 Filter Disposal Filters are of the disposable cartridge type contained in housings having hinged tops. They are replaced when surface dose rate or pressure drop exceeds established levels. Filters which are potentially radioactive are located in individual cubicles in an area close to the drumming station area. If the radiation level of the cartridge requires shielding during removal, a concrete plug in the floor above the housing is removed and another plug with a hole in it is placed in the stepped opening. A filter cask with 3 1/2" lead encased in stainless steel is placed over the hole. The filter housing is opened and the cartridge is drawn into the cask by the use of special tools having extension rods. Once the filter is in place, the cask bottom is closed and the tops installed. The cask is then transported by an overhead crane to a hatch at floor elevation 463 of the Auxiliary Building. This hatch is located above the drumming station area on the floor below. The cask is lowered into the drumming station area. Storage and disposal of all filters is within either high integrity containers or DOT approved containers depending on the specific activity of the filters. For filters requiring shielding, the container is stored in a shielded cask. The filter transfer cask is positioned over a small opening in the shield cask, the bottom slide is pulled open, and the filter is lowered into the shielded container. In this manner, the handling of highly contaminated filters is kept to a minimum. 02-01 RN 08-003

11.5-10 Reformatted November 2011 11.5.5.4 Radioactive Hardware Radioactive hardware can consist of damaged or used equipment or instruments, which due to geometry or materials of fabrication, cannot be readily decontaminated. Such material is disposed of in much the same way as are filter cartridges or as compacted waste, depending upon radiation levels. 11.5.5.5 Compacted Wastes An electromechanical compactor provides 100,500 lbs. of compressive force for the compaction of compressible waste into 90 cubic foot containers. During compaction the container and compacting mechanism are enclosed and the enclosure is vented to the Fuel Handling Building Charcoal Exhaust System through a HEPA filter by a blower. The blower and filter are contained within the compactor. The blower is automatically operated when the door is closed; however, a manual switch is provided so the blower may be operated without compactor operation. The compactor will not operate unless the door is closed, protecting the operator from injury and preventing escape of unfiltered air to the atmosphere. 11.5.6 STORAGE Compactable waste, filled containers of compacted waste, and spent filter cartridges are stored in the shielded areas of the radwaste area or in a location determined by the Manager, Health Physics and Safety Services. Contaminated hardware and tools may also be stored in these rooms. Solidified waste, after solidification is complete, and dewatered resins, once dewatering is complete, may be shipped off-site for immediate burial at a licensed facility. Primary spent resins will normally have at least a one month decay period while being held in the spent resin storage tank. Secondary blowdown resins do not normally require a decay period. If solidified waste and/or dewatered resins require storage for any reason, they will be stored in the radiation control area outside the truck access on the concrete pavement or in a location determined by the Manager, Health Physics and Safety Services. Waste stored in the storage area will be shielded as required by portable shields and/or casks used for shipment. Storage areas for solidified waste, dewatered resins, and compacted waste are sufficient, based on the estimates presented in Section 11.5.4, to accommodate greater than 30 days waste generation. 02-01 02-01 RN 10-022 RN 08-003 RN 08-003 RN 10-022

11.5-11 Reformatted November 2011 11.5.7 SHIPMENT Shipment, in accordance with applicable regulations, is made as necessary--dependent upon operational considerations and storage area availability. The primary activity determination method will be to sample the waste stream (resins and liquid waste) during transfer to a process container and analyze the sample using the appropriate counting instrumentation. An isotopic determination is made of the radionuclides present and the activity of each. Summation of the individual activities is used to calculate the Curie content of the processed container. For cases where the primary method cannot be used, an alternate technique will be implemented. The alternate method entails using the dose rate of the packaged waste in order to calculate the Curie content. The calculation considers the waste characteristics, geometry of the waste package, characteristics of the container and solidification media (if applicable), and the average gamma energy. For spent cartridge filters, this alternate method will be used to determine the Curie content. The appropriate counting instrumentation is used to analyze samples taken from the process stream to identify radionuclides present and the average gamma energy. 11.5.8 POTENTIAL FOR RELEASES 11.5.8.1 Potential for Release during Container Filling The filling operation may be terminated via visual inspection using a remote monitor/television camera. Termination is accomplished by closing valves MOV-2 and MOV-5. There is no airborne release to the atmosphere in the fill areas. Air in the container and gas, if any, from the waste entering the container are vented to the building exhaust, through a local filter, or through a portable ventilation unit. Only one line feeds waste to the container. This is flushed with water as the final phase of the fill cycle. If leaks of any kind or spills are observed, the operation in progress can be immediately terminated. Any spill which may occur will be contained by permanent curbing in the solidification area. Except for the curb in the solidification area, there are no physical barriers in the immediate fill areas to contain spills. Spills from the shipping container would need to be drained to a specific location or container as determined by the type of material spilled. The floor surfaces have a special nonporous finish to permit decontamination of the surface, if required.

11.5-12 Reformatted November 2011 11.5.8.2 Potential for Release from Storage Tanks 11.5.8.2.1 Waste Evaporator Concentrates Tank -No Longer In Service-11.5.8.2.2 Chemical Drain Tank -No Longer In Service-11.5.8.2.3 Primary Spent Resin Storage Tank This tank contains only a negligible quantity of radioactive gases in the gas space. The gas is normally contained in the tank by a closed vent valve. This vent is ducted to the Auxiliary Building Exhaust System and is open only during transfer of resin from the demineralizers or at the conclusion of transferring resin from this tank to the radwaste packaging area. Overflow is not anticipated since primary spent resin storage tank capacity is sufficient to accommodate at least 60 days waste generation under normal plant operating conditions. Overflow protection is provided by a high level alarm at the Solid Waste System control panel. Excess water can either be pumped or drained to the waste holdup tank. Overflow, if it occurs, is to the waste holdup tank through a relief valve. The tank is enclosed within a concrete cubicle with entrance from an overhead shield slab. Any leakage is directed to the floor drain tank through a floor drain. 11.5.8.2.4 Nuclear Blowdown Spent Resin Storage Tank This tank contains only trace amounts of radioactive gas. The gas is normally contained in the tank by a closed vent valve. The tank is vented to the cubicle, which is serviced by the building exhaust system, only during transfer of resin from the demineralizers or at the conclusion of resin transfer from this tank to the radwaste packaging area. Overflow is not anticipated since the nuclear blowdown spent resin storage tank capacity is sufficient to accommodate at least 30 days waste generation under normal plant operating conditions. Overflow protection is provided by a high level alarm at the Solid Waste System control panel. Excess water can either be pumped or drained to the Nuclear Blowdown System reservoir. Overflow, if it occurs, is to the Nuclear Blowdown System reservoir through a relief valve. The tank is enclosed within a concrete cubicle with entrance from an overhead shield slab. Any leakage is directed to the Nuclear Blowdown System reservoir through a floor drain. RN 08-003

11.5-13 Reformatted November 2011 TABLE 11.5-1 SPENT RESIN VOLUMES Demineralizer Number Resin Volume per Bed (ft3) Expected Average Resin Volume Replaced/year (ft3) CVCS Mixed Bed 2 30 60 (minimum) CVCS Cation Bed 1 20 20 (minimum) Recycle Evaporator Feed 2 30 30 (minimum) Boron Thermal Regeneration 4 70 70 Waste Monitor Tank 1 30 540 Nuclear Blowdown Primary 2 150 450 Polishing 2 90 270 Spent Fuel Pool 1 54 54 Excess Liquid Waste 2 30 540 Reactor Grade Process Water 3 28 28 Non-Reactor Grade Process Water 7 28 56 02-01 RN 08-003 RN 08-003

11.5-14 Reformatted November 2011 TABLE 11.5-2 ANTICIPATED TOTAL SOLID WASTE GENERATED PER YEAR VOLUME (FT3) (1) Primary Spent Resins 1340 Reactor Grade Demineralizer Spent Resins 28 Non-Reactor Grade Demineralizer Spent Resins 56 Dry Waste (Compacted) 5000 Nuclear Blowdown Spent Resins 500 Spent Filters Primary 320 Nuclear Blowdown 160 Totals 7404 (1) Before solidification. RN 08-003 RN 08-003

11.5-15 Reformatted November 2011 TABLE 11.5-3 MAXIMUM EXPECTED CONCENTRATIONS OF WASTE TO BE PACKAGED (1) Isotope Spent Resin Activity (µ/cc) Cr-51 1.9 x 101 Mn-54 1.9 x 101 Fe-55 1.3 x 102 Fe-59 1.6 x 101 Co-58 3.9 x 102 Co-60 1.7 x 102 Rb-86 3.2 x 10-1 Sr-89 6.2 Sr-90 8.7 x 10-1 I-131 7.9 x 102 Cs-134 1.1 x 103 Cs-136 3.5 x 101 Cs-137 9.0 x 102 Ba-140 1.0 TOTAL 3.6 x 103 (1) This table is based on N237 Standard. RN 08-003

11.5-16 Reformatted November 2011 TABLE 11.5-4 MAXIMUM EXPECTED ACTIVITIES OF EXPENDED FILTER CARTRIDGE Activity Per Cartridge (Ci) Filter Number of Filters Co-58 Co-60 Cs-134 Cs-137 Others Contact Dose Rate (R/Hr) (1) Reactor Coolant 1 4.0 2.8 2.9 2.0 500.0 Seal Water Injection 2 5.4 x 10-1 3.9 x 10-1 3.8 x 10-1 2.7 x 10-1 100.0 Seal Water Return 1 8.2 x 10-1 4.7 x 10-1 5.7 x 10-1 4.1 x 10-1 100.0 Recycle Evaporator Feed 1 8.2 x 10-1 4.7 x 10-1 5.7 x 10-1 4.1 x 10-1 100.0 Waste Monitor Tank 1 8.1 x 10-1 4.6 x 10-1 5.8 x 10-1 4.2 x 10-1 100.0 Spent Fuel Pool 2 8.2 x 10-1 4.7 x 10-1 5.7 x 10-1 4.1 x 10-1 100.0 Spent Fuel Pool Skimmer 1 8.2 x 10-1 4.7 x 10-1 5.7 x 10-1 4.1 x 10-1 100.0 Spent Resin Sluice 1 8.1 x 10-1 4.6 x 10-1 5.8 x 10-1 4.2 x 10-1 100.0 Floor Drain Tank 1 8.1 x 10-1 4.6 x 10-1 5.8 x 10-1 4.2 x 10-1 100.0 Nuclear Blowdown

a. Demineralizer Inlet 1

8.8 x 10-3 8.8 x 10-3 7.3 x 10-3 2.1 x 10-2 1.111

b. Demineralizer Outlet 1

1.1 x 10-2 2.6 x 10-3 4.8 x 10-2 1.3 x 10-1 3.1

c. Spent Resin Sluice 1

4.4 x 10-3 4.0 x 10-3 3.4 x 10-3 1.0 x 10-2 0.5 (1) Unshielded surface dose rate. 02-01 02-01 RN 08-003 02-01

11.5-17 Reformatted November 2011 TABLE 11.5-5 SOLID WASTE SYSTEM EQUIPMENT DESIGN PARAMETERS Waste Evaporator Concentrates Tank - No Longer In Service - Waste Evaporator Concentrates Pump (XPP0052) - No Longer In Service - Nuclear Blowdown Spent Resin Storage Tank Quantity 1 Volume, ft3 600 Type Vertical Design Pressure, psig 150 Nuclear Blowdown Spent Resin Storage Tank Sluice Pump (XPP0110) Quantity 1 Type Canned Centrifugal Design Capacity, gpm 150 Shutoff Head, ft 249 Design Pressure, psig 150 Design Operating Temperature, °F 100 - 135 Design Temperature, °F 140 02-01 RN 08-003

11.5-18 Reformatted November 2011 TABLE 11.5-6 SOLID RADIOACTIVE WASTE PROCESSED FROM WESTINGHOUSE DESIGNED OPERATING REACTORS Ci (Ft3) Plant 1971 1972 1973 1974 Connecticut Yankee 2.7x102 (2.2x103) 4.0x103 (3.9x103) 5.7x102 (5.6x103) 9.4x102 (7.1x103) Yankee Rowe 2.9 (1.1x103) 2.3 (7.8x103) 2.9 (4.2x102) 1.3x102 (7.8x103) San Onofre 1.2 (7.8x102) 8.0x101 (3.9x103) 3.8x102 (3.9x103) 2.3x102 (2.4x103) Robert E. Ginna 4.7x101 (2.5x104) 1.4x103 (1.3x103) 6.0x102 (7.1x103) 6.1x102 (9.9x103) Note: Annual Average Per Plant, Ci(ft3) = 5.8 x 102 (5.6 x 103) 02-01

11.5-19 Reformatted November 2011 TABLE 11.5-7 VALVES (GAI); VALVES (WESTINGHOUSE); EQUIPMENT; PIPING (GAI); PIPING (WESTINGHOUSE) TYPE DESIGN AND FABRICATION INSPECTION AND TESTING MATERIALS

  • VALVES (GAI)

Diaphragm (Dia.) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/74) Check (H20) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/74) Dia. (auto.) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/74) Plug (auto.) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (3/73) Check ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (2/74) Check (N2) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/74) Dia. (N2) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/74) Dia. (Control N2) ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (6/75) VALVES (WESTINGHOUSE) Dia. ANSI B16.5 (1968) ASME B&PV Sec. III ANSI B16.5 (1968) ASTM (1/73) Dia. (auto) ANSI B16.5 (1968) ASME B&PV Sec. III ANSI B16.5 (1968) ASTM (1/73) Check ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (11/71) Globe ANSI B16.5 (1968) ANSI B16.5 (1968) ASTM (11/71)

  • Dates listed under Materials refer to specification issue dates.

The ASTM Standard applicable to these dates apply. EQUIPMENT EQUIPMENT SPEC. ISSUE DATE

  • DESIGN AND FABRICATION INSPECTION AND TESTING MATERIALS Tank (PSRS) 6/72 ASME B&PV Sec. III ASME B&PV Sec. III ASME B&PV Sec. III Tank (NBSRS) 6/75 ASME Sec. VIII ASME Sec. VIII ASTM Tank (PSR) 11/71 ASME Sec. III, Hydraulic Inst.

Standard (HIS) ASME Sec. III, HIS ASME Sec. III PUMP (NBSR) 1/74 HIS HIS ASTM Filters 7/72 ASME Sec. VIII ASME Sec. VIII ASTM 02-01 02-01 02-01 RN 08-003 RN 08-003

11.5-20 Reformatted November 2011 TABLE 11.5-7 (continued) VALVES (GAI); VALVES (WESTINGHOUSE); EQUIPMENT; PIPING (GAI); PIPING (WESTINGHOUSE) EQUIPMENT SPEC. ISSUE DATE* DESIGN AND FABRICATION INSPECTION AND TESTING MATERIALS PIPING (GAI) Waste, H20 10/73 ANSI B31.10 ANSI B31.10 ASTM N2, Air 10/73 ANSI B31.10 ANSI B31.10 ASTM PIPING (WESTINGHOUSE) Waste, N2, 8/71 ANSI B31.10 ANSI B31.10 ASTM H20 NOTES: CD = Chemical Drain - No Longer In Service - PSRS = Primary Spent Resin Storage NBSRS = Nuclear Blowdown Spent Resin Storage WEC = Waste Evaporator Concentrates - No Longer In Service - B&PV = Boiler and Pressure Vessel Code 02-01 02-01 02-01 RN 08-003

Figure 11.5-1 Rev. 18 - Deleted by RN08-003, January, 2011

11.6-1 AMENDMENT 97-01 AUGUST 1997 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM The offsite Radiological Environmental Monitoring Program is described in The Offsite Dose Calculation Manual.}}

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