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Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 11.0, Radioactive Waste Management
	ML18221A181
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Site:	Summer
Issue date:	05/31/2018
From:	; South Carolina Electric & Gas Co
To:	; Office of Nuclear Reactor Regulation
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ML18221A142	List: ML18221A134; ML18221A145; ML18221A147; ML18221A151; ML18221A152; ML18221A153; ML18221A154; ML18221A155; ML18221A156; ML18221A157; ML18221A158; ML18221A159; ML18221A160; ML18221A161; ML18221A162; ML18221A165; ML18221A166; ML18221A167; ML18221A168; ML18221A169; ML18221A171; ML18221A172; ML18221A175; ML18221A178; ML18221A180; ML18221A181; ML18221A182; ML18221A206; ML18221A210; ML18221A211; ML18221A212; ML18221A213; ML18221A214; ML18221A215; RC-18-0080, Updated Final Safety Analysis Report, Chapter 9, Figures 9.5-1 Through 9.5-11 - Redacted; RC-18-0080, Virgil C. Summer Nuclear Station Unit 1 Submittal of Updated Final Safety Analysis Report, Updated Through May 31, 2018, Cover Sheet; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Approved FSAR Revision Notices; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 10.0, Steam and Power Conversion System; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 11.0, Radioactive Waste Management; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 12.0, Radiation Protection; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 13, Conduct of Operations; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 14, Initial Tests and Operation; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 15, Accident Analyses; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 16, Technical Specifications; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 17, Quality Assurance; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 18, Aging Management Programs and Activities; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 3.0, Design of Structures, Components, Equipment and Systems; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 4.0, Reactor; RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 5.0, Reactor Coolant System; ... further results
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{{#Wiki_filter:11.1-1Reformatted PerAmendment 99-0111.0RADIOACTIVE WASTE MANAGEMENT11.1SOURCE TERMSThe 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 aconventional fuel clad defect model. This conservative model serves as a basis forcalculations 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 basedon ANSI/ANS-18.1-1984.11.1.1RADIOACTIVITIES IN SYSTEMS AND COMPONENTS,CONSERVATIVE MODEL11.1.1.1Reactor Coolant ActivityThe 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 areassumed 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 followingdifferential equations:For parent nuclides in the coolant: dN c dt N c i iRN M D Q M DF DF iF i C ii L C ii i1(11.1-1 ) 99-01 99-01 11.1-2Reformatted PerAmendment 99-01For daughter nuclides in the coolant: dN c dt N c N c j j i jj jj j jRN M fD Q M DF DF F j C ii L C1(11.1-2 )Where: N C=Concentration of nuclide in the reactor coolant (atoms/gram) N F=Inventory of nuclide in the fuel (atoms)t=Operating time(seconds)R=Nuclide release coefficient (1/sec) = F vF=Fraction of fuel rods with defective claddingv=Fission product escape rate coefficient (1/sec) M C=Mass of reactor coolant (grams)=Nuclide decay constant (1/sec)DF=Nuclide demineralizer decontamination factor Q L=Purification or letdown mass flow rate (grams/sec)=Nuclide volume control tank stripping fractionf=Fraction of parent nuclide decay events that result in the formation of thedaughter nuclideD=Dilution coefficient for feed and bleed (1/sec) =B-OtDF*1 B O=Initial boron concentration (ppm)=Boron concentration reduction rate (ppm/sec)and wheresubscript 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 reductiondue to pressurizer operation and that the nuclide concentration in the volume control tank can be approximated by: N DF NVLC1*where N VL = concentration of nuclide in the volume control tank (atoms/gram) 99-01 99-01 11.1-3Reformatted PerAmendment 99-01The corrosion product activities in the Reactor Coolant System are based on measurements at operating reactors[1]. The reactor coolant inventories of corrosionproducts (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 thesilver-indium-cadmium control rods. The presence of Ag-110m has been noted in theprimary coolant or plant discharges at a few of the operating plants. Several mechanisms can contribute to the presence of this isotope; namely, fission productdecay 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 thecontrol rods. The control rods are designed to contain products formed in the absorbermaterial 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.11.1.1.2Gaseous Waste Processing System ActivityThe stripping fractions used in determining the amount of fission gases removed fromthe reactor coolant in the volume control tank and collected by the Gaseous WasteProcessing System (GWPS) are calculated as follows:1 KQKQKLVP(11.1-3)Where:=Nuclide volume control tank stripping fractionK=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=Henry's 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 controltank to obtain the Reactor Coolant System activity, volume control tank activity, andstripping fraction. Stripping fractions are shown in Table 11.1-1. Gaseous waste source terms are discussed in section 11.3. 99-01 11.1-4Reformatted PerAmendment 99-0111.1.1.3Volume Control Tank ActivityTable 11.1-3 lists the maximum activities in the volume control tank using theassumptions 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.11.1.1.4Pressurizer ActivityThe 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-gpmpressurizer 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 Henry's Law Equilibriumwith the liquid in the pressurizer.Pressurizer steam phase iodine concentrations are obtained from the liquid phasenuclide 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 11.1-1.11.1.1.5Liquid Waste Processing System ActivityLiquid waste source terms are discussed in Section 11.2.11.1.1.6Solid WasteSolid waste source terms are discussed in Section 11.5.11.1.2LEAKAGE RATESAs a necessary part of the effort to reduce effluent of radioactive liquid wastes, Westinghouse has been surveying various Pressurized Water Reactor (PWR) facilitieswhich 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 11.1-5Reformatted PerAmendment 99-01Leakage 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.11.1.3RADIOACTIVITIES IN THE FLUID SYSTEMS, REALISTIC MODELThe 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 11.1-5.11.1.4TRITIUMThe release of tritium to the environment from operating Westinghouse PWR's 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.11.1.4.1System SourcesThere 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 Li 6 , Li 7, and deuterium in the reactor water. Tritiumproduction from various sources is shown in Table 11.1-7.11.1.4.1.1Fission SourceThis 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 11.1-6Reformatted PerAmendment 99-01Previous 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. 11.1.4.1.2Control Rod SourceThe 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 arediscussed in Section 11.1.1.1.11.1.4.1.3Boric Acid SourceA direct contribution to the reactor coolant tritium concentration is made by neutronreaction 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. Theprincipal boron reactions are the B 10 (n, 2) H 3 and B 10 (n, ) Li 7 (n, n) H 3 reactions.The Li 7 reaction is controlled by limiting the overall lithium concentration toapproximately 2 ppm during operation. Li 6 is essentially excluded from the system byutilizing 99.9% Li 7.11.1.4.1.4Burnable Shim Rod SourceThese rods are in the core only during the first operating cycle and their potential tritium contribution occurs only during this period.11.1.4.1.5Lithium and DeuteriumLithium 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.11.1.4.2Design BasesThe 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 oftritium released from the fuel rods with Zircaloy cladding is less than originally expected. 99-01 11.1-7Reformatted PerAmendment 99-0111.1.4.3Design EvaluationTable 11.1-7 lists the present expected release of tritium to the reactor coolant. It willbe 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 Buildingair concentrations will be determined by the particular operating c onditions inside theReactor 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.

1.5REFERENCES

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 LightWater Reactors".3."Radiation Analysis Manual," Virgil C. Summer (CGE), CGE / 3-1, Revision 0, 12/92.99-01 99-01 99-01 11.1-8Reformatted PerAmendment 99-01TABLE 11.1-1PARAMETERS USED IN THE CALCULATION OF REACTORCOOLANT 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, ft 3 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 volumesa. Vapor, ft 3b. Liquid, ft 3 150 150 9.Fission product escape rate coefficients:a. Noble gas isotopes, sec -16.5 x 10-8b. Br, Rb, I, and Cs isotopes, sec -11.3 x 10-8c. Te isotopes, sec -11.0 x 10-9d. Mo, Tc, and Ag isotopes, sec -12.0 x 10-9e. Sr and Ba isotopes, sec -11.0 x 10-11f. Y, Zr, Nb, Ru, Rh, La, Ce, and Pr isotopes, sec -11.6 x 10-12 10.Mixed bed demineralizers decontamination factors:a. Br, I, Sr, and Ba 10.0b. Noble gases and all other isotopes 1.0 11.Cation bed demineralizer decontamination factorsa. Kr and Xe isotopes 1b. Sr and Ba isotopes 1c. Rb-86, Cs-134, and Cs-137 10d. Rb-88, Rb-89, Cs-136, and Cs-138 1e. Other isotopes 1 99-01 11.1-9Reformatted PerAmendment 99-01TABLE 11.1-1 (Continued)PARAMETERS USED IN THE CALCULATION OF REACTORCOOLANT FISSION AND CORROSION PRODUCT ACTIVITIES 12.Volume control tank noble gas stripping fractions: IsotopeStripping Fraction (1)Kr-856.0 x 10-5Kr-85m5.6 x 10-1Kr-878.2 x 10-1Kr-886.7 x 10-1Xe-131m1.3 x 10-2Xe-1333.0 x 10-2Xe-133m6.8 x 10-2Xe-1353.0 x 10-1Xe-135m9.4 x 10-1Xe-1389.4 x 10-1 13.Initial boron concentrations High (2)Low (3)2100 1200 14.Pressurizer volumes

a. Vapor, ft 3b. Liquid, ft 3 577 865 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 11.1-10Reformatted PerAmendment 99-01TABLE 11.1-1 (Continued)PARAMETERS USED IN THE CALCULATION OF REACTORCOOLANT FISSION AND CORROSION PRODUCT ACTIVITIES 19.Corrosion product parametersCore wetted areas, effective (in 2)a. Zirlo7.6 x 10 6b. Stainless steel4.9 x 10 5 c. Inconel7.7 x 10 5Out of core wetted area, Inconel (in 2)3.2 x 10 7Coolant velocity (ft/sec)

a. Core

b. Steam generator 14.0 18.0Nominal base metal release rates (mg/dm 2-mo)a. Zirlo 0.0b. Stainless steel 0.5 c. Inconel 1.0 99-01 11.1-11Reformatted PerAmendment 99-01TABLE 11.1-2REACTOR COOLANT EQUILIBRIUM FISSION AND CORROSION PRODUCT ACTIVITIES(Based on parameters given in Table 11.1-1)IsotopeActivity Ci/gmFission ProductsBr-844.2 x 10-2 Rb-88 3.8Sr-894.0 x 10-3Sr-902.0 x 10-4Sr-915.3 x 10-3Sr-921.2 x 10-3Y-905.7 x 10-5Y-915.4 x 10-4Y-921.1 x 10-3Zr-956.7 x 10-4 Nb-956.7 x 10-4Mo-997.9 x 10-1I-131 3.0I-132 3.1I-133 4.6I-1346.0 x 10-1I-135 2.4Te-1322.9 x 10-1Te-1342.8 x 10-2Cs-134 4.4Cs-136 4.5Cs-137 2.1Cs-1389.7 x 10-1Ba-1404.4 x 10-3La-1401.4 x 10-3Ce-1444.7 x 10-4Pr-1444.7 x 10-4Kr-85 7.6Kr-85m 1.8Kr-87 1.1Kr-88 3.2Xe-131m 2.3Xe-1332.9 x 10 2Xe-133m1.9 x 10 1Xe-135 8.6Xe-135m5.2 x 10-1Xe-1386.4 x 10-1Mn-544.1 x 10-4Mn-562.2 x 10-2 Co-581.4 x 10-2 Co-601.3 x 10-3Fe-595.2 x 10-4Cr-515.5 x 10-3 99-01 11.1-12Reformatted PerAmendment 99-01TABLE 11.1-3VOLUME CONTROL TANK EQUILIBRIUM ACTIVITIES(Based on parameters given in Table 11.1-1)

IsotopeVapor Activity (Ci/cc)Kr-83m3.3 x 10 0Kr-851.6 x 10 2Kr-85m1.9 x 10 1Kr-874.9 x 10 0Kr-882.5 x 10 1Xe-131m3.2 x 10 1Xe-1334.2 x 10 3Xe-133m2.7 x 10 2Xe-1351.1 x 10 2Xe-135m9.4 x 10 0Xe-1386.5 x 10-1 IsotopeLiquid Activity (Ci/gm)Kr-888.9 x 10-1Rb-883.8 x 10 0Xe-1331.7 x 10 2Xe-1354.3 x 10 0Cs-1344.4 x 10 0Cs-1364.5 x 10 0Cs-1372.1 x 10 0Cs-1389.7 x 10-1I-1313.0 x 10 0I-1323.1 x 10 0I-1334.6 x 10 0I-1346.0 x 10-1I-1352.4 x 10 0 99-01 11.1-13Reformatted PerAmendment 99-01TABLE 11.1-4PRESSURIZER ACTIVITIES(Based on parameters given in Table 11.1-1) IsotopeVapor Activity (Ci/cc)N-161.5 x 10-3Kr-83m1.8 x 10-2Kr-852.3 x 10 1Kr-85m2.1 x 10-1Kr-872.8 x 10-2Kr-882.2 x 10-1I-1313.0 x 10-2I-1323.1 x 10-2I-1334.6 x 10-2I-1346.0 x 10-3I-1352.4 x 10-2Xe-131m 5.1Xe-1336.3 x 10 2Xe-133m2.8 x 10 1Xe-135 2.1Xe-135m7.8 x 10-4Xe-1387.9 x 10-4 IsotopeLiquid Activity (Ci/gm)N-16 (maximum) 2.9Rb-88 3.8I-132 3.1I-133 4.6I-135 2.4Cs-134 4.4Cs-136 4.5Cs-1389.7 x 10-1Kr-88 3.2Xe-1332.9 x 10 2Xe-135 8.6NOTE:See Section 11.1.1.4 for additional information. 99-01 11.1-14Reformatted PerAmendment 99-01TABLE 11.1-5Normal Plant Operation Source TermsGroup I - Noble Gases(Based on ANSI / ANS-18.1-1984)No VCT PurgeNuclide YParameterReactor CoolantActivity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)Kr-85m5.2 x 10-11.6 x 10-13.9 x 10-8Kr-855.1 x 10-56.1 x 10-11.5 x 10-7Kr-877.9 x 10-11.7 x 10-13.9 x 10-8Kr-886.3 x 10-12.9 x 10-17.2 x 10-8Xe-131m1.1 x 10-27.4 x 10-11.8 x 10-7Xe-133m5.9 x 10-26.6 x 10-21.7 x 10-8Xe-1332.5 x 10-22.5 x 10 06.1 x 10-7Xe-135m9.3 x 10-11.6 x 10-13.8 x 10-8Xe-1352.7 x 10-18.3 x 10-12.1 x 10-7Xe-1379.8 x 10-14.2 x 10-21.0 x 10-8Xe-1389.3 x 10-11.5 x 10-13.5 x 10-8VCT Purge of 2.36 x 10 2 cm 3/secNuclide YParameterReactor CoolantActivity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)Kr-85m7.0 x 10-11.5 x 10-13.7 x 10-8Kr-855.7 x 10-11.1 x 10-22.7 x 10-9Kr-878.3 x 10-11.7 x 10-13.9 x 10-8Kr-887.5 x 10-12.8 x 10-17.0 x 10-8Xe-131m4.7 x 10-17.6 x 10-21.8 x 10-8Xe-133m4.9 x 10-12.4 x 10-26.0 x 10-9Xe-1334.8 x 10-14.8 x 10-11.2 x 10-7Xe-135m9.3 x 10-11.6 x 10-13.8 x 10-8Xe-1355.5 x 10-16.8 x 10-11.7 x 10-7Xe-1379.8 x 10-14.2 x 10-21.0 x 10-8Xe-1389.4 x 10-11.5 x 10-13.5 x 10-8 99-01 11.1-15Reformatted PerAmendment 99-01TABLE 11.1-5 (Continued)Normal Plant Operation Source TermsGroups II, III, IV, and V(Based on ANSI / ANS-18.1-1984)Group II - Halogens NuclideReactor CoolantActivity (Ci/gram)Steam GeneratorLiquid Activity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)Br-842.0 x 10-21.2 x 10-71.2 x 10-9I-1315.0 x 10-23.6 x 10-63.6 x 10-8I-1322.6 x 10-15.4 x 10-65.4 x 10-8I-1331.6 x 10-19.3 x 10-69.3 x 10-8I-134 4.2 x 10-14.0 x 10-64.0 x 10-8I-1353.1 x 10-11.2 x 10-51.2 x 10-7Group III - Rubidium, Cesium NuclideReactor CoolantActivity (Ci/gram)Steam GeneratorLiquid Activity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)Rb-882.3 x 10-18.6 x 10-74.2 x 10-9Cs-1347.5 x 10-35.5 x 10-72.9 x 10-9Cs-1369.3 x 10-46.7 x 10-83.4 x 10-10Cs-1379.9 x 10-37.4 x 10-73.7 x 10-9Group IV - N-16 NuclideReactor CoolantActivity (Ci/gram)Steam GeneratorLiquid Activity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)N-164.0 x 10 11.3 x 10-61.3 x 10-7Group V - Tritium NuclideReactor CoolantActivity (Ci/gram)Steam GeneratorLiquid Activity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)H-31.0 x 10 01.0 x 10-31.0 x 10-3 99-01 11.1-16Reformatted PerAmendment 99-01TABLE 11.1-5 (Continued)Normal Plant Operation Source TermsGroup VI - Miscellaneous Nuclides(Based on ANSI / ANS-18.1-1984) NuclideReactor CoolantActivity (Ci/gram)Steam GeneratorLiquid Activity (Ci/gram)Steam GeneratorSteam Activity (Ci/gram)Na-245.3 x 10-22.8 x 10-61.4 x 10-8Cr-513.3 x 10-32.5 x 10-71.2 x 10-9Mn-541.7 x 10-31.2 x 10-76.2 x 10-10Fe-551.3 x 10-39.3 x 10-84.7 x 10-10Fe-593.2 x 10-42.3 x 10-81.2 x 10-10 Co-584.9 x 10-33.6 x 10-71.8 x 10-9 Co-605.6 x 10-44.2 x 10-82.1 x 10-10Zn-655.4 x 10-44.0 x 10-81.9 x 10-10Sr-891.5 x 10-41.1 x 10-85.5 x 10-11Sr-901.3 x 10-59.3 x 10-104.7 x 10-12Sr-911.1 x 10-35.2 x 10-82.6 x 10-10Y-901.5 x 10-61.1 x 10-105.7 x 10-13Y-91m5.6 x 10-45.3 x 10-92.7 x 10-11Y-915.5 x 10-64.0 x 10-102.1 x 10-12Y-934.8 x 10-32.2 x 10-71.1 x 10-9Zr-954.2 x 10-43.0 x 10-81.5 x 10-10 Nb-953.0 x 10-42.1 x 10-81.1 x 10-10Mo-997.0 x 10-34.7 x 10-72.3 x 10-9Tc-99m5.5 x 10-32.0 x 10-71.0 x 10-9Ru-1038.0 x 10-35.9 x 10-73.0 x 10-9Ru-1069.6 x 10-27.0 x 10-63.4 x 10-8Rh-103m9.2 x 10-35.2 x 10-72.7 x 10-9Rh-1061.1 x 10-16.0 x 10-62.9 x 10-8Ag-110m1.4 x 10-31.0 x 10-75.1 x 10-10Te-129m2.0 x 10-41.5 x 10-87.4 x 10-11Te-1292.9 x 10-23.7 x 10-71.9 x 10-9Te-131m1.7 x 10-31.0 x 10-75.1 x 10-10Te-1319.5 x 10-34.8 x 10-82.5 x 10-10Te-1321.8 x 10-31.2 x 10-76.2 x 10-10Ba-137m9.4 x 10-37.0 x 10-73.5 x 10-9Ba-1401.4 x 10-29.8 x 10-74.9 x 10-9La-1402.7 x 10-21.8 x 10-68.7 x 10-9Ce-1411.6 x 10-41.2 x 10-85.9 x 10-11Ce-1433.1 x 10-31.9 x 10-79.6 x 10-10Ce-1444.3 x 10-33.0 x 10-71.6 x 10-9Pr-1433.7 x 10-32.3 x 10-71.2 x 10-9Pr-1444.9 x 10-32.6 x 10-71.3 x 10-9W-1872.8 x 10-31.6 x 10-78.3 x 10-10Np-2392.4 x 10-31.6 x 10-77.9 x 10-10 99-01 11.1-17Reformatted PerAmendment 99-01TABLE 11.1-6PARAMETERS USED TO DESCRIBE THE REACTOR SYSTEM-REALISTIC MODELParameterSymbolUnitsThermal power PMWt 2958Steam flowrate FSlbs/hr1.3 x 10 7Weight of water in the Reactor Coolant System WPlbs3.9 x 10 5Weight of water in all steam generators WSlbs3.4 x 10 5Reactor coolant letdown flow (purification) FDlbs/hr3.0 x 10 4Reactor coolant letdown flow (yearly average for boron control) FBlbs/hr 300Steam generator blowdown flow (total) FBDlbs/hr4.2 x 10 4Fraction of radioactivity in blowdown stream which is not returned to the secondary coolant systemNBD-1.0Flow through the purification system cation demineralizer FAlbs/hr3.0 x 10 3Ratio of condensate demineralizer flowrate to the total steam flowrate NC-0.0Ratio 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 11.1-5Primary to Secondary Leak Rate-lbs/day 100 99-01 11.1-18Reformatted PerAmendment 99-01TABLE 11.1-7TRITIUM PRODUCTIONTritium SourceExpected Release toReactor CoolantCuries/CycleProduced in CoreTernary Fissions 1420 IFBA's 237Produced in CoolantSoluble Boron 805Soluble Lithium 142Deuterium 3.24Total 2607 Note:Power level = 2958 MWtIFBA B-10 mass = 1730 gmInitial cycle reactor coolant boron concentration = 1200 ppmEquilibrium cycle reactor c oolant boron concentration = 2100 ppmLithium concentration (99.9 atom percent Li

7) = 2.2 ppm 99-01

SITE AREA: 2200 ACRES DAM DAM MONT ICELLO RESERVO IR AUXILIARY eUILDlfliG DA\CANAL::::::../\i/II i!"1\r--.... /v!""::::1,I""".If!'"/oJE ---,V{AiPLANT PROPERTY L17 1500 I iii SCALE IN FEET o 11500 3000 I o o o CD o t1'I o o o t1'I o t1'I Location of Liquid Release Points SOUTH CAROLINA ELECTRIC&GAS CO.VIRGILC.SUMMER NUCLEAR STATION FAIRFIELD PUMPED STORAGE FACILITY PENSTOCKS (A)LIQUID WASTE PROCESSING SYSTEM 1------------------1 (8)PROCESSED STEAM GENERATOR SLOWDOWN CIRULATING WATER DISCHARGE CANAL STEAM GENERATOR SLOWDOWN (S)TURBINE BUILDING FLOOR DRAINSLIQUID RELEASES: Figure 11.2-3 0.".""**",P'EXHIBIT L SHEET 4\..'.//7,0, AIRFIELD'<\/'<\I'OW'RHOUS£ '.,,'"\\\PLAN OF PENSTOCKS\\/ .."/.,0.<".,,>'"-t...'""'-'0 ,07.", ,,0'q,.>..cht'F'f FP-'-' 7'.*'UH"l'll.l..,.IlAH !ltOT-:.<,J: II l n SECTION THROUGH INTAI(E STRUCTURE ,,-*Z SECTION THROUGH!NT AKE CHANNEL------'--.'...(STlll,lCT ....c***-="=5:::':::: _.. ___-- ,__ -_"'ll.l"'-,....---."" ELEVATION OF INTAKE a PENSTOCKS SECTION THROUGH PENSTOCKS....0-.._._u.'$ IS'"0'nc_"NO... '!.oII"10(.UCtlFSE fOll PAGJEtt 010"" MAO!'Iv T>II! """"'0,,"' Of"!l:91'[WJ(" 1'1><'"-0 m-t." r-::!:=l.."![C:o 10'a.3:r-c.,,"CI 3:z...QI C m>>I\)...Q.3...=-0 Zr-Nc:m*a.nq.a=o'" ...0>>n...QI I.Q I\)>>>>:::I'" On

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 wa ste evaporators, r eactor 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 lo w such that site boundary doses will be a small fraction of regulation requirements.

The system is capable of operating under conditi ons 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 retai ned for historical purposes only. Components of a similar design to thos e used in the GWPS have been operating for several years with excellent performance. Systems constructed fr om 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 oper ation 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 inventor y 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 acti vities given in Table 11.1-2 shows that for a given power rating with 1% fuel def ects, the quantity of fission gas activity accumulated after 40 years continuous operati on is about twice the activity accumulated after short term operation. Figure 11.3-2 bas ed on realistic reactor coolant activities given in Table 11.1-5 shows t hat 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 distri buted 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 remova l from the reactor co olant during normal operation, it is expected to reduce plant acti vity levels caused by a leakage of reactor coolant. With operation of this system, it is possible to collect virtua lly all of the Krypton-85 released to the Reactor coolant and to ac hieve 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 inst rumentation diagrams are shown on Figures 11.3-3 and 11.3-4, respectively. Tabl e 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 equ ipment parameter s 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 pre ssure retaining components are allowed by Section III of the ASME Code, an d 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 West inghouse scope. Hence, all components within Westinghouse scope meet the design guidance as out lined 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 disp lacement 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 prov ided. 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 ar e self-contained and desig ned 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 construct ed of carbon steel. There are 8 waste gas decay tanks, 6 are used during normal operat ion 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 applicati on 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 retu rned 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 Instrumentat ion and Control Design The main system instrumentation is descr ibed in Table 11.3-6 and shown on the piping and instrumentation diagr ams, Figure 11.3-4.

The instrumentation readout is located main ly on the Waste Processing System (WPS) panel in the Auxiliary Building. Some inst ruments are read near the equipment location.

All alarms are shown separ ately on the WPS panel and further relayed to 1 common WPS annunciator on the main control board.

Where suitable, instrument li nes 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 s eparator level. During normal operation the proper seal water invent ory is maintained automatically.

Figure 11.3-6 indicates the location of the in struments 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 or ifice located upstream of the recombiner preheater. Local pressure gauges indicate pr essure at the recombiner inlet and the oxygen supply pressure. 11.3-5 Reformatted Per Amendment 00-01 The following controls and alarms are incor porated to maintain the gas composition outside the range of flammable and explosive mixtures:

1. A high flow alarm actuates when the vo lume 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 re combiner feed exceeds 4% by volume, a high hydrogen and high-high hydrogen/oxygen shutdown alarm sounds, the oxygen feed is terminated through TCV 01114, 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 recomb iner feed reaches 2% by volume, a high oxygen and high-high oxygen/shutdown ala rm 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 hy drogen feed, possibl e hydrogen-oxygen catalytic reactor malfuncti on, or loss of oxygen feed.

5. If oxygen in the recombiner disch arge 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, ox ygen feed is terminated. This control prevents an accumulation of oxyg en following system malfunction.

7. High discharge temperatur e 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 re combiner 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. Reac tor 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, fres h hydrogen is charged into the tank. The hydrogen-nitrogen mixture vented fr om the tank enters the circ ulating nitrogen stream at the compressor suction. Since the pressure downstream remains constant by use of a pressure regulating valve, nitrogen added to t he 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 re moved from the coolant. At that time, hydrogen gas is entering the volume contro l tank at 0.7 scfm and mixing with the 40 scfm circulating nitrogen stream to give a 1.8 volume % mi xture of hydrogen in nitrogen at the recombiner inlet. A pproximately 0.35 scfm of oxygen is added in the recombiner and reacted with the hydrogen to yield a dischar ge 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 fr om 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 ar e 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 circ ulated 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 be ing 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 t he reactor coolant must be re moved 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 fo r this operation. In addition, the flow of hydrogen to the volume control tank is st opped, 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 retai ned for historical purposes only. Compressor and recombiner packages are subjec ted to helium leak test after assembly. Initial performance tests are performed to ve rify 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 t he 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 Gu ide 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 stri pped 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 opt ions 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 t he Reactor Building when primary system components are opened or if leak age 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 Build ing Charcoal Cleanup S ystem is operated intermittently to reduce airborne iodine conc entrations 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 shut down) 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 wit h moderate potential for ra dioactive contamination (demineralizers, storage tanks, gas decay t anks, 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 pl enums continuously and is ducted to the main exhaust fans and the main plant vent. There is no by pass around this filter plenum.

The release calculations are based upon the a ssumption that reactor coolant leakage in the Auxiliary Building occurs primarily in the areas exhaust ed 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 distribut ion 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 se condary leakage occurs. Turb ine 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 Syst em 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 cr edit 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 parti tion factors (0.15 fo r 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 G enerator Blowdown The Steam Generator Bl owdown Processing System prov ides for cooling the blowdown in heat exchangers to prevent flashing. Co nsequently, 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 me thod 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 me teorology are made after a sample of decayed gaseous effluent is analyzed and are c ontinuously 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 assure s 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 Appen dix 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 pres ents data for the numbered vents shown by Figure 11.3-8. The data include base and exit elevations of the st acks, cross Section dimensions, volumetric flow rate, exit velocity, and comment

s. 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 fort h 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 calc ulation of annual average (/Q's) values are listed in Table 2.3-133. Exami nation of Table 2.3-1 33 reveals that the highest concentration of gaseous effluents at the exclusion zone boundary is expected to occur in the southeastern sector, w here relative concentration of 5.3 x 10 -6 sec/m 3 was calculated. RN 02-028 11.3-11 Reformatted Per Amendment 00-01 Expected annual gaseous release rates pres ented in Table 11.3-8 were used in conjunction with a ( /Q's) value of 5.3 x 10 -6 sec/m 3 to estimate maximum expected radioisotope concentrations, in air outside the restricted area. T he 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 av erage at the exclusion zone boundary over a 1-year period. It can be seen that the expected concentra tion 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, A ppendix B state that, for a mixture of radionuclides, the following relationship must hold:

N 1 i i i 1 ECL C Where:

C i = concentration of radionuclide i.

ECL i = effluent concentration limit of radionuclide i from 10 CFR 20, Appendix B, Table 2, Column 1.

N = number of radionucli des in the mixture.

The sum of the ratios of expected radi onuclide 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 ma terials 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 pl ant from each of the potentially significant pathways were calcul ated; methodology for the results of the calculations are discussed in the following paragraphs.

 (1) The term "exposure" as used in this Se ction refers only to the disposition of radioactive materials in t he 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, trans port times, bioaccumulation factors, dose conversion fact ors, 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, locat ed 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 acti vity 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 loca tion 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 r adioiodine and radioparticulates in the plant effluent was calculated to be 6.9 x 10 -2 mrem/yr to an adult's thyroid. The predicted dose to an individual obtaining 100% of his vegetable consumption from a garden adjacent to the nearest residence wa s also determined. Maximum calculated exposure from this pathway was 7.7 x 10 -1 mrem/yr to a child's 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 cu rrently being milked, t he suitability of the location for raising dairy cattle and the incr easing 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 10 0 mrem/yr to an infant's 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 indivi dual 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 m eat ingestion pathway was 4.6 x 10 -2 mrem/yr to a child.

Maximum individual doses calculated as de scribed 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 wa s 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 Rele ase Report. Meteorol ogical data used in preparation of the annual effluent report may consist of meteorological data averaged over multiple years to provide a better es timate 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, "C alculation of Releases of Radioactive Materials in Gaseous and Liquid Effluent s 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 RADIOA CTIVITY 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 oper ation 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 RADIOACTIV ITY 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 operati on 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 - c/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 11.3-17Reformatted PerAmendment 02-01TABLE 11.3-4PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM*BASIS: Power Level - 2900 MWtNo. of Units - 1Gas Decay Tanks (Note 4) - 8 Operating Interval - 1 dayStripping Efficiency - 0.4ITEM DESCRIPTIONTEMPPRESSFLOW N 2 H 2ISOTOPIC CONCENTRATION, C/cc(NOTE 1)GAS STREAMSFPSIGSCFM%%KR 85(NOTE 3)KR85MKR87KR88XE-133XE-133MXE-1351.VOLUME CONTROLTANK PURGE 130 150.7 0 1003.07 X 10-21.81 X 10-16.51 X 10-21.67 X 10-11.95 X 10 13.81 X 10-18.74 X 10-12.GAS DECAY TANKDISCH. TO COMP.AMB0.5 4099.90.11.31 X 10 19.95 X 10-25.61 X 10-31.45 X 10-13.79 X 10 11.405.50 X 10-13.COMPRESSORSUCTIONAMB0.540.798.31.71.29 X 10 11.01 X 10-16.63 X 10-31.45 X 10-13.76 X 10 11.395.56 X 10-14.COMP. DISCH. TORECOMBINER 140 4540.798.31.71.29 X 10 11.01 X 10-16.63 X 10-31.45 X 10-13.76 X 10 11.395.56 X 10-15.RECOMBINER DISCH.TO GAS DECAY TANKS 140 30 4099.90.11.31 X 10 11.03 X 10-16.75 X 10-31.48 X 10-13.82 X 10 11.415.66 X 10-16.MISC. VENTS-EVAPS.RCDT. RECYCLE HOLDUP TANKEDUCTOR 1400.5NEG 0 100 0 0 0 0 0 0 07.RECOMBINEROXYGEN SUPPLYAMB 500.35 0 0 0 0 0 0 0 0 08.RECOMBINERCALIBRATING GASAMB 150.004 100 4 0 0 0 0 0 0 09.RECOMBINERCALIBRATING GASAMB ATM0.004 100 4 0 0 0 0 0 0 0 98-01 02-01 02-01 02-01 11.3-18Reformatted PerAmendment 02-01TABLE 11.3-4 (Continued)PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM*ITEM DESCRIPTIONTEMPPRESSFLOW N 2 H 2ISOTOPIC CONCENTRATION, C/cc(NOTE 1)GAS STREAMSFPSIGSCFM%%KR 85(NOTE 3)KR85MKR87KR88XE-133XE-133MXE-13510.WASTE GAS SYSTEMNITROGEN SUPPLYAMB 100 0 100 0 0 0 0 0 0 0 011.NSSS NITROGENSUPPLYAMB 100 0 100 0 0 0 0 0 0 0 012.NITROGEN RELIEF TOPLANT VENTAMB 100 0 100 0 0 0 0 0 0 0 013.NSS HYDROGENSUPPLYAMB 1000.7 0 100 0 0 0 0 0 0 014.VOLUME CONTROLTANK HYDROGENAMB 1000.7 0 100 0 0 0 0 0 0 015.HYDROGEN RELIEF TOPLANT VENTAMB 100 0 0 100 0 0 0 0 0 0 016.WASTE GAS DISCH.TO PLANT VENTAMB ATM 0 100 01.31 x 10 1 0 0 0 0 0 017.RECYCLE GAS TOVOLUME CONTROLTANKAMB 100 0 100 0 0 0 0 0 0 0 018.PRESSURIZER RELIEFTANK VENT AND RETURN 120 3 0 100 0 0 0 0 0 0 0 019.SHUTDOWN TANKRELIEFAMB ATM 0 100 0 0 0 0 0 0 0 0 02-01 02-01 02-01 02-01 11.3-19Reformatted PerAmendment 02-01TABLE 11.3-4 (Continued)PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM*ITEM DESCRIPTIONTEMPPRESSFLOWISOTOPIC CONCENTRATION C/cc(NOTE 2)LIQUID STREAMSFPSIGGPDKR85(NOTE 3)KR85MKR87KR88XE-133XE-133MXE-1351.WASTE GASCOMPRESSOR DRAIN 140 45 03.432.69 X 10-21.77 X 10-33.87 X 10-28.273.05 X 10-11.22 X 10-12.RECOMBINER DRAIN 140 30 62.612.05 X 10-21.35 X 10-32.95 X 10-26.302.32 X 10-19.32 X 10-23.GAS DECAY TANKDRAINSAMB 40 369.18 X 10-16.98 X 10-33.94 X 10-41.02 X 10-22.208.13 X 10-23.19 X 10-24.SYSTEM DRAINS TOVOL CONTROL TANK 140 30-45 421.168.91 X 10-35.30 X 10-41.29 X 10-22.781.02 X 10-14.07 X 10-25.RECOMBINERREACTOR MAKEUPWATERAMB 0 0 0 0 0 0 0 06.COMPRESSORMAKEUP WATERAMB 36 0 0 0 0 0 0 0 02-01 02-01 02-01 11.3-20Reformatted PerAmendment 02-01TABLE 11.3-4 (Continued)PROCESS PARAMETERS FOR GASEOUS WASTE PROCESSING SYSTEM*ITEM COMPONENTTEMPPRESSVOL N 2 H 2COMPONENT INVENTORY, CURIESFPSIG FT 3%%KR 85(NOTE 3)KR85MKR87KR88XE-133XE-133MXE-135ACOMPRESSOR 140 45 498.31.76.024.58 X 10-22.58 X 10-36.67 X 10-21.74 X 10 16.45 X 10-12.53 X 10-1B.RECOMBINER 140 30 499.90.14.513.42 X 10-21.93 X 10-34.99 X 10-11.3 X 10 14.82 X 10-11.89 X 10-1C.GAS DECAY TANKAMB1.0 60099.90.11.04 X 10 31.801.02 X 10-12.626.86 X 10 22.54 X 10 19.96TOTAL SYSTEM6.28 X 10 31.881.06 X 10-12.743.18 X 10 32.65 X 10 11.04 X 10 1* 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 - Ambient6.NEG - Negligible7.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 (N 2 at 60F, 0 psig), scfm 40 Waste Gas Decay Tanks Number 8 Design pressure, psig 150 Design temperature, F 180 Volume (Each). ft 3 600 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 H 2 0 15 inches H 2 0 15 to -10 inches 8 to -5 inch -1 inches H 2 0 WPS panel and local LICA - 1032 Compressor Moisture Separator 150 180 0-30 inches H 2 0 15 inches H 2 0 15 to -10 inches -8 to -5 inches -1 inches H 2 0 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 ft 3 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 02-02802-01 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) NuclideGaseous Waste Processing SystemReactor Building Purge Exhaust Auxiliary Building LeakageTurbine Building Leakage 02-01 Condenser Air Removal System TotalKr-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 10-2Mn-54 4.5 x 10-5 2.1 x 10-4 1.8 x 10-4 (3) (3) 4.3 x 10-4Fe-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 10-6Cs-134 4.5 x 10-5 2.1 x 10-4 1.8 x 10-4 (3) (3) 4.3 x 10-4Cs-137 7.5 x 10-5 3.7 x 10-4 3.0 x 10-4 (3) (3) 7.4 x 10-4C-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 in 2 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 in 2 800 lb/hr 750 fpm 9. Main Steam Dump (5) (3 points) Intermediate 463'-0" 471'-0" 101.6 in 2 each 740,000 lb/hr each (6) 10. Main Steam Safety and Relief Valves(7) (10 points) Intermediate 463'-0" 475'-0" 233.7 in 2 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 in 2 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 in 2 each 740,000 lb/hr

each (6) Main Steam Power Relief Valve (5) (1 point) Auxiliary 485'-0" 497'-0" 113.1 in 2 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 in 2 2.33 x 10 6 lb/hr 40,000 fpm 15. EFW Pump Exhaust Intermediate 420'-0" 475'-0" 135 in 2 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 es timated occurrences are as follows: a. Actual, 22 times per y ear for 10 minutes each time. b. Test, 12 times per year for 1 minute each time.

(6) Main steam dump and power relief exit ve locities 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 st ack prior to diffuser release.

(7) Main steam safety and relief esti mated 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 plan

t. 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. 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. 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. 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 ( Ci/ml) Effluent (2) Concentration Limit In Air ( Ci/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 siteSame 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 siteLocation 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 Commission's App endix I Conformance Option, 40 FR 40816, September 4, 1975. (2) Evaluated at a location that is antici pated to be occupied during plant lifetime or evaluated with respect to such potentia l 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 expos ures in excess of the guidelin e values given above, the applicant should provide reasonable assura nce that a monitoring and surveillance pr ogram will be performed to determine:

1) the quantities of radioactive iodine ac tually released to the atmosphere and deposi ted 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 o ccurred; 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. Dose s 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 02-02510 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 with1%Fuel Defects and 60 gpm Continuous Letdown Figure 11.3-1 3 7 6 54 o.....c CD>C en ctS C9 c2 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 LetdownFigure11.3-2 !.q£N:;:c,.;wQ:,'-'**'1(J,lAl..YTj(llElXNllIE:RS 'N f'NW.J,.,tl -.....,<ii'.i'"'")>-_tJ-'--.... SOUTH CAROLINA ELECTRIC&GAS CO.VIRGllC.SUMMER NUCLEAR STATION Gaseous Waste Processing System Process flow Diagram figure 11.3-3

TO RECOMBINER COMPRESSOR FROM VOLUME CONTROL TANK MOISTURE SEPARATOR COOLER SEAL WATER RETURNT-TEMPERATURE MEASUREMENTP-PRESSURE MEASUREMENTL-LEVEL MEASUREMENT SOUTH CAROLINA ELECTRIC&GAS CO.VIRGILC.SUMMER NUCLEAR STATION Waste Gas Compressor Package Figure 11.3-5 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 TO GAS ANALYZER TO GAS ANALYZERT-TEMPERATURE MEASUREMENT P..PRESSURE MEASUREMENTF-FLOW MEASUREMENT SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION Catalytic Hydrogen Recombiner Package Figure 11.3-6 Gaseous Waste Release Points SOUTH CAROLINA ELECTRIC&GAS CO.VIRGIL C.SUMMER NUCLEAR STATION o WEST PENETRA TlON 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...-IHlcIH P WASTE HOLDUP TANK ()""-____.....__..J PIT HOLDUP TANK HOT MACHINE SHOP L-...JHlp: DECONTAMINATION 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-fHCI 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 PACKING EXHAUSTER AUXILIARY BUILDING SUMP.--LAUNDRY&HOT SHOWER TANK DEAERATOR CONDENSER AIR REMOVALWASTE EVAPORATOR CONDENSATE TANK BAY f-WASTE MONITOR TANKS SYSTEM__SECONDARY SPENT RESIN TANK AUXILIARY BUILDING LEGEND:[!]PRE-FILTER OR ROUGHING FILTER PENETRATION ACCESS AREAS PENETRA TlON 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 -$-DIESEL*GENERATOR\BUILDINGEXHAUST 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 GENERATORo I I FUEL HANDLING BUILDING MAIN CD PLANT VENTI C8J/ill (j)POWER OPERATOR RELIEF\AUXILIARY BUILDING n rl n4-ItT 0 Lb****o CONTROL BU I LO I NG<D o SOUTH CAROLINA ELECTRIC&GAS CO.VIRGILC.SUMMER NUCLEAR STATION Potentially Radioactive Gaseous Waste Release Points Figure 11.3-8

ROOF EL.511'-0" AUX I L1ARY BU I LD ING YENT96" RII-Al3---71 IC"[-ji"'[RM-A 14 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

Figure 11.5-1 Rev. 18 - Deleted by RN08-003, January, 2011 11.6-1AMENDMENT 97-01AUGUST 199711.6OFFSITE RADIOLOGICAL MONITORING PROGRAMThe offsite Radiological Environmental Monitoring Program is described in The Offsite Dose Calculation Manual.}}

RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 11.0, Radioactive Waste Management

Contents

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1.5REFERENCES

SUMMARY

SUMMARY

SUMMARY

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RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 11.0, Radioactive Waste Management

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