ML18053A736
| ML18053A736 | |
| Person / Time | |
|---|---|
| Site: | Lee  |
| Issue date: | 12/19/2017 |
| From: | Donahue J Duke Energy Carolinas |
| To: | Office of New Reactors |
| Hughes B | |
| References | |
| DUKE, DUKE.SUBMISSION.15, LEE.NP, LEE.NP.1 | |
| Download: ML18053A736 (184) | |
Text
UFSAR Table of Contents 1 Introduction and General Description of the Plant 2 Site Characteristics 3 Design of Structures, Components, Equipment and Systems 4 Reactor 5 Reactor Coolant System and Connected Systems 6 Engineered Safety Features 7 Instrumentation and Controls 8 Electric Power 9 Auxiliary Systems 10 Steam and Power Conversion 11 Radioactive Waste Management 12 Radiation Protection 13 Conduct of Operation 14 Initial Test Program 15 Accident Analyses 16 Technical Specifications 17 Quality Assurance 18 Human Factors Engineering 19 Probabilistic Risk Assessment UFSAR Formatting Legend Description Original Westinghouse AP1000 DCD Revision 19 content Departures from AP1000 DCD Revision 19 content Standard FSAR content Site-specific FSAR content Linked cross-references (chapters, appendices, sections, subsections, tables, figures, and references)
11.1 Source Terms............................................................................................... 11.1-1 11.1.1 Design Basis Reactor Coolant Activity ........................................ 11.1-1 11.1.1.1 Fission Products ........................................................ 11.1-1 11.1.1.2 Corrosion Products .................................................... 11.1-3 11.1.1.3 Tritium ........................................................................ 11.1-3 11.1.1.4 Nitrogen-16 ................................................................ 11.1-3 11.1.2 Design Basis Secondary Coolant Activity.................................... 11.1-3 11.1.3 Realistic Reactor Coolant and Secondary Coolant Activity ......... 11.1-4 11.1.4 Core Source Term ....................................................................... 11.1-4 11.1.5 Process Leakage Sources........................................................... 11.1-4 11.1.6 Combined License Information .................................................... 11.1-4 11.1.7 References .................................................................................. 11.1-4 11.2 Liquid Waste Management Systems............................................................ 11.2-1 11.2.1 Design Basis................................................................................ 11.2-1 11.2.1.1 Safety Design Basis................................................... 11.2-1 11.2.1.2 Power Generation Design Basis ................................ 11.2-1 11.2.1.3 Compliance with 10 CFR 20.1406 ............................. 11.2-6 11.2.2 System Description...................................................................... 11.2-7 11.2.2.1 Waste Input Streams ................................................. 11.2-7 11.2.2.2 Other Operations ..................................................... 11.2-10 11.2.2.3 Component Description ........................................... 11.2-10 11.2.2.4 Instrumentation Design ............................................ 11.2-13 11.2.2.5 System Operation and Performance........................ 11.2-13 11.2.3 Radioactive Releases ............................................................... 11.2-16 11.2.3.1 Discharge Requirements ......................................... 11.2-17 11.2.3.2 Estimated Annual Releases..................................... 11.2-17 11.2.3.3 Dilution Factor.......................................................... 11.2-17 11.2.3.4 Release Concentrations........................................... 11.2-18 11.2.3.5 Estimated Doses...................................................... 11.2-18 11.2.3.6 Quality Assurance.................................................... 11.2-20 11.2.4 Preoperational Testing .............................................................. 11.2-20 11.2.4.1 Sump Level Instrument Testing ............................... 11.2-20 11.2.4.2 Discharge Control/Isolation Valve Testing ............... 11.2-20 11.2.4.3 Preoperational Inspection ........................................ 11.2-21 11.2.5 Combined License Information .................................................. 11.2-21 11.2.5.1 Liquid Radwaste Processing by Mobile Equipment................................................................ 11.2-21 11.2.5.2 Cost Benefit Analysis of Population Doses.............. 11.2-21 11.2.5.3 Identification of Ion Exchange and Adsorbent Media ....................................................................... 11.2-21 11.2.5.4 Dilution and Control of Boric Acid Discharge ........... 11.2-21 11.2.6 References ................................................................................ 11.2-21 11.3 Gaseous Waste Management System......................................................... 11.3-1 11.3.1 Design Basis................................................................................ 11.3-1 11.3.1.1 Safety Design Basis................................................... 11.3-1 11.3.1.2 Power Generation Design Basis ................................ 11.3-1 11.3.1.3 Compliance with 10 CFR 20.1406 ............................. 11.3-3 11-i Revision 1
11.3.2.1 General Description ................................................... 11.3-4 11.3.2.2 System Operation ...................................................... 11.3-5 11.3.2.3 Component Description ............................................. 11.3-6 11.3.3 Radioactive Releases .................................................................. 11.3-8 11.3.3.1 Discharge Requirements ........................................... 11.3-8 11.3.3.2 Estimated Annual Releases....................................... 11.3-8 11.3.3.3 Release Points........................................................... 11.3-8 11.3.3.4 Estimated Doses........................................................ 11.3-9 11.3.3.5 Maximum Release Concentrations .......................... 11.3-12 11.3.3.6 Quality Assurance.................................................... 11.3-13 11.3.4 Inspection and Testing Requirements ....................................... 11.3-13 11.3.4.1 Preoperational Testing............................................. 11.3-13 11.3.4.2 Preoperational Inspection ........................................ 11.3-13 11.3.5 Combined License Information .................................................. 11.3-13 11.3.5.1 Cost Benefit Analysis of Population Doses.............. 11.3-13 11.3.5.2 Identification of Adsorbent Media............................. 11.3-14 11.3.6 References ................................................................................ 11.3-14 11.4 Solid Waste Management ............................................................................ 11.4-1 11.4.1 Design Basis................................................................................ 11.4-1 11.4.1.1 Safety Design Basis................................................... 11.4-1 11.4.1.2 Power Generation Design Basis ................................ 11.4-1 11.4.1.3 Functional Design Basis ............................................ 11.4-1 11.4.1.4 Compliance with 10 CFR 20.1406 ............................. 11.4-3 11.4.2 System Description...................................................................... 11.4-3 11.4.2.1 General Description ................................................... 11.4-3 11.4.2.2 Component Description ............................................. 11.4-6 11.4.2.3 System Operation ...................................................... 11.4-7 11.4.2.4 Waste Processing and Disposal Alternatives .......... 11.4-10 11.4.2.5 Facilities ................................................................... 11.4-11 11.4.3 System Safety Evaluation.......................................................... 11.4-12 11.4.4 Tests and Inspections................................................................ 11.4-12 11.4.5 Quality Assurance ..................................................................... 11.4-12 11.4.6 Combined License Information for Solid Waste Management System Process Control Program ............................................. 11.4-12 11.4.6.1 Procedures............................................................... 11.4-13 11.4.6.2 Third Party Vendors ................................................. 11.4-13 11.4.7 References ................................................................................ 11.4-13 11.5 Radiation Monitoring .................................................................................... 11.5-1 11.5.1 Design Basis................................................................................ 11.5-1 11.5.1.1 Safety Design Basis................................................... 11.5-1 11.5.1.2 Power Generation Design Basis ................................ 11.5-2 11.5.2 System Description...................................................................... 11.5-2 11.5.2.1 Radiation Monitoring System ..................................... 11.5-2 11.5.2.2 Monitor Functional Description .................................. 11.5-3 11.5.2.3 Monitor Descriptions .................................................. 11.5-3 11.5.2.4 Inservice Inspection, Calibration, and Maintenance ............................................................ 11.5-11 11-ii Revision 1
11.5.4 Process and Airborne Monitoring and Sampling ....................... 11.5-12 11.5.4.1 Effluent Sampling..................................................... 11.5-12 11.5.4.2 Representative Sampling......................................... 11.5-12 11.5.5 Post-Accident Radiation Monitoring .......................................... 11.5-13 11.5.6 Area Radiation Monitors ............................................................ 11.5-14 11.5.6.1 Design Objectives .................................................... 11.5-14 11.5.6.2 Post-Accident Area Monitors ................................... 11.5-15 11.5.6.3 Normal Range Area Monitors .................................. 11.5-16 11.5.6.4 Fuel Handling Area Criticality Monitors.................... 11.5-16 11.5.6.5 Quality Assurance.................................................... 11.5-17 11.5.7 Preoperational Testing .............................................................. 11.5-17 11.5.8 Combined License Information .................................................. 11.5-17 11.5.9 References ................................................................................ 11.5-18 11-iii Revision 1
Activities ....................................................................................................... 11.1-5 1-2 Design Basis Reactor Coolant Activity ......................................................... 11.1-6 1-3 Tritium Sources ............................................................................................ 11.1-7 1-4 Parameters Used to Calculate Secondary Coolant Activity ......................... 11.1-8 1-5 Design Basis Steam Generator Secondary Side Liquid Activity .................. 11.1-9 1-6 Design Basis Steam Generator Secondary Side Steam Activity ................ 11.1-10 1-7 Parameters Used to Describe Realistic Sources ....................................... 11.1-11 1-8 Realistic Source Terms .............................................................................. 11.1-12 1-201 Not Used .................................................................................................... 11.1-15 1-202 Not Used .................................................................................................... 11.1-16 1-203 Not Used .................................................................................................... 11.1-17 2-1 Liquid Inputs and Disposition ..................................................................... 11.2-23 2-2 Component Data - Liquid Radwaste System ............................................. 11.2-24 2-3 Summary of Tank Level Indication, Level Annunciators, and Overflows ... 11.2-31 2-4 Tank Surge Capacity .................................................................................. 11.2-32 2-5 Decontamination Factors ........................................................................... 11.2-33 2-6 Input Parameters for the GALE Computer Code ........................................ 11.2-34 2-7 Releases to Discharge Canal (Ci/Yr) Calculated by GALE Code .............. 11.2-37 2-8 Comparison of Annual Average Liquid Release Concentrations with 10 CFR 20 for Expected Releases Effluent Concentration Limits .............. 11.2-39 2-9 Comparison of Annual Average Liquid Release Concentrations with 10 CFR 20 Effluent Concentration Limits for Releases with Maximum Defined Fuel Defects .................................................................................. 11.2-41 2-201 Impoundment Model Parameters ............................................................... 11.2-43 2-202 LADTAP II Input Parameters ............................................................................11.2-44 2-203 Annual Dose to a Maximally Exposed Individual from Liquid Effluents (per unit) ..................................................................................................................11.2-45 2-204 Annual Population Dose from Liquid Effluents (per unit) ............................ 11.2-47 2-205 Liquid Pathway Doses Compared to 40 CFR Part 190 Limits .................... 11.2-48 2-206 Liquid and Gaseous Pathway Doses Compared to 40 CFR Part 190 Limits .......................................................................................................... 11.2-49 2-207 Liquid Pathway Comparison of Maximum Individual Dose to 10 CFR Part 50, Appendix I Criteria ........................................................................ 11.2-50 2-208 Liquid Pathway Comparison of Maximum Individual Dose to 10 CFR Part 20.1301 Criteria .................................................................................. 11.2-51 3-1 Gaseous Radwaste System Parameters ................................................... 11.3-15 3-2 Component Data (Nominal) Gaseous Radwaste System ...................... 11.3-16 3-3 Expected Annual Average Release of Airborne Radionuclides as Determined by the PWR-GALE Code, Revision 1 ..................................... 11.3-18 3-4 Comparison of Calculated Offsite Airborne Concentrations with 10 CFR 20 Limits ........................................................................................ 11.3-21 3-201 GASPAR II Input Parameters ..................................................................... 11.3-23 3-202 Individual Dose Rates ................................................................................ 11.3-24 3-203 Dose in MilliRads at Special Locations ...................................................... 11.3-27 3-204 Population Doses ....................................................................................... 11.3-28 3-205 Calculated Maximum Individual Doses Compared to 10 CFR Part 50 Appendix I Design Objectives .................................................................... 11.3-29 11-iv Revision 1
Effluents Compared to 10 CFR 20.1301 Limits .......................................... 11.3-30 3-207 Collective Gaseous Doses Compared to 40 CFR Part 190 Limits ............. 11.3-31 3-208 Population Dose by Isotopic Group ............................................................ 11.3-32 4-1 Estimated Solid Radwaste Volumes .......................................................... 11.4-15 4-2 Expected Annual Curie Content of Primary Influents ................................. 11.4-16 4-3 Maximum Annual Curie Content of Primary Influents ................................ 11.4-18 4-4 Expected Annual Curie Content of Shipped Primary Wastes .................... 11.4-20 4-5 Maximum Annual Curie Content of Shipped Primary Wastes .................... 11.4-22 4-6 Expected Annual Curie Content of Secondary Waste as Generated ......... 11.4-24 4-7 Maximum Annual Curie Content of Secondary Waste as Generated ........ 11.4-26 4-8 Expected Annual Curie Content of Shipped Secondary Wastes ................ 11.4-28 4-9 Maximum Annual Curie Content of Shipped Secondary Wastes ............... 11.4-30 4-10 Component Data Solid Waste Management System (Nominal) ............ 11.4-32 5-1 Radiation Monitor Detector Parameters ..................................................... 11.5-19 5-2 Area Radiation Monitor Detector Parameters ............................................ 11.5-21 5-201 Minimum Sampling Frequency ................................................................... 11.5-22 5-202 Minimum Sensitivities ................................................................................. 11.5-23 11-v Revision 1
Diagram ...................................................................................................... 11.2-52 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 1 of 8)............................................................................................... 11.2-53 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 2 of 8)............................................................................................... 11.2-54 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 3 of 8)............................................................................................... 11.2-55 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 4 of 8)............................................................................................... 11.2-56 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 5 of 8)............................................................................................... 11.2-57 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 6 of 8)............................................................................................... 11.2-58 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 7 of 8)............................................................................................... 11.2-59 2-2 Liquid Radwaste System Piping and Instrumentation Diagram (Sheet 8 of 8)............................................................................................... 11.2-60 3-1 Gaseous Radwaste System Simplified Sketch .......................................... 11.3-33 3-2 Gaseous Radwaste System Piping and Instrumentation Diagram ............. 11.3-34 4-1 Waste Processing System Flow Diagram .................................................. 11.4-34 5-1 Process In-Line Radiation Monitor ............................................................. 11.5-24 5-2 Safety-Related Containment High Range Radiation Monitor ..................... 11.5-25 5-3 Containment Atmosphere Radiation Monitor ............................................. 11.5-26 5-4 Plant Vent Radiation Monitor ...................................................................... 11.5-27 5-5 In-Line HVAC Duct Radiation Monitor ........................................................ 11.5-28 5-6 Safety-Related Main Control Room Supply Duct Radiation Monitor .......... 11.5-29 5-7 Liquid Offline Radiation Monitor ................................................................. 11.5-30 5-8 Adjacent to Line Radiation Monitor ............................................................ 11.5-31 5-9 HVAC Duct Particulate Radiation Monitor .................................................. 11.5-32 11-vi Revision 1
section addresses the sources of radioactivity that are treated by the liquid and gaseous waste systems. Radioactive materials are generated within the core (fission products) and have potential of leaking to the reactor coolant system by way of defects in the fuel cladding. The core ation field also results in activation of the coolant to form N-16 from oxygen and the activation of osion products in the reactor coolant system.
source terms are presented for the primary and the secondary coolant. The first is a servative, or design basis, source term that assumes the design basis fuel defect level. This rce term serves as a basis for system design and shielding requirements.
second source term is a realistic model. This source term represents the expected average centrations of radionuclides in the primary and the secondary coolant. These values are rmined using the model in the PWR-GALE code (Reference 1) and which provides the bases for mating typical concentrations of the principal radionuclides that are expected to occur. This rce term model reflects the industry experience at a large number of operating PWR plants.
.1 Design Basis Reactor Coolant Activity
.1.1 Fission Products the design basis source term it is assumed that there is a significant fuel defect level, well above anticipated during normal operation. It is assumed that small cladding defects are present in fuel producing 0.25 percent of the core power output (also stated as 0.25 percent fuel defects). The cts are assumed to be uniformly distributed throughout the core.
parameters used in the calculation of the reactor coolant fission product concentrations, uding pertinent information concerning the fission product escape rate coefficients, coolant nup rate, and demineralizer effectiveness, are listed in Table 11.1-1. Since the fuel defects are umed to be uniformly distributed in the core, the fission product escape rate coefficients are ed on average fuel temperature.
determination of reactor coolant activity is based on time-dependent fission product core ntories that are calculated by the ORIGEN code (Reference 2).
fission product activity in the reactor coolant is calculated using the following differential ations.
parent nuclides in the coolant:
dN c p FR p N F p
=
dt Mc Q DFp - 1
- p + D p + L Nc M c DFp p 11.1-1 Revision 1
dN c d FR d N F d
= + f p p N cp -
dt Mc QL DFd - 1 d + D d + N c d Mc DFd re:
Nc = Concentration of nuclide in the reactor coolant (atoms/gram)
NF = Population of nuclide in the fuel (atoms) t = Operating time (seconds)
R = Nuclide release coefficient (1/sec)
F = Fraction of fuel rods with defective cladding Mc = Mass of reactor coolant (grams)
= Nuclide decay constant (1/sec) 1 D = Dilution coefficient by feed and bleed (1/sec) = x B O - t DF Bo = Initial boron concentration (ppm)
= Boron concentration reduction rate (ppm/sec)
DF = Nuclide demineralizer decontamination factor QL = Purification or letdown mass flow rate (grams/sec) f = Fraction of parent nuclide decay events that result in the formation of the daughter nuclide Subscript p refers to the parent nuclide.
Subscript d refers to the daughter nuclide.
le 11.1-2 lists the resulting reactor coolant radionuclide concentrations. The values presented are maximum values calculated to occur during the fuel cycle from startup through the equilibrium
- e. Thus, the source term does not represent any particular time in the fuel cycle but is a servative composite.
design basis source term based on 0.25 percent fuel defects is used to ensure a consistent set esign values for interfaces among the radioactive waste processing systems. The Technical cifications in Chapter 16, which are related to fuel failure are also based upon 0.25 percent fuel 11.1-2 Revision 1
.1.2 Corrosion Products reactor coolant corrosion product activities are based on operating plant data and are pendent of fuel defect level. The concentrations of corrosion products are included in le 11.1-2.
.1.3 Tritium umber of tritium production processes add tritium to the reactor coolant:
Fission product formation in the fuel (ternary fission) forms tritium which can diffuse through the fuel clad or leak through fuel clad defects Neutron reactions with soluble boron in the reactor coolant Burnable neutron absorber Neutron reactions with soluble lithium in the reactor coolant Neutron reactions with deuterium in the reactor coolant first two processes are the principal contributors to tritium in the reactor coolant. Table 11.1-3 the tritium introduced to the reactor coolant from each of the processes.
um exists in the reactor coolant primarily combined with hydrogen (that is, a tritium atom replaces drogen atom in a water molecule) and thus cannot be readily separated from the coolant by mal processing methods. The maximum concentration of tritium in the reactor coolant is less than microcuries per gram as a result of losses due to leakage and the controlled release of tritiated er to the environment.
.1.4 Nitrogen-16 vation of oxygen in the coolant results in the formation of N-16 which is a strong gamma emitter.
ause of its short half-life of 7.11 seconds, N-16 is not of concern outside the containment.
le 12.2-3 provides N-16 concentrations at various points in the reactor coolant system. After tdown, N-16 is not a source of radiation inside of containment.
.2 Design Basis Secondary Coolant Activity am generator tube defects cause the introduction of reactor coolant into the secondary cooling em. The resulting radionuclide concentrations in the secondary coolant depend upon the ary-to-secondary leak rate, the nuclide decay constant, and the steam generator blowdown rate.
reactor coolant leakage into the secondary system is assumed to have radionuclide centrations as defined in Table 11.1-2. The parameters used in the calculation of the secondary activities are provided in Table 11.1-4 and the resulting radionuclide concentrations in the steam erator secondary side water and steam are presented in Tables 11.1-5 and 11.1-6.
11.1-3 Revision 1
g the modeling in ANSI-18.1 (Reference 3). This modeling is also incorporated in the R-GALE code. The reference plant values provided in ANSI-18.1 were adjusted to be consistent the AP1000 parameters listed in Table 11.1-7. The adjustment factors are applied to the fission ducts. The realistic source term is listed in Table 11.1-8.
.4 Core Source Term core fission product inventories used to establish source terms for accident radiological sequence analyses are provided in Appendix 15A.
.5 Process Leakage Sources systems containing radioactive liquids are potential sources for the release of radioactive erial to plant buildings and then to the environment. The leakage sources and the resulting orne concentrations are discussed in Section 12.2.
ease pathways for radioactive materials are discussed in Sections 11.2 and 11.3.
.6 Combined License Information section contained no requirement for information.
.7 References Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code), NUREG-0017, Revision 1, March 1985.
RSIC Computer Code Collection CCC-371, ORIGEN 2.1 Isotope Generation and Depletion Code - Matrix Exponential Method, August 1, 1991.
ANSI/ANS-18.1-1984, Radioactive Source Term for Normal Operation of Light Water Reactors.
11.1-4 Revision 1
re thermal power (MWt) 3,400 3)(a) 9,575 actor coolant liquid volume (ft actor coolant full-power average temperature (°F) 578.1 rification flow rate (gal/min)(b)
Maximum 100 Normal 91.3 ective cation demineralizer flow, annual average (gal/min)(b) 9.1 clide release coefficients (the product of the failed fuel fraction and the fission duct escape rate coefficient)
Equivalent fraction of core power produced by fuel rods containing small cladding 0.0025 defects (failed fuel fraction) sion product escape rate coefficients during full-power operation (s-1):
Kr and Xe isotopes 6.5 x 10-8 Br, Rb, I, and Cs isotopes 1.3 x 10-8 Mo, Tc, and Ag isotopes 2.0 x 10-9 Te isotopes 1.0 x 10-9 Sr and Ba isotopes 1.0 x 10-11 Y, Zr, Nb, Ru, Rh, La, Ce, and Pr isotopes 1.6 x 10-12 emical and volume control system mixed bed demineralizers Resin volume (ft3) 50 mineralizer isotopic decontamination factors:
Kr and Xe isotopes 1 Br and I isotopes 10 Sr and Ba isotopes 10 Other isotopes 1 emical and volume control system cation bed demineralizer Resin volume (ft3) 50 mineralizer isotopic decontamination factors:
Kr and Xe isotopes 1 Sr and Ba isotopes 1 Rb-86, Cs-134, and Cs-137 10 Rb-88, Rb-89, Cs-136, and Cs-138 1 Other isotopes 1 her isotopic removal mechanisms See Note c.
ial boron concentration (ppm) 1,400 eration time (effective full-power hours) 12,492 tes:
Reactor coolant mass used in defining fission product activities is based on above stated conditions before thermal expansion (conservative).
Flow calculated at 2250 psia and 250°F.
For all isotopes, except the isotopes of Kr, Xe, Br, I, Rb, Cs, Sr, and Ba, a removal decontamination factor of 10 is assumed to account for removal mechanisms other than ion exchange, such as plateout or filtration. This decontamination factor is applied to the normal purification letdown flow.
11.1-5 Revision 1
Activity Activity Nuclide (Ci/g) Nuclide (Ci/g)
Kr-83m 1.8 x 10-1 Rb-88 1.5
-1 Rb-89 6.9 x 10-2 Kr-85m 8.4 x 10 Kr-85 3.0 Sr-89 1.1 x 10-3 Kr-87 4.7 x 10-1 Sr-90 4.9 x 10-5 Kr-88 1.5 Sr-91 1.7 x 10-3 Kr-89 3.5 x 10-2 Sr-92 4.1 x 10-4 Xe-131m 1.3 Y-90 1.3 x 10-5 Xe-133m 1.7 Y-91m 9.2 x 10-4 Xe-133 1.2 x 102 Y-91 1.4 x 10-4 Xe-135m 1.7 x 10-1 Y-92 3.4 x 10-4 Xe-135 3.5 Y-93 1.1 x 10-4 Xe-137 6.7 x 10-2 Zr-95 1.6 x 10-4 Xe-138 2.5 x 10-1 Nb-95 1.6 x 10-4 Br-83 3.2 x 10-2 Mo-99 2.1 x 10-1 Br-84 1.7 x 10-2 Tc-99m 2.0 x 10-1 Br-85 2.0 x 10-3 Ru-103 1.4 x 10-4 I-129 1.5 x 10-8 Rh-103m 1.4 x 10-4 I-130 1.1 x 10-2 Rh-106 4.5 x 10-5 I-131 7.1 x 10-1 Ag-110m 4.0 x 10-4 I-132 9.4 x 10-1 Te-127m 7.6 x 10-4 I-133 1.3 Te-129m 2.6 x 10-3 I-134 2.2 x 10-1 Te-129 3.8 x 10-3 I-135 7.8 x 10-1 Te-131m 6.7 x 10-3 Cs-134 6.9 x 10-1 Te-131 4.3 x 10-3 Cs-136 1.0 Te-132 7.9 x 10-2 Cs-137 5.0 x 10-1 Te-134 1.1 x 10-2 Cs-138 3.7 x 10-1 Ba-137m 4.7 x 10-1 Cr-51 1.3 x 10-3 Ba-140 1.0 x 10-3 Mn-54 6.7 x 10-4 La-140 3.1 x 10-4 Mn-56 1.7 x 10-1 Ce-141 1.6 x 10-4 Fe-55 5.0 x 10-4 Ce-143 1.4 x 10-4 Fe-59 1.3 x 10-4 Pr-143 1.5 x 10-4 Co-58 1.9 x 10-3 Ce-144 1.2 x 10-4 Co-60 2.2 x 10-4 Pr-144 1.2 x 10-4 e activities are used for shielding and radwaste system interface design. For 1 percent fuel defect calculations (maximum se and liquid and gaseous radwaste system capability) multiply the activities above by 4 except for corrosion products 1, Mn-54, Mn-56, Fe-55, Fe-59, Co-58 and Co-60).
11.1-6 Revision 1
Release to Coolant (curies/cycle1)
Tritium Source Design Basis Best Estimate oduced in core Ternary fission 1770 354 Burnable absorbers 279 56 oduced in coolant Soluble boron 734 734 Soluble lithium 168 168 Deuterium 4 4 TAL 2955 1316 Cycle length of 18 months. Design basis case reflects the historical assumption that 10% of the tritium produced in the core is released to the coolant. Best-estimate case is based on a release of only 2% of the tritium.
11.1-7 Revision 1
tal secondary side water mass (lb/steam generator) 1.68 x 105 eam generator steam fraction 0.058 otal steam flow rate (lb/hr) 1.5 x 107 oisture carryover (percent) 0.1 tal makeup water feed rate (lb/hr) 700 otal blowdown rate (gpm) 186 tal primary-to-secondary leak rate (gpd) 300 dine partition factor (mass basis) 100 11.1-8 Revision 1
Activity Activity Nuclide (Ci/g) Nuclide (Ci/g)
Br-83 1.4 x 10-5 Y-93 8.2 x 10-8 Br-84 2.4 x 10-6 Zr-95 1.5 x 10-7 Br-85 3.1 x 10-8 Nb-95 1.5 x 10-7 I-129 1.3 x 10-11 Mo-99 1.9 x 10-4 I-130 7.9 x 10-6 Tc-99m 1.7 x 10-4 I-131 6.3 x 10-4 Ru-103 1.2 x 10-7 I-132 4.2 x 10-4 Rh-106 4.1 x 10-8 I-133 1.0 x 10-3 Rh-103m 1.2 x 10-7 I-134 4.9 x 10-5 Rh-106 4.1 x 10-8 I-135 5.0 x 10-4 Ag-110m 3.0 x 10-6 Rb-86 1.4 x 10-5 Te-125m 1.5 x 10-7 Rb-88 1.4 x 10-4 Te-127m 7.0 x 10-7 Rb-89 5.6 x 10-6 Te-127 2.2 x 10-6 Cs-134 1.1 x 10-3 Te-129m 2.4 x 10-6 Cs-136 1.7 x 10-3 Te-129 2.1 x 10-6 Cs-137 8.2 x 10-4 Te-131m 5.6 x 10-6 Cs-138 5.9 x 10-5 Te-131 1.6 x 10-6 H-3 3.8 x 10-1 Te-132 7.0 x 10-5 Cr-51 1.3 x 10-6 Te-134 2.0 x 10-6 Mn-54 6.6 x 10-7 Ba-137m 7.7 x 10-4 Mn-56 7.8 x 10-5 Ba-140 9.4 x 10-7 Fe-55 5.0 x 10-7 La-140 3.3 x 10-7 Fe-59 1.3 x 10-7 Ce-141 1.4 x 10-7 Co-58 1.9 x 10-6 Ce-143 1.2 x 10-7 Co-60 2.2 x 10-7 Ce-144 1.1 x 10-7 Sr-89 1.8 x 10-6 Pr-143 1.4 x 10-7 Sr-90 8.0 x 10-8 Pr-144 1.1 x 10-7 Sr-91 1.9 x 10-6 Sr-92 2.4 x 10-7 Y-90 1.4 x 10-8 Y-91m 1.0 x 10-6 Y-91 1.3 x 10-7 Y-92 2.8 x 10-7 11.1-9 Revision 1
Nuclide Activity (Ci/g)
Kr-83m 1.1 x 10-6 Kr-85m 4.3 x 10-6 Kr-85 1.5 x 10-5 Kr-87 2.4 x 10-6 Kr-88 7.7 x 10-6 Kr-89 1.8 x 10-7 Xe-131m 6.9 x 10-6 Xe-133m 8.7 x 10-6 Xe-133 6.4 x 10-4 Xe-135m 5.5 x 10-6 Xe-135 1.9 x 10-5 Xe-137 3.4 x 10-7 Xe-138 1.3 x 10-6 I-129 1.5 x 10-13 I-130 8.7 x 10-8 I-131 6.9 x 10-6 I-132 4.7 x 10-6 I-133 1.1 x 10-5 I-134 5.4 x 10-7 I-135 5.5 x 10-6 H-3 3.8 x 10-1 11.1-10 Revision 1
AP1000 Parameter Symbol Units Value Nominal Value ermal power P MWt 3400 3400 eam flow rate FS lb/hr 1.5 x 107 1.5 x 107 ight of water in reactor coolant system WP lb 4.3 x 105 5.5 x 105 ight of water in all steam generators WS lb 3.5 x 105 4.5 x 105 actor coolant purification flow FD lb/hr 4.3 x 104 3.7 x 104 actor coolant letdown flow (yearly FB lb/hr 1.5 x 102 5.0 x 102 erage for boron control) eam generator blowdown flow (total) FBD lb/hr 7.5 x 104 7.5 x 104 action of radioactivity in blowdown NBD - 0.0 1.0 eam which is not returned to the condary coolant system w through the purification system cation FA lb/hr 4.3 x 103 3.7 x 103 mineralizer tio of condensate demineralizer flow NC - 0.33 0.0 e to the total steam flow rate action of the noble gas activity in the Y - 0.0 0.0 down stream which is not returned to the ctor coolant system mary-to-secondary leakage FL lb/day 75 75 11.1-11 Revision 1
Noble Gases Reactor Coolant Steam Generator Activity Steam Activity Nuclide (Ci/g) (Ci/g)
Kr-85m 0.21 4.4 x 10-8 Kr-85 1.4 2.9 x 10-7 Kr-87 0.19 3.9 x 10-8 Kr-88 0.36 7.7 x 10-8 Xe-131m 1.1 2.3 x 10-7 Xe-133m 0.093 2.0 x 10-8 Xe-133 3.6 7.6 x 10-7 Xe-135m 0.17 3.5 x 10-8 Xe-135 1.1 2.3 x 10-7 Xe-137 0.044 9.2 x 10-9 Xe-138 0.15 3.2 x 10-8 Halogens Reactor Coolant Steam Generator Steam Generator Activity Liquid Activity Steam Activity Nuclide (Ci/g) (Ci/g) (Ci/g)
Br-84 0.02 1.2 x 10-7 1.2 x 10-9 I-131 0.04 2.7 x 10-6 2.7 x 10-8 I-132 0.25 5.1 x 10-6 5.1 x 10-8 I-133 0.14 7.4 x 10-6 7.4 x 10-8 I-134 0.42 3.9 x 10-6 3.9 x 10-8 I-135 0.28 1.1 x 10-5 1.1 x 10-7 Rubidium, Cesium Reactor Coolant Steam Generator Steam Generator Activity Liquid Activity Steam Activity Nuclide (Ci/g) (Ci/g) (Ci/g)
Rb-88 0.24 8.9 x 10-7 4.4 x 10-9 Cs-134 5.9 x 10-3 1.5 x 10-6 7.6 x 10-9 Cs-136 7.4 x 10-4 1.7 x 10-7 8.7 x 10-10 Cs-137 7.9 x 10-3 2.0 x 10-6 9.9 x 10-9 11.1-12 Revision 1
Tritium Reactor Coolant Steam Generator Steam Generator Activity Liquid Activity Steam Activity Nuclide (Ci/g) (Ci/g) (Ci/g)
H-3 1 1.0 x 10-3 1.0 x 10-3 Miscellaneous Nuclides Reactor Coolant Steam Generator Steam Generator Activity Liquid Activity Steam Activity Nuclide (Ci/g) (Ci/g) (Ci/g)
Na-24 4.6 x 10-2 3.6 x 10-6 1.8 x 10-8 Cr-51 2.6 x 10-3 3.6 x 10-7 1.8 x 10-9 Mn-54 1.3 x 10-3 1.8 x 10-7 9.2 x 10-10 Fe-55 1.0 x 10-3 1.4 x 10-7 7.0 x 10-10 Fe-59 2.5 x 10-4 3.3 x 10-8 1.7 x 10-10 Co-58 3.9 x 10-3 5.3 x 10-7 2.6 x 10-9 Co-60 4.4 x 10-4 6.1 x 10-8 3.1 x 10-10 Zn-65 4.3 x 10-4 5.9 x 10-8 2.8 x 10-10 Sr-89 1.2 x 10-4 1.6 x 10-8 8.1 x 10-11 Sr-90 1.0 x 10-5 1.4 x 10-9 7.0 x 10-12 Sr-91 9.8 x 10-4 6.4 x 10-8 3.2 x 10-10 Y-90 1.2 x 10-6 1.6 x 10-10 8.0 x 10-13 Y-91m 5.7 x 10-4 5.6 x 10-9 2.8 x 10-11 Y-91 4.4 x 10-6 5.9 x 10-10 3.1 x 10-12 Y-93 4.3 x 10-3 2.8 x 10-7 1.4 x 10-9 Zr-95 3.3 x 10-4 4.5 x 10-8 2.2 x 10-10 Nb-95 2.4 x 10-4 3.1 x 10-8 1.6 x 10-10 Mo-99 5.6 x 10-3 6.7 x 10-7 3.2 x 10-9 Tc-99m 5.1 x 10-3 2.4 x 10-7 1.2 x 10-9 Ru-103 6.3 x 10-3 8.6 x 10-7 4.5 x 10-9 Ru-106 7.5 x 10-2 1.0 x 10-5 5.0 x 10-8 Rh-103m 6.3 x 10-3 8.6 x 10-7 4.5 x 10-9 Rh-106 7.5 x 10-2 1.0 x 10-5 5.0 x 10-8 Ag-110m 1.1 x 10-3 1.5 x 10-7 7.5 x 10-10 Te-129m 1.6 x 10-4 2.2 x 10-8 1.1 x 10-10 11.1-13 Revision 1
Miscellaneous Nuclides Reactor Coolant Steam Generator Steam Generator Activity Liquid Activity Steam Activity Nuclide (Ci/g) (Ci/g) (Ci/g)
Te-129 2.9 x 10-2 3.9 x 10-7 2.0 x 10-9 Te-131m 1.4 x 10-3 1.4 x 10-7 7.0 x 10-10 Te-131 9.7 x 10-3 4.9 x 10-8 2.5 x 10-10 Te-132 1.5 x 10-3 1.8 x 10-7 8.9 x 10-10 Ba-137m 7.4 x 10-3 1.9 x 10-6 9.3 x 10-9 Ba-140 1.1 x 10-2 1.4 x 10-6 7.2 x 10-9 La-140 2.3 x 10-2 2.4 x 10-6 1.2 x 10-8 Ce-141 1.3 x 10-4 1.7 x 10-8 8.6 x 10-11 Ce-143 2.6 x 10-3 2.6 x 10-7 1.3 x 10-9 Ce-144 3.4 x 10-3 4.5 x 10-7 2.3 x 10-9 Pr-143 3.0 x 10-3 3.3 x 10-7 1.8 x 10-9 Pr-144 3.4 x 10-3 4.5 x 10-7 2.3 x 10-9 W-187 2.3 x 10-3 2.2 x 10-7 1.1 x 10-9 Np-239 2.0 x 10-3 2.2 x 10-7 1.1 x 10-9 11.1-14 Revision 1
Not Used 11.1-15 Revision 1
Not Used 11.1-16 Revision 1
Not Used 11.1-17 Revision 1
ose of liquids containing radioactive material. These include the following:
Steam generator blowdown processing system (Subsection 10.4.8);
Radioactive waste drain system (Subsection 9.3.5);
Liquid radwaste system (WLS) (Section 11.2).
section primarily addresses the liquid radwaste system. The other systems are also addressed ubsection 11.2.3, which discusses the expected releases from the liquid waste management ems.
liquid radwaste system is designed to control, collect, process, handle, store, and dispose of d radioactive waste generated as the result of normal operation, including anticipated operational urrences.
.1 Design Basis section 1.9.1 discusses the conformance of the liquid radwaste system design with the criteria of ulatory Guide 1.143.
.1.1 Safety Design Basis liquid radwaste system serves no safety-related functions except for:
Containment isolation; see Subsection 6.2.3.
Draining the passive core cooling system compartments to the containment sump to prevent flooding of these compartments and possible immersion of safety-related components.
Back flow prevention check valves in the drain lines from the chemical and volume control system compartment and the passive core cooling system compartments to the containment sump, which prevent cross flooding of these compartments. Each drain line has two check valves in series so that a single failure does not compromise the back flow prevention safety function. See Subsection 6.3.3.3.2 for a discussion of containment flooding.
.1.2 Power Generation Design Basis
.1.2.1 Capacity liquid radwaste system provides holdup capacity as shown in Table 11.2-2, and permanently alled processing capacity of 75 gpm through the ion exchange/filtration train. This is adequate acity to meet the anticipated processing requirements of the plant. The projected flows of various d waste streams to the liquid radwaste system under normal conditions are identified in le 11.2-1.
liquid radwaste system design can accept equipment malfunctions without affecting the ability of the system to handle both anticipated liquid waste flows and possible surge load due to essive leakage. Table 11.2-4 contains information on the surge capacity of individual tanks.
ions of the liquid radwaste system may become unavailable as a result of the malfunctions listed ubsection 11.2.1.2.2.
11.2-1 Revision 1
ccommodate the anticipated operational occurrences described in Subsection 11.2.1.2.3.
.1.2.2 Failure Tolerance
.1.2.2.1 Pump Failure ere operation is not essential and surge capacity is available, a single pump is provided. This lies to most applications in the liquid radwaste system. Two reactor coolant drain tank pumps and containment sump pumps are provided because the relative inaccessibility of the containment ng power operation would hinder maintenance. The containment sump pumps are submersible ps with permanently lubricated bearings and mechanical seals. To protect them from damage to loss of suction, each pump is interlocked to stop on a low level condition in the sump. The tor coolant drain tank pumps are vertical sump type pumps with motors above the reactor lant drain tank shaft coupled to pumps submersed in the liquid within the reactor coolant drain
. This arrangement minimizes contamination of the motors and permits removal and ntenance of the motors outside of the radiation area.
cess pumps located outside containment are air-operated, double diaphragm type. These pumps capable of significant suction lifts, and can thus be located on or near the top of the associated te tank, with internal suction piping. They can pump slurries with high solids fractions, run dheaded, and run dry without damage. In addition, they can operate over a wide range of raulic conditions by varying the driving air input. This makes it possible to fulfill many different lications with a single pump model, thereby facilitating maintenance and reducing the inventory of re parts.
.1.2.2.2 Filter or Ion Exchanger Plugging rumentation is provided to give indication of the pressure drop across filters and ion exchangers.
odic checks of the pressure drops provide indication of equipment fouling, thus permitting ective action to be taken before an excessive pressure drop is reached. Change of filter ridges and ion exchange beds is expected to occur based upon radiation survey.
.1.2.3 Anticipated Operational Occurrences
.1.2.3.1 High Primary Coolant System Leakage Rate system is designed to handle an abnormal primary coolant system leak in addition to the ected leakage during normal operation. Operation of the system is the same as for normal ration, except that the load on the system is increased.
.1.2.3.2 High Use of Decontamination Water rge quantities of water are used to decontaminate areas or equipment, the load on the liquid waste system is increased. However, the liquid radwaste system is designed to handle a large, tinuous input to the waste holdup tanks. If the water can be discharged without processing based ampling which shows acceptably low activity, the overall liquid radwaste system capacity is eased.
ccommodate the possible use of special decontamination fluids or very large volumes of ontamination fluids, mobile equipment is used as discussed in Subsection 11.2.1.2.5.2.
11.2-2 Revision 1
cribed in Subsection 10.4.8. However, if excessive radioactivity is detected, the blowdown is rted to the liquid radwaste system for processing and disposal.
blowdown fluid is brought into the waste holdup tanks, which provide some surge capacity to the fluid during processing. It is then processed in the same fashion as, and combined with, r inputs.
e event of a steam generator tube rupture, the condensate storage tank may also become taminated. In this event, the tank is cleaned by the use of temporary equipment brought to the site he purpose, as described in Subsection 11.2.1.2.5.2.
.1.2.3.4 Refueling load on the liquid radwaste system is expected to increase during refueling because of the eased level of maintenance activities in the plant, but operation is the same as for normal plant ration. There is no significant effect on the performance capability of the liquid radwaste system.
.1.2.4 Controlled Release of Radioactivity liquid radwaste system provides the capability to reduce the amounts of radioactive nuclides ased in the liquid wastes through the use of demineralization and time delay for decay of short-d nuclides.
assumed equipment decontamination factors appear in Table 11.2-5. Estimates of the oactive source terms and annual average flow rate that will be processed in the liquid radwaste em or discharged to the environment during normal operation appear in Table 11.2-1.
ore radioactive liquid waste is discharged, it is pumped to a monitor tank. A sample of the monitor contents is analyzed, and the results are recorded. In this way, a record is kept of planned ases of radioactive liquid waste.
liquid waste is discharged from the monitor tank in a batch operation, and the discharge flow rate stricted as necessary to maintain an acceptable concentration when diluted by the circulating er discharge flow. These provisions preclude uncontrolled releases of radioactivity.
ddition, the discharge line contains a radiation monitor with diverse methods of stopping the harge. The first method closes an isolation valve in the discharge line, which prevents any further harge from the liquid radwaste system. The valve automatically closes and an alarm is actuated e activity in the discharge stream reaches the monitor setpoint. The second method stops the itor tank pumps.
minimize leakage from the liquid radwaste system, the system is of welded construction except re flanged connections are required to facilitate component maintenance or to allow connection mporary or mobile equipment. Air-operated diaphragm pumps or pumps having mechanical ls are used. These pumps minimize system leakage thereby minimizing the release of radioactive that might be entrained in the leaking fluid to the building atmosphere.
visions are made to control spills of radioactive liquids due to tank overflows. Table 11.2-3 lists the isions for tank level indication, alarms, and overflow disposition for liquid radwaste system tanks ide containment. In addition, the radioactive waste collection tanks (i.e., the effluent holdup s, waste holdup tanks, and chemical tank) are located within the auxiliary building, which is well 11.2-3 Revision 1
ding floor drain systems are to the liquid radwaste system. This eliminates the potential for etected tank leakage to the environment and supports compliance with 10 CFR 20.1406 ference 5).
liquid radwaste system is designed so that the annual average concentration limits established 0 CFR 20 (Appendix B, table 2, column 2) (Reference 1) for liquid releases are not exceeded ng plant operation. Subsection 11.2.3 describes the calculated releases of radioactive materials the liquid radwaste system and other portions of the liquid waste management systems lting from normal operation.
monitored radwaste discharge pipeline is engineered to preclude leakage to the environment.
pipe is routed from the auxiliary building to the radwaste building (the short section of pipe ween the two buildings is fully available for visual inspection as noted above) and then out of the waste building to the licensed release point for dilution and discharge. The discharge radiation itor and isolation valve are located inside the radiologically controlled area. The exterior piping is igned to preclude inadvertent or unidentified releases to the environment; it is either enclosed in a guard pipe and monitored for leakage, or accessible for visual inspection. No valves or uum breakers are incorporated outside of monitored structures. This greatly reduces the potential undetected leakage from this discharge to the environment at a non-licensed release point, and ports compliance with 10 CFR 20.1406 (Reference 5).
exterior liquid radwaste system discharge pipeline is routed from the Radwaste Building to the tern bank of the Broad River where this effluent mixes with the blowdown sump discharge. This rface point between the waste water and liquid radwaste systems is upstream of the plant outfall e Ninety-Nine Islands Reservoir via the outfall pipe/diffuser. The plant outfall is described in section 9.2.9.2.2.
exterior liquid radwaste system discharge pipe is stainless steel and is enclosed within a high-sity polyethylene guard pipe. No valves or vacuum breakers are incorporated in exterior radwaste lines outside of monitored structures. The annular space between the liquid radwaste discharge and the guard pipe is monitored for leakage at low points along the path. The guard pipe is tinuous up to the underground pit where the liquid radwaste pipe ties into the outfall pipe. The erground pit is monitored for leakage. Monitoring points are provided to facilitate manual pling for leakage consistent with NEI 08-08A and 10 CFR 20.1406 contamination minimization uirements. Leakage monitoring of the liquid radwaste system discharge pipeline and the erground pit where the liquid radwaste pipe ties into the outfall pipe will be implemented as part of radiation protection program (See Appendix 12AA).
.1.2.4.1 Abnormal Operation sections 11.2.1.2.2 and 11.2.1.2.3 describe the capability of the liquid radwaste system to ommodate abnormal conditions for various equipment and other anticipated operational urrences. During these anticipated occurrences, the effectiveness of the liquid radwaste system ontrolling releases of radioactivity remains unaffected, so releases are limited as during normal ration.
section 11.2.3 discusses the calculated releases of radioactive materials from the liquid radwaste em for abnormal situations.
11.2-4 Revision 1
liquid radwaste system equipment design parameters are provided in Table 11.2-2.
seismic design classification and safety classification for the liquid radwaste system components structures are listed in Section 3.2. The components listed are located in the Seismic Category I lear Island and in the radwaste building.
monitor tanks in the non-seismic radwaste building are used to store processed water. The oactivity content of processed water in each tank will be less than the A1 and A2 levels of CFR 71 Appendix A, Table A-1.
.1.2.5.2 Use of Mobile and Temporary Equipment liquid radwaste system is designed to handle most liquid effluents and other anticipated events g installed equipment. However, for events occurring at a very low frequency or producing ents not compatible with the installed equipment, temporary equipment may be brought into the waste building mobile treatment facility truck bays.
nections are provided to and from various locations in the liquid radwaste system to these mobile ipment connections. This allows the mobile equipment to be used in series with installed ipment, as an alternate to it with the treated liquids returned to the liquid radwaste system, or as ltimate disposal point for liquids that are to be removed from the plant site for disposal where.
use of temporary equipment is common practice in operating plants. The radwaste building truck s and laydown space for mobile equipment, in addition to the flexibility of numerous piping nections to the liquid radwaste system, allow the plant operator to incorporate mobile equipment n integrated fashion.
porary equipment is also used to clean up the condensate storage tank if it becomes taminated following steam generator tube leakage. This use of temporary equipment is similar to just described, except that the equipment is used in the yard rather than in the radwaste building k bays.
en mobile or temporary equipment is selected to process liquid effluents, the equipment design testing meets the applicable requirements of Regulatory Guide 1.143. When confirmed through pling that the radioactive waste contents result in an inventory on a mobile system that is below A2 quantity limit for radionuclides specified in Appendix A to 10 CFR Part 71, the liquid effluent be processed with the mobile liquid waste processing system in the Radwaste Building. When process sampling and controls indicate that A2 quantity limits may be exceeded by processing d effluent in the Radwaste Building, liquid waste is processed in the seismic Category I Auxiliary ding. Procedural controls also ensure that the total cumulative source term of unpackaged tes including liquid waste, wet waste, solid waste, gaseous waste, activated or contaminated als and components, and contaminated waste present at any time in the Radwaste Building is ed consistent with RG 1.143, Revision 2, unmitigated radiological release criteria, so that an itigated release, occurring over a two hour time period, would not result in a dose of greater than millirem at the protected area boundary, or an unmitigated exposure, occurring over a two hour period, would not result in a dose of greater than 5 rem to site personnel located 10 feet from the l cumulative radioactive inventory. The unmitigated, unshielded worker dose is calculated at eet from the source. Unlimited worker occupancy workstations and low dose rate waiting areas 11.2-5 Revision 1
ile and temporary equipment are designed in accordance with the applicable mobile and porary radwaste treatment systems guidance provided in Regulatory Guide 1.143, including the es and standards listed in Table 1 of the Regulatory Guide.
ile and temporary equipment have the following features:
Level indication and alarms (high-level) on tanks.
Screwed connections are permitted only for instrument connections beyond the first isolation valve.
Remote operated valves are used where operations personnel would be required to frequently manipulate a valve.
Local control panels are located away from the equipment, in low dose areas.
Instrumentation readings are accessible from the local control panels (i.e., temperature, flow, pressure, liquid level, etc.).
Wetted parts are 300 series stainless steel, except flexible hose and gaskets.
Flexible hose is used only for mobile equipment within the designated black box locations between mobile components and at the interface with the permanent plant piping.
The contents of tanks are capable of being mixed, either through recirculation or with a mixer.
Grab sample points are located in tanks and upstream and downstream of the process equipment.
ection and testing of mobile or temporary equipment is in accordance with the codes and dards listed in Table 1 of Regulatory Guide 1.143 with the following additions:
After placement in the station, the mobile or temporary equipment is hydrostatically, or pneumatically, tested prior to tie-in to permanent plant piping.
A functional test, using demineralized water, is performed. Remote operated valves are stroked (open-closed-open or closed-open-closed) under full flow conditions. The proper function of the instrumentation, including alarms, is verified. The operating procedures are verified correct during the functional test.
Tank overflows are routed to floor drains.
Floor drains are confirmed to be functional prior to placing mobile or temporary equipment into operation.
.1.3 Compliance with 10 CFR 20.1406 ccordance with the requirements of 10 CFR 20.1406 (Reference 5), the liquid radwaste system is igned to minimize, to the extent practicable, contamination of the facility and the environment, itate decommissioning, and minimize, to the extent practicable, the generation of radioactive te. This is done through appropriate selection of design technology for the system, and 11.2-6 Revision 1
.2 System Description liquid radwaste system, shown in Figure 11.2-1, includes tanks, pumps, ion exchangers, and rs. The liquid radwaste system is designed to process, or store for processing by mobile ipment, radioactively contaminated wastes in four major categories:
Borated, reactor-grade, waste water - this input is collected from the reactor coolant system (RCS) effluents received through the chemical and volume control system (CVS), primary sampling system sink drains and equipment leakoffs and drains.
Floor drains and other wastes with a potentially high suspended solids content - this input is collected from various building floor drains and sumps.
Detergent wastes - this input comes from the plant hot sinks and showers, and some cleanup and decontamination processes. It generally has low concentrations of radioactivity.
Chemical waste - this input comes from the laboratory and other relatively small volume sources. It may be mixed hazardous and radioactive wastes or other radioactive wastes with a high dissolved-solids content.
radioactive secondary-system waste is not processed by the liquid radwaste system. Secondary-em effluent is normally handled by the steam generator blowdown processing system, as cribed in Subsection 10.4.8, and by the turbine building drain system.
ioactivity can enter the secondary systems from steam generator tube leakage. If significant oactivity is detected in secondary-side systems, blowdown is diverted to the liquid radwaste em for processing and disposal.
.2.1 Waste Input Streams
.2.1.1 Reactor Coolant System Effluents effluent subsystem receives borated and hydrogen-bearing liquid from two sources: the reactor lant drain tank and the chemical and volume control system. The reactor coolant drain tank ects leakage and drainage from various primary systems and components inside containment.
ent from the chemical and volume control system is produced mainly as a result of reactor lant system heatup, boron concentration changes and RCS level reduction for refueling.
t collected by the effluent subsystem normally contains hydrogen and dissolved radiogases.
refore, it is routed through the liquid radwaste system vacuum degasifier before being stored in effluent holdup tanks.
liquid radwaste system degasifier can also be used to degas the reactor coolant system before tdown by operating the chemical and volume control system in an open loop configuration. This is e by taking one of the effluent holdup tanks out of normal waste service and draining it. Then mal chemical and volume control system letdown is directed through the degasifier to the icated effluent holdup tank. From there, it is pumped back to the suction of the chemical and me control system makeup pumps with the effluent holdup tank pump. The makeup pumps rn the fluid to the reactor coolant system in the normal fashion. This process is continued as essary for degassing the reactor coolant system as described in Subsection 9.3.6.
11.2-7 Revision 1
trolled to maintain an essentially fixed tank level to minimize tank pressure variation.
ctor coolant system effluents from the chemical and volume control system letdown line or the tor coolant drain subsystem pass through the vacuum degasifier, where dissolved hydrogen and on gases are removed. These gaseous components are sent via a water separator to the eous radwaste system. A degasifier discharge pump then transfers the liquid to the currently cted effluent holdup tank. If flows from the letdown line and the reactor coolant drain tank are ed to the degasifier concurrently, the letdown flow has priority and the drain tank input is matically suspended.
e event of abnormally high degasifier water level, inputs are automatically stopped by closing the own control and containment isolation valves.
effluent holdup tanks vent to the radiologically controlled area ventilation system and, in ormal conditions, may be purged with air to maintain a low hydrogen gas concentration in the s' atmosphere. Hydrogen monitors are included in the tanks vent lines to alert the operator of ated hydrogen levels.
contents of the effluent holdup tanks may be recirculated and sampled, recycled through the asifier for further gas stripping, returned to the reactor coolant system via the chemical and me control system makeup pumps, discharged to the mobile treatment facility, processed ugh the ion exchangers, or directed to the monitor tanks for discharge without treatment.
cessing through the ion exchangers is the normal mode.
AP1000 liquid radwaste system processes waste with an upstream filter followed by four ion hange resin vessels in series. Any of these vessels can be manually bypassed and the order of last two can be interchanged, so as to provide complete usage of the ion exchange resin.
top of the first vessel is normally charged with activated carbon, to act as a deep-bed filter and ove oil from floor drain wastes. Moderate amounts of other wastes can also be routed through vessel. It can be bypassed for processing of relatively clean waste streams. This vessel is ewhat larger than the other three, with an extra sluice connection to allow the top bed of vated carbon to be removed. This feature is associated with the deep bed filter function of the sel; the top layer of activated carbon collects particulates, and the ability to remove it without urbing the underlying zeolite bed minimizes solid-waste production.
second, third and fourth beds are in identical ion exchange vessels, which are selectively loaded resin, depending on prevailing plant conditions.
r deionization, the water passes through an after-filter where radioactive particulates and resin s are removed. The processed water then enters one of the monitor tanks. When one of the itor tanks is full, the system is automatically realigned to route processed water to another tank.
contents of the monitor tank are recirculated and sampled. In the unlikely event of radioactivity in ess of operational targets, the tank contents are returned to a waste holdup tank for additional essing.
mally, however, the radioactivity will be well below the discharge limits, and the dilute boric acid is harged for dilution to the circulating water blowdown. The discharge flow rate is set to limit the c acid concentration in the circulating water blowdown stream to an acceptable concentration for 11.2-8 Revision 1
down is not available for the discharge path.
.2.1.2 Floor Drains and Other Wastes with Potentially High Suspended Solid Contents entially contaminated floor drain sumps and other sources that tend to be high in particulate ing are collected in the waste holdup tank. Additives may be introduced to the tank to improve tion and ion exchange processes. Tank contents may be recirculated for mixing and sampling.
tanks have sufficient holdup capability to allow time for realignment and maintenance of the ess equipment.
waste water is processed through the waste pre-filter to remove the bulk of the particulate ing. Next it passes through the ion exchangers and the waste after-filter before entering a itor tank. The monitor tank contents are sampled and, if necessary, returned to a waste holdup or recirculated directly through the filters and ion exchangers.
ste water meeting the discharge limits is discharged to the circulating water blowdown through a ation detector that stops the discharge if high radiation is detected.
.2.1.3 Detergent Wastes detergent wastes from the plant hot sinks and showers contain soaps and detergents.
se wastes are generally not compatible with the ion exchange resins described in sections 11.2.2.1.1 and 11.2.2.1.2. The detergent wastes are not processed and are collected e chemical waste tank. If the detergent wastes activity is low enough, the wastes can be harged without processing.
en sufficient detergent wastes are produced and processing is necessary, mobile processing ipment is brought into one of the radwaste building mobile systems facility truck bays provided for purpose.
.2.1.4 Chemical Wastes ts to the chemical waste tank normally are generated at a low rate. These wastes are only ected; no internal processing is provided. Chemicals can be added to the tank for pH or other stment. Since the volume of these wastes is low, they can be treated by the use of mobile ipment or by shipment offsite.
.2.1.5 Steam Generator Blowdown am generator blowdown is normally accommodated within the steam generator blowdown em, which is described in Subsection 10.4.8.
eam generator tube leakage results in significant levels of radioactivity in the steam generator down stream, this stream is redirected to the liquid radwaste system for treatment before ase. In this event, one of the waste holdup tanks is drained to prepare it for blowdown processing.
blowdown stream is brought into that holdup tank, and continuously or in batches pumped ugh the waste ion exchangers. The number of ion exchangers in service is determined by the rator to provide adequate purification without excessive resin usage. The blowdown is then ected in a monitor tank, sampled, and discharged in a monitored fashion.
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b sampling taps are provided where required to monitor influent boron and radioactivity centrations; to monitor performance of various components; to determine tank water racteristics before transfer, processing or discharge; to verify performance of the on-line lyzers; and to collect samples of discharges to the environs for analysis and documentation.
ples are taken in low radiation areas.
.2.2.2 Tank Cleaning aordinary measures for tank cleaning are not normally required because the pumps take suction the low point of the tank, and the tank bottoms are sloped so that the tank can be fully drained.
irculation connections are provided to allow the tanks to be effectively mixed. Also, the air-rated double-diaphragm pumps used can pump air, water or slurries without damage, and can dry to clear the bottoms of the tanks.
visions are made for tank cleaning using a portable tank cleaning rig. Suction is taken from the bottom via a temporary hose. The pump discharge passes through a filter and the hose to a tank ning lance, which is manually inserted through a manway on the tank. The operator can direct high-velocity water throughout the inside of the tank.
.2.3 Component Description general descriptions and summaries of the design basis requirements for the liquid radwaste em components follow. Table 11.2-2 contains the operating parameters for the liquid radwaste em components.
itional information regarding the applicable codes and classifications is also available in tion 3.2.
.2.3.1 Liquid Radwaste System Pumps ctor Coolant Drain Tank Pumps full-capacity, stainless steel, reactor coolant drain tank pumps recirculate the reactor coolant n tank contents for cooling and to discharge the reactor coolant drain tank contents to the asifier or to an effluent holdup tank. These vertical sump pumps have permanently lubricated rings and mechanical seals. The pumps start and stop on high and low level.
ntainment Sump Pumps full-capacity containment sump pumps are provided. These pumps discharge the containment p contents to the waste holdup tank. These submersible sump pumps have permanently icated bearings and mechanical seals. The pumps start and stop on high and low level.
asifier Vacuum Pumps stainless steel, full-capacity, liquid ring type, degasifier vacuum pumps maintain the degasifier at w pressure for efficient gas stripping.
se liquid ring pumps use water as the compressant. The water is recycled to minimize sumption. Excess water from vapor condensation is discharged to an effluent holdup tank.
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ess compressor water accumulation in the separator to an effluent holdup tank. The pumps start stop to share the duty. The pump is constructed of stainless steel and has a mechanical seal.
er Pumps following air-operated double-diaphragm pumps are mounted near the associated tanks with rnal suction piping. Construction is of stainless steel, with elastomeric diaphragms.
Degasifier discharge pumps (2)
Effluent holdup tank pumps (2)
Waste holdup tank pumps (2)
Monitor tank pumps (6)
Chemical waste tank pump (1)
.2.3.2 Liquid Radwaste System Heat Exchangers ctor Coolant Drain Tank Heat Exchanger horizontal U-tube heat exchanger is provided. The heat exchanger has a flanged tubesheet that mits removal of the tube bundle for inspection and cleaning.
heat exchanger is designed to prevent the reactor coolant drain tank contents from boiling with leakage influent as shown in Table 11.2-4.
reactor coolant drain tank contents flow through the tubes which are stainless steel component ling water flows through the carbon steel shell.
or Condenser horizontal U-tube heat exchanger assists in drying the gases drawn out of the liquid waste by vacuum pump, before they are sent to the gaseous radwaste system. As the gas bearing water cades down through the packing in the degasifier vessel, it boils in the low pressure. To minimize size of the vacuum pumps, a vapor condenser is provided between the degasifer vessel and the uum pumps. In the vapor condenser, most of the water vapor is condensed out of the gas stream re it enters the vacuum pump. The vapor condenser is cooled by chilled water. Chilled water s through the tubes, which are stainless steel. Water vapor condenses on the tubes and drains ugh a subcooling section in the stainless steel shell. The non-condensible gases and condensate recombined in a common pipe leading to the suction of the liquid ring type vacuum pumps.
.2.3.3 Liquid Radwaste System Tanks ctor Coolant Drain Tank reactor coolant drain tank is provided. The tank is sized to accommodate two vertical sump type ps and to have a volume above the normal operating water level sufficient to accept the influent shown in Table 11.2-4.
reactor coolant drain tank is a stainless steel, horizontal, cylindrical tank with dished heads. It is ided with a vacuum breaker to prevent excess external pressure during containment leak testing.
protected from excess internal pressure by a relief valve which vents to the containment sump.
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asifier Column e-stage, stainless steel degasifier column is provided. The degasifier column is designed to meet performance parameters shown in Table 11.2-5.
ation and surface exposure are accomplished by spraying the influent onto the top of a column of king which breaks up the flow and spreads it into thin films as it cascades downward. The low sure causes the inlet water to boil. The flashed vapor accompanies the gas bearing water nward through the packing. Exposure to low pressure draws out the non-condensible gases sistent with Henry's Law and they pass out the vacuum connection. The vacuum connection is ted near the last point of contact with the degassed water where the vacuum is greatest and ditions are least conducive to reabsorption. A stainless steel mesh demister is provided at the sel vacuum connection to remove water droplets which are entrained in the gas/vapor mixture as exiting to the vapor condenser.
asifier Separator stainless steel separator is provided. It is designed to remove compressor water from the uum pump discharge flow for reuse. It also serves as a silencer.
uent Holdup Tanks se stainless steel tanks contain effluent waste prior to processing. They are horizontal cylinders internal pump suction piping at the low point of the tank, and with side manways for ntenance.
ste Holdup Tanks se stainless steel tanks contain floor and equipment drain waste before processing. They are ical cylinders with internal pump suction piping at the low points of the tanks and with side ways for maintenance.
nitor Tanks se stainless steel tanks contain processed waste before discharge. They are vertical cylinders internal pump suction piping at the low points of the tanks and with side manways for ntenance.
mical Waste Tank stainless steel tank contains chemical waste and hot sinks and shower drains before processing mobile equipment. The configuration is a vertical cylinder with internal pump suction piping at the point of the tank and with a side manway for maintenance.
.2.3.4 Liquid Radwaste System Ion Exchangers r ion exchange vessels are provided, with resin volumes as shown in Table 11.2-2. The media will elected by the plant operator to optimize system performance. The ion exchange vessels are nless steel, vertical, cylindrical pressure vessels with inlet and outlet process nozzles plus nections for resin addition, sluicing, and draining. The process outlet and flush water outlet nections are equipped with resin retention screens designed to minimize pressure drop.
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filter is provided to collect particulate matter in the process stream before ion exchange. The unit onstructed of stainless steel and has disposable filter bags.
ste After-Filter filter is provided downstream of the ion exchangers to collect particulate matter, such as resin
- s. The unit is constructed of stainless steel and has disposable filter cartridges.
.2.4 Instrumentation Design rumentation readout is available in the main control room and on portable display and control els.
ms are provided to the data display system including a radwaste system annunciator in the main trol room.
ssure indicators provide pressure drops across demineralizers, filters, and strainers.
eases to the environment are monitored for radioactivity. Section 11.5 describes this rumentation.
h tank is provided with level instrumentation that actuates an alarm on high liquid level in the
, thus warning of potential tank overflow. High level in redundant tank pairs also diverts the flow e standby tank. Table 11.2-3 provides a summary of the tank level alarms.
.2.5 System Operation and Performance
.2.5.1 Reactor Coolant System Effluent Processing
.2.5.1.1 Reactor Coolant Systems Effluent: Letdown Line mical and volume control system letdown is directed to the degasifier. This letdown flow matically takes priority by causing isolation of influent to the degasifier from the reactor coolant n tank pumps to prevent the design capacity of the degasifier from being exceeded.
en the degasifier and waste gas system are placed in operation one of the degasifier vacuum ps operates to maintain a vacuum in the degasifier column. The degasifier separator pump rates to return compressor water to the vacuum pump. The degasifier separator vents to the eous radwaste system. Its level is automatically controlled by discharging excess water (due to densation of vapor carryover from the degasifier column) to an effluent holdup tank. In the event bnormally high level, chemical and volume control system letdown flow is automatically stopped.
effluent holdup tanks are provided. One is aligned to receive inputs. When it fills to the ropriate level, an alarm alerts the operator that the tank is full and ready for processing. The inlet rsion valve automatically realigns the system to route input to the other tank upon high-high m.
.2.5.1.2 Reactor Coolant System Effluent: Reactor Coolant Drain Tank reactor coolant drain tank receives input from the reactor coolant system and other drains inside tainment that have the potential to contain radioactive gas or hydrogen.
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matically controlled to maintain reactor coolant drain tank water level within a narrow band to imize tank pressure variation. An alarm alerts the operator if the reactor coolant drain tank hes a temperature consistent with the design leak of saturated RCS coolant. The system matically realigns valves and recirculates the tank contents through the reactor coolant drain heat exchanger.
cumulative quantity discharged from the reactor coolant drain tank is totalized and indicated for in reactor coolant leakage evaluations.
discharge may have a relatively high dissolved hydrogen concentration and is therefore aligned e degasifier. However, during reactor coolant system loop drain operations the hydrogen and oactive gas concentrations should be low and discharge may be directly aligned to an effluent up tank.
.2.5.1.3 Processing of the Reactor Coolant System Effluents h effluent holdup tank vent includes a hydrogen detector to monitor the hydrogen concentration in tank atmosphere. In the event of high alarm, the operator initiates air purge through the tank to e the hydrogen gas and maintain it below the flammable limits. The tanks vent to the ologically controlled area ventilation system.
effluent holdup tank high level alarm alerts the operator that the tank is full and ready for essing. The inlet diversion valve automatically directs the influent to the other tank upon high-alarm.
rocess the contents of the filled tank, the effluent holdup tank pump is started to recirculate and ple the tank contents. If additional gas stripping is required, the tank contents may be recirculated ugh the degasifier. The degasifier functions automatically as described in section 11.2.2.5.1.1.
discharge of either effluent holdup tank pump can be aligned to the suction of the chemical and me control system makeup pumps. This mode of operation is used during reactor coolant system assing operations. Reactor coolant from the chemical and volume control system letdown is assed in the degasifier, collected in one of the effluent holdup tanks, and continuously pumped k to the chemical and volume control system makeup pumps. The pump returns the degassed er to the reactor coolant system.
ctor coolant collected in an effluent holdup tank during reactor coolant system loop drain rations may also be pumped to the chemical and volume control system makeup pumps for refill e reactor coolant system. Before beginning this process, the operator fully drains the effluent up tank receiving the reactor coolant so that the boron concentration of the reactor coolant em is not significantly affected.
effluent may be transferred to the mobile treatment facility for concentration or solidification. This osal method is used only during unusual conditions that restrict the normal processed waste harge mode described in the following paragraphs.
normal mode of operation is to process the effluent by ion exchange and filtration to remove the oactive materials. The ion exchangers operate in series as described in Subsection 11.2.2.1.1.
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en the analysis of samples taken periodically downstream of the ion exchange processing cates an increase in radioactivity above prescribed limits, the operator isolates the expended (s) for resin replacement. Flow continues through the other units until a fresh resin bed is ready.
en one of the last two ion exchangers has been replenished, the fresh unit is then brought online he downstream unit.
after-filter removes resin fines and other particulate matter that may pass through the ion hangers. A high differential pressure alarm alerts the operator to the need for filter element acement. Normally, filter element replacement is initiated on high radioactivity determined by odic survey.
cess discharge is normally aligned to one of the monitor tanks. When one of the tanks is full, an m alerts the operator that the tank is full and ready to be discharged. The inlet diversion valve matically realigns the system to route processed waste to another tank upon high-high level.
operator then starts the monitor tank pump to recirculate the tank contents and samples the essed waste. Since the ion exchangers operate in the borated saturated mode, the water tains boric acid. The radioactivity and chemistry of the processed waste is determined by sample lysis. In the unlikely event that radioactivity exceeds discharge limitations, the tank contents are rned to a waste holdup tank for reprocessing.
e it is confirmed that the waste water is within radioactivity discharge limitations, the operator pares the system for discharge. The operator initiates discharge by starting the monitor tank pump opening the remotely operated discharge valve. During controlled discharge, grab samples are n for laboratory analysis and documentation of discharge.
e radiation monitor in the discharge line detects high radiation, the valve automatically closes.
operator is alerted to this condition by a high radiation alarm, and is required to take corrective on. A manual drain valve is opened to flush the radiation monitor and confirm low radiation before stablishing discharge to the circulating water blowdown. Low monitor tank level automatically s the monitor tank pump.
.2.5.2 Floor Drain and Equipment Drain Waste Processing cellaneous liquid wastes normally include influent from the radioactive floor drains, equipment ns and auxiliary building sump and excess water from the solid radwaste system. These wastes ect in one of two waste holdup tanks.
gh level alarm in the tank alerts the operator that the tank is full and ready to be processed. The diversion valve automatically directs influents to the second waste holdup tank upon high-high
- l. The waste holdup tank pump is started to recirculate and sample the tank contents. Additives be introduced to the waste holdup tank to optimize filtration and ion exchange processes.
r drain wastes are also brought into the waste holdup tanks from the containment sump. High p level automatically opens the containment isolation valves and starts a pump to transfer the p contents. Low level automatically stops the pump and closes the isolation valves. An alarm is ided to alert the operator to abnormally high containment sump level and the standby pump is matically started. Cumulative flow is totalized and indicated to support reactor coolant leakage lysis.
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.2.5.3 Detergent Waste Processing detergent wastes from the plant hot sinks and showers are routed to the chemical waste tank.
mally, these wastes are sampled and confirmed suitable for discharge without processing. If essing prior to discharge is necessary, three courses of action are available. The waste water be transferred to a waste holdup tank and processed in the same manner as other radioactively taminated waste water. If the onsite processing capabilities are not suitable for the composition of detergent waste, processing can be performed using mobile equipment brought into one of the k bays of the radwaste building or the waste water can be shipped offsite for processing. After essing by mobile equipment the water may be transferred to a waste holdup tank for further essing by the onsite equipment or transferred to a monitor tank for sampling and discharge.
.2.5.4 Chemical Waste Processing ioactively contaminated chemical wastes are collected in the chemical waste tank. Chemicals be added to the tank for pH or other adjustment. The volume of these wastes is expected to be The design includes alternatives for processing or discharge of chemical wastes. They may be essed onsite without being combined with other wastes using mobile equipment. When bined with detergent wastes, they may be suitable for discharge without treatment or for essing by onsite equipment before discharge. When not suitable for onsite processing, they can reated using mobile equipment or shipped offsite for processing. After processing by mobile ipment the water may be transferred to a waste holdup tank for further processing by the onsite ipment or transferred to a monitor tank for sampling and discharge.
.2.5.5 Steam Generator Blowdown Processing mal steam generator blowdown processing is accommodated by the steam generator blowdown em, which is described in Subsection 10.4.8.
eam generator tube leakage results in levels of radioactivity in the blowdown stream above what be accommodated by the secondary-side systems, this stream is directed to the liquid radwaste em. For this function, the operator aligns the steam generator blowdown system to the inlet of the te holdup tank. The blowdown waste is then processed in the same way as other wastes.
.2.5.6 Ion Exchange Media Replacement initial and subsequent fill of ion exchange media is made through a resin fill nozzle on the top of ion exchange vessel. When the media are spent and ready to be transferred to the solid radwaste em, the vessel is isolated from the process flow. The flush water line is opened to the sluice ng and demineralized water is pumped into the vessel through the normal process outlet nection upward through the media retention screen. The media fluidize in the upward, reverse
. When the bed has been fluidized, the sluice connection is opened and the bed is sluiced to the nt resin tanks in the solid radwaste system (WSS). Demineralized water flow continues until the has been removed and the sluice lines are flushed clean of spent resin.
.3 Radioactive Releases id waste is produced both on the primary side (primarily from adjustment of reactor coolant boron centration and from reactor coolant leakage) and the secondary side (primarily from steam erator blowdown processing and from secondary side leakage). Primary and secondary coolant 11.2-16 Revision 1
ept for reactor coolant system degasification in anticipation of shutdown, the AP1000 does not cle primary side effluents for reuse. Primary effluents are discharged to the environment after essing. Fluid recycling is provided for the steam generator blowdown fluid which is normally rned to the condensate system.
only liquid effluent site interface parameter outside of the Westinghouse scope is the release t to the Broad River at the Ninety-Nine Islands Dam.
.3.1 Discharge Requirements release of radioactive liquid effluents from the plant may not exceed the concentration limits cified in Reference 1 nor may the releases result in the annual offsite dose limits specified in CFR 50, Appendix I (Reference 2) being exceeded.
.3.2 Estimated Annual Releases annual average release of radionuclides from the plant is determined using the PWR-GALE e (Reference 3). The PWR-GALE code models releases which use source terms derived from obtained from the experience of operating PWRs. The code input parameters used in the lysis to model the AP1000 plant are listed in Table 11.2-6. The annual releases for a single-unit are presented in Table 11.2-7.
greement with Reference 3, the total releases include an adjustment factor of 0.16 curies per r to account for anticipated operational occurrences. The adjustment uses the same distribution uclides as the calculated releases.
.3.3 Dilution Factor dilution factor provided for the activity released is site dependent; the value of 6000 gpm used ein is based on cooling tower blowdown requirements and is expected to be conservatively low.
plant operator will select dilution flow rates to ensure that the effluent concentration limits of CFR Part 20, the annual offsite dose limits in 10 CFR 50 Appendix I, and any local requirements continuously met. If the available dilution is low, the discharge rate can be reduced to maintain eptable concentrations.
required dilution flow is dependent on the liquid waste discharge rate and, while the monitor tank ps have a design flow rate of 100 gpm, the discharge flow is controlled to be compatible with the ilable dilution flow. With a typical liquid waste release of 1925 gallons per day, the nominal ulating water blowdown flow of 6000 gpm provides sufficient dilution flow to maintain the annual rage discharge concentrations well below the effluent concentration limits. Actual plant operation ependent on the waste liquid activity level and the available dilution flow.
ffuser pipe upstream of the Ninety-Nine Islands Dam is the discharge point for the plant liquid ological effluent. The diffuser pipe mixes the effluent with the Ninety-Nine Islands Reservoir, ch acts as an impoundment, as described in Regulatory Guide 1.113. The annual average rates for the liquid radwaste effluent and the Broad River at the Ninety-Nine Islands Dam are d in the dose calculations. The dilution factors for points downstream of the dam are set at one.
conservatively assumes that no additional dilution occurs other than the dilution that takes place tream of the dam.
summary of parameters used in the impoundment model is presented in Table 11.2-201.
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onstrate compliance with the Reference 1 effluent concentration limits, the discharge centrations have been evaluated for the release of a typical daily liquid waste volume of 5 gallons per day and using the nominal circulating water blowdown flow of 6000 gpm.
le 11.2-8 lists the annual average nuclide release concentrations and the fraction of the effluent centration limits using base GALE code assumptions. As shown in Table 11.2-8, the overall tion of the effluent concentration limit is 0.11, which is well below the allowable value of 1.0.
annual releases from the plant have also been evaluated based on operation with the maximum ned fuel defect level. The maximum defined fuel defect level corresponds to the Technical cification limit on coolant activity which is based on 0.25 percent fuel defects. Table 11.2-9 lists annual average nuclide release concentrations and the fractions of the effluent concentration s for the maximum defined fuel defects. As shown in Table 11.2-9, the overall fraction of the ent concentration limit for the maximum defined fuel defect level is 0.53, which is well below the wable value of 1.0.
.3.5 Estimated Doses e and dose rate to man was calculated using the LADTAP II computer code. This code is based he methodology presented in Regulatory Guide 1.109.
tors common to both estimated individual dose rates and estimated population dose are ressed here. Unique data are discussed in the respective sections.
vity pathways considered are drinking water, sport fishing, and recreational activities.
nearest drinking water takeoff downstream of the Lee site is approximately 21 miles downstream nion, South Carolina.
.3.5.1 Estimated Individual Dose Rate e rates to individuals are calculated for drinking water, fish consumption, and recreational vities.
le 11.2-202 contains LADTAPII input data for dose rate calculations.
le 11.2-203 gives the maximum individual dose rates.
maximum doses to individuals resulting from routine liquid effluents per unit are presented and pared to the regulatory criteria set forth in 10 CFR Part 50, Appendix I, and 10 CFR 20.1301 in le 11.2-207 and Table 11.2-208, respectively.
maximum doses to individuals resulting from routine liquid effluents per unit are given in le 11.2-203. These doses are multiplied by two (2) to account for both units at the site. The total imum doses for both units are summarized in Table 11.2-205 for comparison to the regulatory s set forth in 40 CFR Part 190.
annual doses to a maximally exposed individual from gaseous effluents are given in le 11.3-207. The total site dose compared with the 40 CFR Part 190 criteria is provided in le 11.2-206. The liquid effluent doses per unit presented in Table 11.2-203 are added to the 11.2-18 Revision 1
section 12.4.2. Direct radiation from containment and other plant buildings is negligible for the es in Table 11.2-206. Additionally, there is no contribution from refueling water because the eling water is stored inside the containment instead of in an outside storage tank. In addition, e is no outside storage of solid radwaste. There are no radiation sources outside of the manent plant structures. There are no other uranium fuel cycle facilities in the vicinity of the site would contribute to the dose received by the maximally exposed individual. Thus, only the dose effluent released from the site and direct radiation from the site need be considered.
.3.5.2 Estimated Population Dose population dose is based on the fraction of the 50-mile population that will be exposed to the luated pathways. These pathways are drinking water, recreational activities, and sport fishing.
sport fishing harvest given in Table 11.2-202 is estimated using observations of usage and South olina Department of Natural Resources harvest limits for the number and size applicable to the ad River in the vicinity of the Lee Nuclear Station site.
reational activities considered are swimming, boating, and shoreline use. The annual usage for h of these activities, which is given in Table 11.2-202, is based on consideration of the eational value of the Broad River, the population in the vicinity of the Broad River downstream of Nuclear Station, and the average shoreline exposure time per person for each age group taken Table E-4 of Regulatory Guide 1.109. The population doses are shown in Table 11.2-204.
.3.5.3 Liquid Radwaste Cost Benefit Analysis Methodology application of the methodology of Regulatory Guide 1.110 was used to satisfy the cost benefit lysis requirements of 10 CFR Part 50, Appendix I, Section II.D. The parameters used in ulating the Total Annual Cost (TAC) are fixed and are given for each radwaste treatment system ment listed in Regulatory Guide 1.110, including the Annual Operating Cost (AOC) (Table A-2),
ual Maintenance Cost (AMC) (Table A-3), Direct Cost of Equipment and Materials (DCEM) le A-1), and Direct Labor Cost (DLC) (Table A-1). The following variable parameters were used:
Capital Recovery Factor (CRF) - This factor is taken from Table A-6 of Regulatory Guide 1.110 and reflects the cost of money for capital expenditures. A cost-of-money value of 7% per year is assumed in this analysis, consistent with the Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission (NUREG/BR-0058). A CRF of 0.0806 was obtained from Table A-6.
Indirect Cost Factor (ICF) - This factor takes into account whether the radwaste system is unitized or shared (in the case of a multi-unit site) and is taken from Table A-5 of Regulatory Guide 1.110. It is assumed that the radwaste system for this analysis is a unitized system at a 2-unit site, which equals an ICF of 1.625.
Labor Cost Correction Factor (LCCF) - This factor takes into account the differences in relative labor costs between geographical regions and is taken from Table A-4 of Regulatory Guide 1.110. A LCCF of 1.0 (the lowest value) is assumed in this analysis.
endix I to 10 CFR Part 50 prescribes a $1,000 per person-rem criterion for determining the cost efit of actions to reduce radiation exposure.
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shold value. The lowest-cost option for liquid radwaste treatment system augments is a 20 gpm tridge Filter at $11,140 per year, which yields a threshold value of 11.14 person-rem total body or oid dose from liquid effluents.
AP1000 sites with population dose estimates less than 11.14 person-rem total body or thyroid e from liquid effluents, no further cost-benefit analysis is needed to demonstrate compliance with CFR 50, Appendix I Section II.D.
.3.5.4 Liquid Radwaste Cost Benefit Analysis total body and thyroid population doses for liquid effluents given in Table 11.2-204 are a small tion of the threshold dose of 11.14 person-rem. Thus, no further cost-benefit analysis is needed to onstrate compliance with 10 CFR 50, Appendix I Section II.D.
.3.6 Quality Assurance quality assurance program for design, fabrication, procurement, and installation of the liquid waste system is in accordance with the overall quality assurance program described in pter 17.
e the impact of radwaste systems on safety is limited, the extent of control required by endix B to 10 CFR Part 50 is similarly limited. Thus, a supplemental quality assurance program licable to design, construction, installation and testing provisions of the liquid radwaste system is blished by procedures that comply with the guidance presented in Regulatory Guide 1.143.
.4 Preoperational Testing
.4.1 Sump Level Instrument Testing of the diverse methods of detecting small reactor coolant pressure boundary leaks is monitoring containment sump level. (See Subsection 5.2.5 for a full discussion.) A sump capacity calibration is performed so the containment sump level instruments can provide a display that is correlated e contained volume of water in the sump.
ddition to a normal level accuracy calibration of the containment sump level instruments, S-LT-034 and WLS-LT-035, their displays will be correlated to the volume of water during operational testing. A known volume of water will be added to the containment sump. The change ump level will be measured by marking the sump wall before and after the addition of water. The nge in the display of the sump level instruments will be compared to the level change measured he sump wall. A sump level change corresponding to a volume of water which is smaller than that ased in an hour by 0.5 gpm reactor coolant system leak can be detected.
.4.2 Discharge Control/Isolation Valve Testing AP1000 effluent discharge line includes a radiation monitor, WLS-RE-229, as described in section 11.5.2.3.3. A concentration of radioactivity in the effluent, which exceeds the radiation itor setpoint, causes a high radiation signal to automatically close the discharge control/isolation e.
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.4.3 Preoperational Inspection performance of the liquid radwaste system has been evaluated based upon using a determined quantity and type of ion-exchange media. An inspection will confirm that the proper me of media, as listed in Table 11.2-2, Component Data - Liquid Radwaste System, has been alled into the appropriate liquid radwaste system components, MV03 and MV04A/B/C.
.5 Combined License Information
.5.1 Liquid Radwaste Processing by Mobile Equipment mobile or temporary equipment used for storing or processing liquid radwaste is addressed in section 11.2.1.2.5.2.
.5.2 Cost Benefit Analysis of Population Doses application of the methodology of Regulatory Guide 1.110 is addressed in Subsection 11.2.3.5.3.
site specific cost-benefit analysis to address the requirements of 10 CFR 50, Appendix I, arding population doses due to liquid effluents is addressed in Subsections 11.2.3.3, 11.2.3.5,
.3.5.1, 11.2.3.5.2, and 11.2.3.5.4.
.5.3 Identification of Ion Exchange and Adsorbent Media types of liquid waste ion exchange and absorbent media to be used in the liquid radwaste em (WLS) are addressed in APP-GW-GLR-008 (Reference 6).
.5.4 Dilution and Control of Boric Acid Discharge planned discharge flow rate for borated wastes and controls for limiting the boric acid centration in the circulating water system blowdown is addressed in APP-GW-GLR-014 ference 7).
.6 References "Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage," 10 CFR Part 20, Appendix B, Issued by 58 FR 67657, April 28, 1995.
"Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion 'As Low As Is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," 10 CFR Part 50, Appendix I.
"Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," NUREG-0017, Revision 1, March 1985.
ANSI/ANS-55.6-1993, "Liquid Radioactive Waste Processing Systems for Light Water Reactor Plants."
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COL Items 11.2-3 and 11.3-2," Westinghouse Electric Company LLC.
APP-GW-GLR-014, "Closure of COL Items in DCD Chapter 11, Dilution and Control of Boric Acid Discharge," Westinghouse Electric Company LLC.
"Appendix A to Part 71 - Determination of A1 and A2," 10 CFR 71 Appendix A, Table A-1.
11.2-22 Revision 1
Collection Tank Expected and Sources Input Rate Activity Basis Disposition Effluent holdup tanks Filtered, demineralized, and discharged Chemical and volume control 159,000 gpy 100% of AP1000-specific system letdown reactor coolant calculations(b)
Leakage inside containment (to 10 gpd 167% of ANSI/ANS-55.6 reactor coolant drain tank) reactor coolant Leakage outside containment 80 gpd 100% of ANSI/ANS-55.6 (to effluent holdup tanks) reactor coolant Sampling drains 200 gpd 100% of ANSI/
reactor coolant ANS-55.6(a)
Waste holdup tank Filtered, demineralized and discharged Reactor containment cooling 500 gpd 0.1% of ANSI/ANS-55.6 reactor coolant Spent fuel pool liner leakage 25 gpd 0.1% of ANSI/ANS-55.6 reactor coolant Misc. drains 675 gpd 0.1% of ANSI/ANS-55.6 reactor coolant Detergent waste Filtered, monitored, and discharged. If necessary, processed with mobile equipment.
Hot shower 0 gpd 10-7 µCi/g ANSI/ANS-55.6 Hand wash 200 gpd 10-7 µCi/g ANSI/ANS-55.6 Equipment and area 40 gpd 0.1% of ANSI/ANS-55.6 decontamination reactor coolant Laundry Offsite laundry Chemical wastes 2 gpd reactor Estimate Processed with mobile coolant equipment tes:
ANSI/ANS-55.6 identifies sampling drains activity of 5 percent of reactor coolant; 100 percent is used as a conservative input for GALE Code analysis.
Average letdown for all normal reactor fuel cycle operations; initial heatup, dilutions and borations.
11.2-23 Revision 1
mps ntainment sump pumps Number 2 Type Submersible centrifugal Design pressure (psig) 15 external Design temperature (°F) 250 Design flow (gpm) 100 Material Stainless steel actor coolant drain tank pumps Number 2 Type Vertical sump type, centrifugal Design pressure (psig) 15 external Design temperature (°F) 250 Design flow (gpm) 100 Material Stainless steel gasifier separator pump rt of vacuum degasifier)
Number 2 Type Centrifugal Design pressure (psig) 125 Design temperature (°F) 200 Design flow (gpm) 7 Material Stainless steel 11.2-24 Revision 1
ndard waste processing pump Standard waste processing pump used for:(1)
Number Application 2 Degasifier discharge pumps 2 Effluent holdup tank pumps 2 Waste holdup tank pumps 6 Monitor tank pumps 1 Chemical waste tank pump Type Air-operated, double-diaphragm Design pressure (psig) 125 Design temperature (°F) 200 Design flow (gpm) 100 (can be varied by varying air supply flow)
Material Stainless steel body, Elastomeric diaphragm gasifer vacuum pumps rt of vacuum degasifier package)
Number 2 Type Liquid ring Design pressure (psig) 125 Design temperature (°F) 200 Design flow (scfm) 0.5 steady, 150 hogging Material Stainless steel 11.2-25 Revision 1
ste pre-filter Number 1 Type Disposable bag Design pressure (psig) 150 Design temperature (°F) 150 Design flow (gpm) 75 Particle size (micron, 98% retention) 25 Materials Housing Stainless steel Filter Polypropylene/pleated paper ste after-filter Number 1 Type Disposable bag or cartridge Design pressure (psig) 150 Design temperature (°F) 150 Design flow (gpm) 75 Particle size (micron, 98% retention) 0.5 Materials Housing Stainless steel Filter medium Polypropylene/pleated paper 11.2-26 Revision 1
actor Coolant drain tank heat exchanger Number 1 Type Horizontal U-tube Design pressure (psig) 150 tubeside, 200 shellside Design temperature (°F) 250 tubeside, 200 shellside Design flow (lb/hr) 48,700 tubeside, 62,200 shellside Heat Transfer Design Case Temperature inlet (°F) 175 tubeside, 95 shellside Temperature outlet (°F) 143 tubeside, 120 shellside Material SS tubeside, CS shellside por condenser Number 1 Type Horizontal U-tube Design pressure (psig) 150 Design temperature (°F) 150 Design flow (lb/hr) 100,000 tubeside, 1700 shellside Heat Transfer Design Case Temperature inlet (°F) 45 tubeside, 84 shellside Temperature outlet (°F) 63 tubeside, 60 shellside Material SS 11.2-27 Revision 1
ep bed filter Number 1 Design pressure (psig) 150 Design temperature (°F) 150 Design flow (gpm) 75 Nominal resin volume (ft3) 50 Material Stainless steel Resin type Layered: Activated charcoal on zeolite resin (Adjustable for plant conditions)
Process decontamination factors See Table 11.2-5 ste ion exchangers Number 3 Design pressure (psig) 150 Design temperature (°F) 150 Design flow (gpm) 75 Nominal resin volume (ft3) 30 Materials Stainless steel Resin type One cation, Two mixed (Adjustable for plant conditions)
Process decontamination factors See Table 11.2-5 11.2-28 Revision 1
actor coolant drain tank Number 1 Nominal volume (gal) 900 Type Horizontal Design pressure (psig) 10 internal, 15 external Material Stainless steel ntainment sump Number 1 Nominal volume (gal) 220 Type Rectangular Design pressure (psig) Atmospheric Design temperature (°F) 200 Material Stainless steel uent holdup tanks Number 2 Nominal volume (gal) 28,000 Type Horizontal Design pressure (psig) Atmospheric Design temperature (°F) 150 Material Stainless steel ste holdup tanks Number 2 Nominal volume (gal) 15,000 Type Vertical Design pressure (psig) Atmospheric Design temperature (°F) 150 Material Stainless steel 11.2-29 Revision 1
Number 6 Nominal volume (gal) 15,000 Type Vertical Design pressure (psig) Atmospheric Design temperature (°F) 150 Material Stainless steel emical waste tank Number 1 Nominal volume (gal) 8,900 Type Vertical Design pressure (psig) Atmospheric Design temperature (°F) 150 Material Stainless steel gasifier separator rt of vacuum degasifier package)
Number 1 Nominal volume (gal) 45 Type Vertical Design pressure (psig) 75 Design temperature (°F) 200 Material Stainless steel gasifier column rt of vacuum degasifier package)
Number 1 Nominal volume (gal) 900 Type Vertical Design pressure (psig) 75 internal 15 external Design temperature (°F) 150 Material Stainless steel This same pump is also used for other applications, such as sumps outside containment.
11.2-30 Revision 1
Level Indication Alarm Tank Location (Note 3) Location Alarm Overflow To uent holdup MCR MCR High Room drains to auxiliary building sump which is pumped to waste holdup tank (Note 2) ste holdup MCR MCR High Room (Note 4) emical waste MCR MCR High Room (Note 2) nitor MCR MCR High Room (Note 5) s:
MCR = main control room Room is piped to a floor drain within the auxiliary building, which is seismic Category I and water-tight with curbs or walls of sufficient height to contain the entire contents of the contained tank.
Monitoring of the liquid radwaste system is performed through the data display and processing system. Control functions are performed by the plant control system. Appropriate alarms and displays are available in the control room. Local indication and control are available on portable displays which may be connected to the data display and processing system. See Chapter 7.
Room is within the auxiliary building, which is seismic Category I and water-tight with curbs or walls of sufficient height to contain the entire contents of the contained tank.
Room is piped to a floor drain within the auxiliary building, which is seismic Category I and water-tight with curbs or walls of sufficient height to contain the entire contents of the contained tank, or to a floor drain within the radwaste building, which is water tight with curbs or walls of sufficient height to contain the entire contents of the contained tank.
11.2-31 Revision 1
actor Coolant Drain Tank Sized to accept 10 gpm of saturated reactor coolant for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without discharge or overflow.
Reactor coolant drain tank heat exchanger designed to limit the temperature to less than 175°F with this input assumed to be at 580°F.
ntainment Sump Sized to allow collection of 160 gallons of water between pumping cycles.
luent Holdup Tanks Sized to allow (together) a back-to-back plant shutdown and restart without delay at any time during the first 85 percent of core life. This operation requires nominal processing of the effluent monitor tanks and normal discharge with temporary storage of waste fluid in the cask loading pit.
Sized to allow (together) a single plant shutdown and restart without delay at any time during the first 80 percent of core life. This operation requires nominal processing to the monitor tanks, but no discharge from the plant.
her Tanks Sized based on accommodating maximum input without operator intervention for reasonable lengths of time.
11.2-32 Revision 1
contamination factors assumed per NUREG-0017, Revision 1 (PWR-GALE code input) to be as follows:
sin Type/Component Iodine Cs/Rb Other olite/deep bed filter (Note 1) 1 100 1 tion/waste ion exchanger 1 1 10 10 ed/waste ion exchanger 2 100 2 (Note 2) 100 ed/waste ion exchanger 3 10 10 (Note 2) 10 (Note 2) her components not directly involved in discharge from the plant:
Degasifier Column Reduce hydrogen by a factor of 40 Assuming inlet flow of 100 gpm at 130°F.
s:
This component is not included in NUREG-0017. DFs based upon "Reduction of Cesium and Cobalt Activity in Liquid Radwaste Processing Using Clinoptilolite Zeolite at Duke Power Company," by O.E. Ekechokwu, et al., Proc. Waste Management '92, Tucson, Arizona, March 1992, University of Arizona, Tucson.
Credit for this decontamination factor not taken in determination of anticipated annual releases.
11.2-33 Revision 1
rmal power level (MWt) 3400 s of primary coolant (lb) 4.35 x 105 ary system letdown rate (gpm) 100 down cation demineralizer flow rate, annual average (gpm) 10 mber of steam generators 2 l steam flow (lb/hr) 14.97 x 106 s of liquid in each steam generator (lb) 1.75 x 105 l blowdown rate (lb/hr) 4.2 x 104 wdown treatment method 0(1) densate demineralizer regeneration time N/A densate demineralizer flow fraction 0.33 ary coolant bleed for boron control leed flow rate (gpd) 435 econtamination factor for I 103 econtamination factor for Cs and Rb 103 econtamination factor for others 103 ollection time (day) 30 rocess and discharge time (day) 0 raction discharged 1.0 ipment Drains and Clean Waste quipment drains flow rate (gpd) 290 raction of reactor coolant activity 1.023 econtamination factor for I 103 econtamination factor for Cs and Rb 103 econtamination factor for others 103 ollection time (day) 30 rocess and discharge time (day) 0 raction discharged 1.0 11.2-34 Revision 1
irty waste input flow rate (gpd) 1200 raction of reactor coolant activity 0.001 econtamination factor for I 103 econtamination factor for Cs and Rb 103 econtamination factor for others 103 ollection time (day) 10 rocess and discharge time (day) 0 raction discharged 1.0 wdown Waste lowdown fraction processed 1 econtamination factor for I 100 econtamination factor for Cs and Rb 10 econtamination factor for others 100 ollection time N/A rocess and discharge time N/A raction discharged 0 enerant Waste N/A 11.2-35 Revision 1
ontinuous gas stripping of full letdown purification flow None Holdup time for xenon, (days) 38 Holdup time for krypton, (days) 2 Fill time of decay tanks for gas stripper N/A Gas waste system: HEPA filter None Auxiliary building: Charcoal filter None Auxiliary building: HEPA filter None Containment volume (ft3) 2.1 x 106 Containment atmosphere internal cleanup rate (ft3/min) N/A Containment high volume purge:
Number of purges per year (in addition to two shutdown 0 purges)
Charcoal filter efficiency (%) 90 HEPA filter efficiency (%) 99 Containment normal continuous purge rate (ft3/min) 500 (based on 20 hrs/week at 4000 ft3/min)
Charcoal filter efficiency (%) 90 HEPA filter efficiency (%) 99 Fraction of iodine released from blowdown tank vent N/A Fraction of iodine removed from main condenser air ejector 0.0 release ergent Waste Decontamination Factor 0.0(2) s:
A "0" is input to indicate that the blowdown is recycled to the condensate system after treatment in the blowdown system.
A "0.0" is input to indicate that the plant does not have an onsite laundry.
11.2-36 Revision 1
Turbine Combined Total Nuclide Shim Bleed Misc. Wastes Building Releases Releases(1)
Corrosion and Activation Products Na-24 0.00053 0.0(2) 0.00008 0.00061 0.00163 Cr-51 0.00068 0.0 0.0 0.00070 0.00185 Mn-54 0.00048 0.0 0.0 0.00049 0.00130 Fe-55 0.00037 0.0 0.0 0.00037 0.00100 Fe-59 0.00008 0.0 0.0 0.00008 0.00020 Co-58 0.00125 0.0 0.00001 0.00126 0.00336 Co-60 0.00016 0.0 0.0 0.00017 0.00044 Zn-65 0.00015 0.0 0.0 0.00015 0.00041 W-187 0.00004 0.0 0.0 0.00005 0.00013 Np-239 0.00008 0.0 0.0 0.00009 0.00024 Fission Products Br-84 0.00001 0.0 0.0 0.00001 0.00002 Rb-88 0.00010 0.0 0.0 0.00010 0.00027 Sr-89 0.00004 0.0 0.0 0.00004 0.00010 Sr-90 0.0 0.0 0.0 0.0 0.00001 Sr-91 0.00001 0.0 0.0 0.00001 0.00002 Y-91m 0.0 0.0 0.0 0.00001 0.00001 Y-93 0.00003 0.0 0.0 0.00004 0.00009 Zr-95 0.00010 0.0 0.0 0.00011 0.00023 Nb-95 0.00009 0.0 0.0 0.00009 0.00021 Mo-99 0.00028 0.0 0.00001 0.0003 0.00057 Tc-99m 0.00027 0.0 0.00001 0.00028 0.00055 Ru-103 0.00183 0.00001 0.00002 0.00185 0.00493 Rh-103m 0.00183 0.00001 0.00002 0.00185 0.00493 Ru-106 0.02729 0.00011 0.00021 0.02761 0.07352 Rh-106 0.02729 0.00011 0.00021 0.02761 0.07352 Ag-110m 0.00039 0.0 0.0 0.00039 0.00105 Ag-110 0.00005 0.0 0.0 0.00005 0.00014 Te-129m 0.00004 0.0 0.0 0.00005 0.00012 Te-129 0.00006 0.0 0.0 0.00006 0.00015 11.2-37 Revision 1
Turbine Combined Total Nuclide Shim Bleed Misc. Wastes Building Releases Releases(1)
Te-131m 0.00003 0.0 0.0 0.00003 0.00009 Te-131 0.00001 0.0 0.0 0.00001 0.00003 I-131 0.00512 0.00004 0.00015 0.00531 0.01413 Te-132 0.00009 0.0 0.0 0.00009 0.00024 I-132 0.00054 0.00001 0.00007 0.00062 0.00164 I-133 0.00211 0.00003 0.00038 0.00252 0.00670 I-134 0.00030 0.0 0.0 0.00031 0.00081 Cs-134 0.00370 0.00001 0.00002 0.00373 0.00993 I-135 0.00144 0.00002 0.00041 0.00187 0.00497 Cs-136 0.00023 0.0 0.0 0.00024 0.00063 Cs-137 0.00496 0.00001 0.00003 0.00500 0.01332 Ba-137m 0.00464 0.00001 0.00002 0.00468 0.01245 Ba-140 0.00203 0.00001 0.00003 0.00207 0.00552 La-140 0.00272 0.00002 0.00005 0.00279 0.00743 Ce-141 0.00003 0.0 0.0 0.00004 0.00009 Ce-143 0.00006 0.0 0.00001 0.00007 0.00019 Pr-143 0.00005 0.0 0.0 0.00005 0.00013 Ce-144 0.00117 0.0 0.00001 0.00119 0.00316 Pr-144 0.00117 0.0 0.00001 0.00119 0.00316 All others 0.00001 0.0 0.0 0.00001 0.00002 l 0.09398 0.00043 0.00182 0.09623 0.25623 ept tritium)
Tritium release = 1010 curies per year s:
The release totals include an adjustment of 0.16 Ci/yr added by PWR-GALE code to account for anticipated operational occurrences such as operator errors that result in unplanned releases.
An entry of 0.0 indicates that the value is less than 10-5 Ci/yr.
11.2-38 Revision 1
Expected Releases Effluent Concentration Limits Discharge Concentration Effluent Concentration Fraction of Nuclide (Ci/ml)(1) Limit (Ci/ml)(2) Concentration Limit Na-24 1.7E-10 5.0E-05 3.4E-06 Cr-51 1.9E-10 5.0E-04 3.9E-07 Mn-54 1.4E-10 3.0E-05 4.5E-06 Fe-55 1.0E-10 1.0E-04 1.0E-06 Fe-59 2.1E-11 1.0E-05 2.1E-06 Co-58 3.5E-10 2.0E-05 1.8E-05 Co-60 4.6E-11 3.0E-06 1.5E-05 Zn-65 4.3E-11 5.0E-06 8.6E-06 W-187 1.4E-11 3.0E-05 4.5E-07 Np-239 2.5E-11 2.0E-05 1.3E-06 Br-84 2.1E-12 4.0E-04 5.2E-09 Rb-88 2.8E-11 4.0E-04 7.1E-08 Sr-89 1.0E-11 8.0E-06 1.3E-06 Sr-91 2.1E-12 2.0E-05 1.0E-07 Y-91m 1.0E-12 2.0E-03 5.2E-10 Y-93 1.2E-11 2.0E-05 5.8E-07 Zr-95 2.9E-11 2.0E-05 1.5E-06 Nb-95 2.6E-11 3.0E-05 8.7E-07 Mo-99 8.4E-11 2.0E-05 4.2E-06 Tc-99m 8.0E-11 1.0E-03 8.0E-08 Ru-103 5.2E-10 3.0E-05 1.7E-05 Rh-103m 5.2E-10 6.0E-03 8.6E-08 Ru-106 7.7E-09 3.0E-06 2.6E-03 11.2-39 Revision 1
Discharge Concentration Effluent Concentration Fraction of Nuclide (Ci/ml)(1) Limit (Ci/ml)(2) Concentration Limit Ag-110m 1.1E-10 6.0E-06 1.8E-05 Te-129m 1.3E-11 7.0E-06 1.8E-06 Te-129 1.6E-11 4.0E-04 3.9E-08 Te-131m 9.4E-12 8.0E-06 1.2E-06 Te-131 3.1E-12 8.0E-05 3.9E-08 I-131 1.5E-09 1.0E-06 1.5E-03 Te-132 2.5E-11 9.0E-06 2.8E-06 I-132 1.7E-10 1.0E-04 1.7E-06 I-133 7.0E-10 7.0E-06 1.0E-04 I-134 8.5E-11 4.0E-04 2.1E-07 Cs-134 1.0E-09 9.0E-07 1.2E-03 I-135 5.2E-10 3.0E-05 1.7E-05 Cs-136 6.6E-11 6.0E-06 1.1E-05 Cs-137 1.4E-09 1.0E-06 1.4E-03 Ba-140 5.8E-10 8.0E-06 7.2E-05 La-140 7.8E-10 9.0E-06 8.6E-05 Ce-141 9.4E-12 3.0E-05 3.1E-07 Ce-143 2.0E-11 2.0E-05 9.9E-07 Pr-143 1.4E-11 2.5E-05 5.4E-07 Ce-144 3.3E-10 3.0E-06 1.1E-04 Pr-144 3.3E-10 6.0E-04 5.5E-07 H-3 1.1E-04 1.0E-03 1.1E-01 Total = 0.11 s:
Annual average discharge concentration based on release of average daily discharge for 292 days per year with 6000 gpm dilution flow.
Effluent concentration limits are from Reference 1.
11.2-40 Revision 1
Concentration Limits for Releases with Maximum Defined Fuel Defects Discharge Concentration Effluent Concentration Fraction of Nuclide (Ci/ml)(1) Limit (Ci/ml)(2) Concentration Limit Na-24 1.7E-10 5.0E-05 3.4E-06 Cr-51 1.6E-10 5.0E-04 3.2E-07 Mn-54 1.4E-10 3.0E-05 4.5E-06 Fe-55 1.0E-10 1.0E-04 1.0E-06 Fe-59 2.1E-11 1.0E-05 2.1E-06 Co-58 3.5E-10 2.0E-05 1.8E-05 Co-60 4.6E-11 3.0E-06 1.5E-05 Zn-65 4.3E-11 5.0E-06 8.6E-06 W-187 1.4E-11 3.0E-05 4.5E-07 Np-239 2.5E-11 2.0E-05 1.3E-06 Br-84 4.6E-12 4.0E-04 1.1E-08 Rb-88 2.9E-10 4.0E-04 7.1E-07 Sr-89 1.8E-10 8.0E-06 2.3E-05 Sr-91 9.1E-12 2.0E-05 4.5E-07 Y-91m 7.0E-12 2.0E-03 3.5E-09 Y-93 1.2E-11 2.0E-05 5.8E-07 Zr-95 4.3E-11 2.0E-05 2.2E-06 Nb-95 4.6E-11 3.0E-05 1.5E-06 Mo-99 5.4E-09 2.0E-05 2.7E-04 Tc-99m 4.9E-09 1.0E-03 4.9E-06 Ru-103 3.4E-10 3.0E-05 1.1E-05 Rh-103m 3.4E-10 6.0E-03 5.7E-08 Ru-106 1.6E-08 3.0E-06 5.5E-03 11.2-41 Revision 1
Discharge Concentration Effluent Concentration Fraction of Nuclide (Ci/ml)(1) Limit (Ci/ml)(2) Concentration Limit Ag-110m 1.4E-10 6.0E-06 2.3E-05 Te-129m 3.9E-10 7.0E-06 5.6E-05 Te-129 1.6E-11 4.0E-04 3.9E-08 Te-131m 7.4E-11 8.0E-06 9.3E-06 Te-131 4.0E-12 8.0E-05 5.0E-08 I-131 1.2E-08 1.0E-06 1.2E-02 Te-132 2.3E-09 9.0E-06 2.5E-04 I-132 3.6E-10 1.0E-04 3.6E-06 I-133 3.3E-09 7.0E-06 4.6E-04 I-134 8.5E-11 4.0E-04 2.1E-07 Cs-134 2.0E-07 9.0E-07 2.3E-01 I-135 9.1E-10 3.0E-05 3.0E-05 Cs-136 1.5E-07 6.0E-06 2.6E-02 Cs-137 1.5E-07 1.0E-06 1.5E-01 Ba-140 5.8E-10 8.0E-06 7.2E-05 La-140 7.8E-10 9.0E-06 8.6E-05 Ce-141 2.9E-11 3.0E-05 9.5E-07 Ce-143 2.0E-11 2.0E-05 9.9E-07 Pr-143 1.4E-11 2.5E-05 5.4E-07 Ce-144 3.3E-10 3.0E-06 1.1E-04 Pr-144 3.3E-10 6.0E-04 5.5E-07 H-3 1.1E-04 1.0E-03 1.1E-01 Total = 5.3E-01 s:
Annual average discharge concentration based on release of average daily discharge for 292 days per year with 6000 gpm dilution flow.
Effluent concentrations limits are from Reference 1.
11.2-42 Revision 1
Impoundment Model Parameters rameter Average Annual Condition poundment Model Fully Mixed ant Discharge Rate (cfs) 18.3 poundment Volume (cubic feet) 1,746,300 poundment Blowdown Rate (cfs) 2,538 11.2-43 Revision 1
LADTAP II Input Parameters(a) t Parameter Value shwater Site Selected harge Flowrate (cfs) 18.3 mile Population 3,455,395 (Tables 11.2-203 and 2.1-204, Year 2036)(b) rce Term Table 11.2-7 oundment Model Table 11.2-201 re Width Factor 0.2 tion Factors 1.0 nsit Time - Drinking Water (hr) 14.2 nsit Time - Fish and Recreational Uses (hr) 0 rt Fish Annual Harvest (lb/yr) 15,000 mercial Fish Annual Harvest (lb/yr) 0 reline Use (person-hrs/yr) 6,620,364(c) mming Exposure (person-hrs/yr) 6,620,364(c) ting Exposure (person-hrs/yr) 6,620,364(c) king Water Intake Union, SC Distance 21 miles Projected 2036 Population 24,725 Input parameters not specified use LADTAP II default values.
The population is conservatively projected to 2036; a date more than five years beyond the date of commercial operation of Unit 2.
Annual use is based on consideration of the recreational value of the Broad River, the population in the vicinity of the Broad River downstream of Lee Nuclear Station, and the average shoreline exposure time per person for each age group taken from Table E-4 of Regulatory Guide 1.109.
Fifteen percent of the 50-mile 2036 population is assumed to access Broad River recreational area annually.
11.2-44 Revision 1
Annual Dose to a Maximally Exposed Individual from Liquid Effluents (per unit)
Dose (mrem/yr) lt hway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LLI 3.13E-02 5.50E-02 4.06E-02 4.17E-03 1.88E-02 6.48E-03 4.38E-03 king 7.00E-04 2.04E-02 2.02E-02 2.79E-02 2.00E-02 1.96E-02 2.42E-02 reline 4.72E-05 4.03E-05 4.03E-05 4.03E-05 4.03E-05 4.03E-05 4.03E-05 4.03E-05 l 4.72E-05 3.20E-02 7.55E-02 6.09E-02 3.21E-02 3.88E-02 2.61E-02 2.86E-02 nager hway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LLI 3.29E-02 5.64E-02 2.32E-02 3.82E-03 1.90E-02 7.46E-03 3.30E-03 king 6.75E-04 1.46E-02 1.41E-02 2.09E-02 1.42E-02 1.38E-02 1.72E-02 reline 2.64E-04 2.25E-04 2.25E-04 2.25E-04 2.25E-04 2.25E-04 2.25E-04 2.25E-04 l 2.64E-04 3.38E-02 7.13E-02 3.75E-02 2.50E-02 3.34E-02 2.15E-02 2.07E-02 d
hway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LLI 4.08E-02 4.92E-02 9.19E-03 3.90E-03 1.60E-02 5.89E-03 1.45E-03 king 1.94E-03 2.82E-02 2.67E-02 4.37E-02 2.73E-02 2.65E-02 2.97E-02 reline 5.51E-05 4.71E-05 4.71E-05 4.71E-05 4.71E-05 4.71E-05 4.71E-05 4.71E-05 l 5.51E-05 4.28E-02 7.75E-02 3.60E-02 4.77E-02 4.34E-02 3.25E-02 3.12E-02 11.2-45 Revision 1
Dose (mrem/yr) nt hway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LLI 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 king 2.11E-03 2.82E-02 2.61E-02 5.32E-02 2.69E-02 2.61E-02 2.80E-02 reline 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l 0.00E+00 2.11E-03 2.82E-02 2.61E-02 5.32E-02 2.69E-02 2.61E-02 2.80E-02 11.2-46 Revision 1
Annual Population Dose from Liquid Effluents (per unit)
Dose (person-rem per yr) hway Skin Total Body Thyroid Kidney Lung GI-LLI Liver Bone
- 1.25E-02 9.58E-04 6.90E-03 2.45E-03 1.31E-03 2.05E-02 1.25E-02 king - 2.60E-01 3.69E-01 2.59E-01 2.54E-01 3.06E-01 2.66E-01 1.15E-02 reline 2.60E-02 2.23E-02 2.23E-02 - - - - -
mming - 5.39E-04 5.39E-04 - - - - -
ting - 2.69E-04 2.69E-04 - - - - -
l 2.60E-02 2.96E-01 3.93E-01 2.66E-01 2.56E-01 3.07E-01 2.87E-01 2.40E-02 11.2-47 Revision 1
Liquid Pathway Doses Compared to 40 CFR Part 190 Limits Dose (mrem/yr, per site)
Dose 40 CFR 190 Requirements Assessment of Both Units ole Body Dose Equivalent 25 1.22E-01(a) roid Dose 75 1.06E-01(b) e to Another Organ 25 1.55E-01(c) an adult receives the maximum individual whole body dose an infant receives the maximum thyroid dose a child receives the maximum individual organ dose which is to the liver 11.2-48 Revision 1
Liquid and Gaseous Pathway Doses Compared to 40 CFR Part 190 Limits Dose (mrem/yr, per site)(a)
Dose 40 CFR 190 Requirements Assessment of Both Units ole Body Dose Equivalent 25 3.74E+00(b) roid Dose 75 2.00E+01(c) e to Another Organ 25 9.05E+00(d)
Direct radiation from containment and other plant buildings is negligible based on information presented in the AP1000 DCD, Tier 2, Chapter 12, Subsection 12.4.2.1.
This value was conservatively calculated by summing the maximum whole body dose due to the liquid pathway (to an adult) and the maximum whole body dose due to the gaseous pathway (to a child).
An infant receives the maximum thyroid dose.
A child receives the maximum other individual organ dose which is to the bone.
11.2-49 Revision 1
Liquid Pathway Comparison of Maximum Individual Dose to 10 CFR Part 50, Appendix I Criteria Dose (mrem/yr, per unit)
Appendix I Dose Objective Unit 1 or 2 Assessment l Body Shoreline 4.03E-05 Drinking 2.02E-02 Fish 4.06E-02 Total 3 6.09E-02(a) imum Organ Shoreline 4.71E-05 Drinking 2.82E-02 Fish 4.92E-02 Total 10 7.75E-02(b)
An adult receives the maximum individual total body dose.
A child receives the maximum individual organ dose which is to the liver.
11.2-50 Revision 1
Liquid Pathway Comparison of Maximum Individual Dose to 10 CFR Part 20.1301 Criteria Dose (mrem/yr, per unit)
Dose 10 CFR 20.1301 Objective Unit 1 or 2 Assessment l Body - 6.09E-02(a) roid Dose - 5.32E-02(b)
E 100 6.25E-02(c) imum dose in any hour em/hr) 2 7.13E-06 An adult receives the maximum individual total body dose.
An infant receives the maximum thyroid dose.
Per the guidance of Regulatory Guide 1.183, the total effective dose equivalent (TEDE) is approximated by the sum of the total body dose and 3% of the thyroid dose.
11.2-51 Revision 1
WLS 1&2 - UFSAR Figure 11.2-1 Liquid Radwaste System Simplified Piping and Instrumentation Diagram (REF) WLS 11.2-52 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Figure 11.2-2 (Sheet 1 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 001 11.2-53 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Inside Auxiliary Building Figure 11.2-2 (Sheet 2 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 002 11.2-54 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Figure 11.2-2 (Sheet 3 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF WLS 003) 11.2-55 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Figure 11.2-2 (Sheet 4 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 004 11.2-56 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Figure 11.2-2 (Sheet 5 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 005 11.2-57 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement.
Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Inside Auxiliary Building Figure 11.2-2 (Sheet 6 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 006 11.2-58 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Inside Radwaste Building Figure 11.2-2 (Sheet 7 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 008 11.2-59 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement.
Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Inside Radwaste Building Figure 11.2-2 (Sheet 8 of 8)
Liquid Radwaste System Piping and Instrumentation Diagram (REF) WLS 009 11.2-60 Revision 1
ducts. A portion of these radionuclides is released to the reactor coolant because of a small ber of fuel cladding defects. Leakage of reactor coolant thus results in a release to the tainment atmosphere of the noble gases. Airborne releases can be limited both by restricting tor coolant leakage and by limiting the concentrations of radioactive noble gases and iodine in reactor coolant system.
ne is removed by ion exchange in the chemical and volume control system (CVS). Removal of noble gases from the reactor coolant system (RCS) is not normally necessary because the gases not build up to unacceptable levels when fuel defects are within normally anticipated ranges. If le gas removal is required because of high reactor coolant system concentration, the chemical volume control system can be operated in conjunction with the liquid radwaste system degasifier, emove the gases. See Subsection 9.3.6 for a description of these operations.
AP1000 gaseous radwaste system (WGS) is designed to perform the following major functions:
Collect gaseous wastes that are radioactive or hydrogen bearing Process and discharge the waste gas, keeping off-site releases of radioactivity within acceptable limits.
ddition to the gaseous radwaste system release pathway, release of radioactive material to the ironment occurs through the various building ventilation systems. These systems are described in tion 9.4 with a summary of system air flow rates and filter efficiencies provided in Table 9.4-1. The mated annual release reported in Subsection 11.3.3 includes contributions from the major ding ventilation pathways.
.1 Design Basis section 1.9.1 discusses the conformance of the gaseous radwaste system design with the criteria egulatory Guide 1.143.
.1.1 Safety Design Basis gaseous radwaste system serves no safety-related functions and therefore has no nuclear ty design basis.
.1.2 Power Generation Design Basis
.1.2.1 Capacity
.1.2.1.1 Gaseous Waste Collection gaseous radwaste system is designed to receive hydrogen bearing and radioactive gases erated during process operation. The radioactive gas flowing into the gaseous radwaste system rs as trace contamination in a stream of hydrogen and nitrogen.
design basis period of operation is the last 45 days of a fuel cycle. During this time, reactor lant system dilution and subsequent letdown from the chemical and volume control system into liquid radwaste system is at a maximum. Gaseous radwaste system inputs are as follows:
11.3-1 Revision 1
Letdown diversion for reactor coolant system degassing, assumed to remove gases from the reactor coolant system to a level of 1 cc/kg beginning with the reactor coolant system at the maximum hydrogen concentration of 40 cc/kg. At its maximum this input is 0.5 scfm hydrogen carrying a very small volume of radiogas yielding 245 scf total hydrogen.
Reactor coolant drain tank liquid transfer to maintain proper reactor coolant drain tank level, assuming 0.25 gallons per minute liquid input from the reactor coolant system, intermittently yielding 0.5 scfm hydrogen and nitrogen carrying a very small volume of radiogas, yielding about 80 scf hydrogen and nitrogen total.
Reactor coolant drain tank gas venting, conservatively estimated at 1 scf per day, yielding 45 scf total nitrogen and hydrogen.
.1.2.1.2 Waste Gas Processing gaseous radwaste system is designed to reduce the controlled activity releases in support of the rall AP1000 release goals.
en the various inputs to the gaseous radwaste system, with licensing basis assumptions for lysis and with normally operating gaseous radwaste system equipment available, the combined t releases must be within the limits outlined in 10 CFR 20 and 10 CFR 50 Appendix I ferences 1 and 2, respectively).
.1.2.2 Failure Tolerance
.1.2.2.1 System Leakage gaseous radwaste system operates at low pressures, slightly above atmospheric pressure, thus ing the potential for leakage. Manual valves are the type which eliminate the potential for stem age. The system is of welded construction to further limit leakage.
.1.2.2.2 Water Incursion umber of features prevent wetting the activated carbon delay beds. These features include trols and alarms in the liquid radwaste system to prevent high degasifier separator water level, gas cooler, moisture separator, drain traps, and automatic isolation of the guard bed inlet on high sture separator level in the gaseous radwaste system. Additional protection is provided by the vated carbon guard bed, which removes residual moisture as well as iodine from the gas stream.
oisture enters the first activated carbon delay bed, the operator bypasses that bed and either s it with a nitrogen purge or replaces the activated carbon.
.1.2.3 Anticipated Operational Occurrences
.1.2.3.1 Prevention of Hydrogen Ignition e the carrier gas for the radiogas inputs to the gaseous radwaste system includes hydrogen, the eous radwaste system is designed to prevent hydrogen ignition both within its own boundaries in connected systems (the liquid radwaste system and the nuclear island radioactive ventilation em).
11.3-2 Revision 1
lysis, using independent, redundant monitors, is provided within the gaseous radwaste system.
n high oxygen level in the system, an alarm alerts the operator. At an operator selectable oxygen centration of 4 percent or less, the liquid radwaste system vacuum pumps automatically stop to ate potentially oxygenated inputs to the gaseous radwaste system, and a valve automatically ns to initiate a nitrogen purge. The discharge isolation valve of the gaseous radwaste system is tinuously pressurized with nitrogen to prevent ingress of air into the system from the discharge gaseous radwaste system also eliminates sources of hydrogen ignition. The system rporates spark-proof valves, electrical grounding, and a nitrogen purge. Discharge to the heating, tilating and air-conditioning duct is downstream of the exhaust fans to provide additional ection against hydrogen ignition.
.1.2.4 Controlled Release of Radioactivity
.1.2.4.1 Expected Releases AP1000 design prevents the annual average concentration limits established by 10 CFR 20 pendix B, table 2, column 1) (Reference 1) for gaseous releases from being exceeded due to the ases resulting during plant operation. Subsection 11.3.3 describes the calculated releases of oactive materials from the gaseous radwaste system and other pathways during normal ration.
section 11.3.3 also contains an evaluation which demonstrates that the doses to individuals, r beyond the site boundary, resulting from the expected releases from the gaseous waste nagement systems are within numerical design objectives of Appendix I of 10 CFR 50 ference 2).
.1.2.4.2 Monitoring Releases eases from the gaseous radwaste system are continuously monitored by a radiation detector in discharge line. In addition, the system includes provisions for taking grab samples of the harge flow stream for analysis. In this manner, the requirements of General Design Criterion 64 met as described in Section 3.1. Section 11.5 discusses radiation monitoring.
.1.2.4.3 Operator Error or Equipment Malfunction revent the release of radioactive gases resulting from equipment failure or operator error, a ation monitor is located in the discharge line. This instrument provides an alarm signal at a high l setpoint to alert operators of rising radiation levels. The monitor is also interlocked with an ation valve in the discharge line; the valve closes at a higher level setpoint.
operator actions are required during gaseous radwaste system operation since, once aligned for ration, the system operates automatically in response to the control signals from the rumentation.
.1.3 Compliance with 10 CFR 20.1406 ccordance with the requirements of 10 CFR 20.1406 (Reference 4), the gaseous radwaste em is designed to minimize, to the extent practicable, contamination of the facility and the 11.3-3 Revision 1
.2 System Description
.2.1 General Description AP1000 gaseous radwaste system, as shown on Figure 11.3-1 is a once-through, ambient-perature, activated carbon delay system. The system includes a gas cooler, a moisture separator, ctivated carbon-filled guard bed, and two activated carbon-filled delay beds. Also included in the em are an oxygen analyzer subsystem and a gas sampling subsystem.
radioactive fission gases entering the system are carried by hydrogen and nitrogen gas. The ary influent source is the liquid radwaste system degasifier. The degasifier extracts both rogen and fission gases from the chemical and volume control system letdown flow which is rted to the liquid radwaste system or from the reactor coolant drain tank discharge.
ctor coolant degassing is not required during power operation with fuel defects at or below the ign basis level of 0.25 percent. However, the gaseous radwaste system periodically receives ent when chemical and volume control system letdown is processed through the liquid radwaste em degasifier during reactor coolant system dilution and volume control operations. Since the asifier is a vacuum type and requires no purge gas, the maximum gas influent rate to the gaseous waste system from the degasifier equals the rate that hydrogen enters the degasifier (dissolved in d).
other major source of input to the gaseous radwaste system is the reactor coolant drain tank.
rogen dissolved in the influent to the reactor coolant drain tank enters the gaseous radwaste em either via the tank vent or the liquid radwaste system degasifier discharge.
tank vent is normally closed, but is periodically opened on high pressure to vent the gas that has e out of solution. The reactor coolant drain tank liquid is normally discharged to the liquid waste system via the degasifier, where the remaining hydrogen is removed.
reactor coolant drain tank is purged with nitrogen gas to discharge nitrogen and fission gases to gaseous radwaste system before operations requiring tank access. The reactor coolant drain is also purged with nitrogen gas to dilute and discharge oxygen after tank servicing or inspection rations which allow air to enter the tank.
ents to the gaseous radwaste system first pass through the gas cooler where they are cooled to ut 40°F by the chilled water system. Moisture formed due to gas cooling is removed in the sture separator.
r leaving the moisture separator, the gas flows through a guard bed that protects the delay beds abnormal moisture carryover or chemical contaminants. The gas then flows through two delay s in series where the fission gases undergo dynamic adsorption by the activated carbon and are eby delayed relative to the hydrogen or nitrogen carrier gas flow. Radioactive decay of the fission es during the delay period significantly reduces the radioactivity of the gas flow leaving the em.
effluent from the delay bed passes through a radiation monitor and discharges to the ventilation aust duct. The radiation monitor is interlocked to close the gaseous radwaste system discharge ation valve on high radiation. The discharge isolation valve also closes on low ventilation system aust flow rate to prevent the accumulation of hydrogen in the aerated vent.
11.3-4 Revision 1
gaseous radwaste system is used intermittently. Most of the time during normal operation of the 000, the gaseous radwaste system is inactive. When there is no waste gas inflow to the system, discharge isolation valve closes, which maintains the gaseous radwaste system at a positive sure, preventing the ingress of air during the periods of low waste gas flow.
en the gaseous radwaste system is in use, its operation is passive, using the pressure provided he influent sources to drive the waste gas through the system.
largest input to the gaseous radwaste system is from the liquid radwaste system degasifier, ch processes the chemical and volume control system letdown flow when diverted to the liquid waste system and the liquid effluent from the liquid radwaste system reactor coolant drain tank.
chemical and volume control system letdown flow is diverted to the liquid radwaste system only ng dilutions, borations, and reactor coolant system degassing in anticipation of shutdown. The ign basis influent rate from the liquid radwaste system degasifier is the full diversion of the mical and volume control system letdown flow, when the reactor coolant system is operating with imum allowable hydrogen concentration. Since the liquid radwaste system degasifier is a uum type that operates without a purge gas, this input rate is very small, about 0.5 scfm.
liquid radwaste system degasifier is also used to degas liquid pumped out of the reactor coolant n tank. The amount of fluid pumped out, and therefore the gas sent to the gaseous radwaste em, is dependent upon the input into the reactor coolant drain tank. This is smaller than the input the chemical and volume control system letdown line.
final input to the gaseous radwaste system is from the reactor coolant drain tank vent. A nitrogen er gas is maintained in the reactor coolant drain tank. This input consists of nitrogen, hydrogen, radioactive gases. The tank operates at nearly constant level, with its vent line normally closed, his input is minimal. Venting is required only after enough gas has evolved from the input fluid to ease the reactor coolant drain tank pressure.
influent first passes through a gas cooler. Chilled water flows through the gas cooler at a fixed to cool the waste gas to about 40°F regardless of waste gas flow rate. Moisture formed due to cooling is removed in the moisture separator, and collected water is periodically discharged matically. To reduce the potential for waste gas bypass of the gas cooler in the event of valve age, a float-operated drain trap is provided which automatically closes on low water level.
gas leaving the moisture separator is monitored for temperature, and a high alarm alerts the rator to an abnormal condition requiring attention. Oxygen concentration is also monitored. On a oxygen alarm, a nitrogen purge is automatically injected into the influent line.
waste gas then flows through the guard bed, where iodine and chemical (oxidizing) taminants are removed. The guard bed also removes any remaining excessive moisture from the te gas.
waste gas then flows through the two delay beds where xenon and krypton are delayed by a amic adsorption process. The discharge line is equipped with a valve that automatically closes on er high radioactivity in the gaseous radwaste system discharge line or low ventilation exhaust t flow.
11.3-5 Revision 1
ed water or other causes) results in a gradual reduction in gaseous radwaste system ormance. Reduced performance is indicated by high temperature and discharge radiation ms. High-high radiation automatically terminates discharge.
.2.2.2 Purge Operations gaseous radwaste system is purged with nitrogen gas to expel residual oxygen gas after icing operations. The system is purged until the effluent from the outlet indicates a low oxygen centration. The gaseous radwaste system oxygen analyzer is temporarily aligned to monitor the in the discharge line. Nitrogen connections are also provided to the sample system and to the em discharge line for purge before and after maintenance operations.
.2.3 Component Description general descriptions and summaries of the design basis requirements for the gaseous radwaste em components follow. Table 11.3-2 lists the key design parameters for the gaseous radwaste em components.
seismic design classification and safety classification for the gaseous radwaste system ponents are listed in Section 3.2. The components listed are located in the Seismic Category I lear Island.
.2.3.1 Sample Pumps sample pumps are provided. One sample pump normally operates continuously to provide flow ugh the oxygen analyzers. The other sample pump is periodically used to provide flow from ous sample points through a sample cylinder. It is used as a backup to provide flow through the gen analyzers.
.2.3.2 Gas Cooler gas cooler heat exchanger is designed to cool the gas flow to near the temperature of the chilled er supply (40°F) for efficient moisture removal. The pressure of the gas flow through the gas ler is less than the chilled water pressure to minimize the potential for contaminating the chilled er system.
.2.3.3 Gaseous Radwaste System Tanks sture Separator moisture separator is sized for the design basis purge gas flow rate and is oversized for the er normal flow rate. The unit includes connections for high and low water level sensors.
ard Bed activated carbon guard bed protects the delay beds from abnormal moisture or chemical taminants. Under normal operating conditions, the guard bed provides increased delay time for on and krypton and removes iodine entering the system.
flow through the activated carbon bed is downward. A retention screen on the outlet of the guard prevents the loss of activated carbon from the unit. Activated carbon can be added to or uumed from the unit via a blind flange port.
11.3-6 Revision 1
ides adequate performance. This provides operational flexibility to permit continued operation of gaseous radwaste system in the event of operational upsets in the system that requires isolation ne bed.
waste gas flows vertically through columns of activated carbon. The activated carbon volume is n in Table 11.3-1.
etention screens are required on the delay beds since the flow enters and leaves each delay bed s top.
guard bed and the delay beds, including supports, in the gaseous radwaste system are designed eismic loads in conformance with Regulatory Guide 1.143. These are the only AP1000 ponents used to store or delay the release of gaseous radioactive waste. The beds are located in seismic Category I auxiliary building at elevation 666.
.2.3.4 Remotely Operated Valves sture Separator Level Control Valve normally closed, fail-closed globe valve is located in the liquid drain line from the moisture arator outlet line. It maintains the level in the moisture separator by regulating the flow from the sture separator to the liquid radwaste system. The valve receives a signal to automatically open high level in the moisture separator and to close on low level. The valve can also be manually trolled from the gaseous waste panel.
at-operated drain trap serves as a backup to this valve. This drain trap automatically closes on a water level in the moisture separator to stop drain flow to the liquid radwaste system in the event valve or instrument failure. This prevents waste gas bypass around the gas cooler due to level trol valve failure.
eous Radwaste System Discharge Isolation Valve normally closed, fail-closed globe valve is at the outlet of the system. The valve is interlocked to e on a high-high radiation signal in the gaseous radwaste system discharge line to prevent the ase of radioactivity in the event of a gaseous radwaste system failure. The valve also receives a al to automatically close in the event of a low ventilation system exhaust flow rate which prevents umulation of a flammable or explosive concentration of hydrogen in the aerated vent line.
ual control is provided on the gaseous radwaste panel.
ogen Purge Pressure Control Valve is a self-contained pressure regulating valve in the nitrogen purge line. It is set to maintain a ll positive pressure in the gaseous radwaste system to prevent ingress of air during periods of flow.
11.3-7 Revision 1
Venting of the containment which contains activity as a result of leakage of reactor coolant and as a result of activation of naturally occurring Ar-40 in the atmosphere to form radioactive Ar-41 Ventilation discharges from the auxiliary building which contains activity as a result of leakage from process streams Ventilation discharges from the turbine building Condenser air removal system (gaseous activity entering the secondary coolant as a result of primary to secondary leakage is released via this pathway)
Gaseous radwaste system discharges.
se releases are on-going throughout normal plant operations. There is no gaseous waste holdup ability in the gaseous waste management system and thus no criteria are required for rmining the timing of releases or the release rates to be used.
re are no gaseous effluent site interface parameters outside of the Westinghouse scope.
.3.1 Discharge Requirements release of radioactive gaseous and particulate effluents to the atmosphere may not exceed the centration limits specified in Reference 1 nor may the releases result in the annual offsite dose s specified in 10 CFR 50, Appendix I (Reference 2) being exceeded.
.3.2 Estimated Annual Releases annual average airborne releases of radionuclides from the plant are determined using the R-GALE code (Reference 3). The GALE code models releases using realistic source terms ved from data obtained from the experience of many operating pressurized water reactors. The e input parameters used in the analysis to model the AP1000 plant are provided in Table 11.2-6.
expected annual releases for a single unit site are presented in Table 11.3-3.
emonstrate compliance with the effluent concentration limits in Reference 1, the expected ases from Table 11.3-3 are used to determine the annual average concentration at the site ndary, and the results are compared with the Reference 1 concentration limits for unrestricted as in Table 11.3-4. As shown in Table 11.3-4, the overall fraction of the effluent concentration limit he expected releases is 0.030, which is significantly below the allowable value of 1.0.
.3.3 Release Points orne effluents are normally released through the plant vent or the turbine building vent. The plant t provides the release path for containment venting releases, auxiliary building ventilation ases, annex building releases, radwaste building releases, and gaseous radwaste system harge. The turbine building vents provide the release path for the condenser air removal system, d seal condenser exhaust and the turbine building ventilation releases.
11.3-8 Revision 1
ndary are 2.1 mrad for gamma radiation and 10.1 mrad for beta radiation. These doses are based he annual average atmospheric dispersion factor from Section 2.3 (2.0 x 10-5 seconds per cubic er). These doses are below the 10 CFR 50, Appendix I, design objectives of 10 mrad per year for ma radiation or 20 mrad per year for beta radiation.
radiological consequences due to a single failure of an active component in the gaseous waste system are evaluated assuming a 1-hour bypass of the delay beds and 30 minutes of ay before release to the environs. This analysis assumes a pre-existing condition of operation reactor coolant activity corresponding to 1 percent fuel defects as described in the Note for le 11.1-2. Using the site boundary (0 to 2 hr) atmospheric dispersion factor from Table 2.0-201, site boundary whole body dose is 0.1 rem.
calculated gaseous doses for the maximum exposed individual are compared to the regulatory ria in Appendix I of 10 CFR Part 50 and 10 CFR Part 20.1301 for acceptance. Table 11.3-205 Table 11.3-206 display this comparison and demonstrate that the calculated gaseous doses for maximally exposed individual are less than the regulatory criteria. The Lee Nuclear Station site-cific values are bounded by the DCD identified acceptable releases. With the annual airborne ases listed in Table 11.3-3, the site-specific air doses at ground level at the site boundary are mrad per year for gamma radiation and 7.32 mrad per year for beta radiation. These doses are ed on the annual average atmospheric dispersion factor from Section 2.3. These doses are below 10 CFR Part 50, Appendix I design objectives of 10 mrad per year for gamma radiation or mrad per year for beta radiation.
e and dose rate to man were calculated using the GASPAR II computer code. This code is based he methodology presented in Regulatory Guide 1.109. Factors common to both estimated vidual dose rates and estimated population dose are addressed in this subsection. Unique data discussed in the respective subsections.
vity pathways considered are plume, ground deposition, inhalation, and ingestion of vegetables, t, and milk (cow or goat).
ed on site meteorological conditions, the highest combined dose rate from plume exposure and und deposition occurs at the site boundary 0.27 mi. (427 m) NW of the Effluent Release ndary.
cultural products are estimated from U.S. Department of Agriculture National Agricultural istics Service. GASPAR II evenly distributes the food production over the entire 50 miles when n a total production for calculating dose.
ulation distribution within the 50-mi. radius is presented in Tables 2.1-203 and 2.1-204.
.3.4.1 Estimated Individual Doses e rates to individuals are calculated for airborne decay and deposition, inhalation, and ingestion ilk (goat or cow), meat and vegetables. Dose from plume and ground deposition are calculated ffecting all age groups equally.
me exposure approximately 0.27 mi. NW of the Effluent Release Boundary produced a maximum e rate to a single organ of 4.90 mrem/yr to skin. The maximum total body dose rate was ulated to be 7.32E-1 mrem/yr.
11.3-9 Revision 1
lation Dose at the site boundary, 0.27 mi. NW of the Effluent Release Boundary, results in a imum dose rate to a single organ of 1.54 mrem/yr to a childs thyroid. The maximum total body e rate is calculated to be 1.24E-1 mrem/yr to a teenager.
etable consumption assumes that the dose is received from the garden special location, roximately 1.0 mi. SSE of the plant. GASPAR II default vegetable consumption values are used eu of site-specific vegetable consumption data as permitted by Regulatory Guide 1.109. The mated maximum dose rate to a single organ is 2.42 mrem/yr to a childs thyroid. The maximum l body dose rate is calculated to be 4.59E-1 mrem/yr to a child.
t consumption assumes that the dose is received from the cow special location, approximately mi. SE of the plant. GASPAR II default meat consumption values are used in lieu of site-specific t consumption data as permitted by Regulatory Guide 1.109. The estimated maximum dose rate single organ is 2.74E-1 mrem/yr to a childs bone. The maximum total body dose rate is ulated to be 5.81E-2 mrem/yr to a child.
milk consumption assumes that the dose is received from the cow special location, roximately 1.65 mi. SE of the plant. GASPAR II default cow milk consumption values are used eu of site-specific cow milk consumption data as permitted by Regulatory Guide 1.109. The mated maximum dose rate to a single organ is 6.23 mrem/yr to an infants thyroid. The maximum l body dose rate is calculated to be 3.99E-1 mrem/yr to an infant.
t milk consumption assumes that the dose is received from the nearest milk goat special location, roximately 1.05 mi. SSW of the plant. GASPAR II default goat milk consumption values are used eu of site-specific goat milk consumption data as permitted by Regulatory Guide 1.109. The mated maximum dose rate to a single organ is 7.58 mrem/yr to an infants thyroid. The maximum l body dose rate is calculated to be 3.26E-1 mrem/yr to an infant.
maximum dose rate to any organ considering every pathway is calculated to be 9.95 mrem/yr to nfant's thyroid. The maximum total body dose rate is calculated to be 1.81 mrem/yr to a child.
se are below the 10 CFR 50, Appendix I design objectives of 5 mrem/yr to total body, and mrem/yr to any organ, including skin.
le 11.3-201 contains GASPAR II input data for dose rate calculations. Information regarding the cial locations for cow, goat, garden, site boundary and the EAB is located in Section 2.3.
le 11.3-203 contains total organ dose rates based on age group and pathway. Table 11.3-203 tains total air dose at each special location.
.3.4.2 Estimated Population Dose population dose analysis performed to determine off-site dose from gaseous effluents is based n the AP1000 generic site parameters included in Chapter 1 and Tables 11.3-1, 11.3-2 11.3-4, and the year 2056 population data in Tables 2.1-203 and 2.1-204. The population doses shown in Table 11.3-204.
.3.4.3 Gaseous Radwaste Cost-Benefit Analysis Methodology guidance for performing cost-benefit analysis for the gaseous radwaste system is similar to that d and described for the liquid radwaste system in Section 11.2. The gaseous radwaste treatment em augments annual costs were determined and the lowest annual cost considered a threshold 11.3-10 Revision 1
AP1000 sites with population dose estimates less than 6.32 person-rem total body or thyroid e from gaseous effluents, no further cost-benefit analysis is needed to demonstrate compliance 10 CFR 50, Appendix I, Section II.D.
.3.4.4 Gaseous Radwaste Cost Benefit Analysis population doses are given in Tables 11.3-204 and 11.3-208. The lowest cost gaseous radwaste em augment is $6,320. Assuming 100 percent efficiency of this augment, the minimum possible per person-rem is determined by dividing the cost of the augment by the population dose. This is 64 per person-rem total body ($6,320/5.00 person-rem). The total body exposure-related costs person-rem reduction exceed the $1,000 per person-rem criterion prescribed in Appendix I to CFR Part 50 and are therefore not cost-beneficial. Realistic efficiencies would increase the cost person-rem further above the $1,000 criterion.
hown in Tables 11.3-204 and 11.3-208, the Lee thyroid dose from gaseous effluents is person-rem, which exceeds the 6.32 person-rem threshold value. Based on the estimated person-rem/year thyroid dose, those augments with a Total Annual Cost (TAC) less than 00 are considered below.
R Air Ejector Charcoal/HEPA Filtration Unit TAC for this augment is $9,140. To be cost-beneficial at $1000 per person-rem, this augment t remove sufficient activity to decrease the population dose by at least 9.14 person-rem (thyroid);
is, decrease the thyroid dose from 9.80 person-rem (initial level) to a final level of person-rem. No iodine is released through the condenser air removal (offgas) system as shown able 11.3-3, sheet 2 of 3. This augment does not affect the iodine discharged by the plant which ounts for a total 4.85 person-rem in the thyroid population dose. Therefore, it would be impossible chieve the necessary dose reduction, and this augment is not cost-beneficial.
on Charcoal Adsorber TAC for this augment is $8,770. To be cost-beneficial at $1,000 per person-rem, this augment t remove sufficient activity to decrease the population dose by at least 8.77 person-rem (thyroid);
is, decrease the thyroid dose from 9.80 person-rem (initial level) to a final level of person-rem.
3-Ton Charcoal Adsorber unit in Regulatory Guide 1.110 is based on a 200 cubic foot charge of vated charcoal for an add-on vessel to an existing system per the information contained within documents Total Direct Cost Estimate Sheet attachments. For the AP1000, it is assumed that augment would be appended to the Gaseous Radwaste System where it would increase the y time of noble gases exiting the existing activated carbon delay beds. No iodine is released ugh the Gaseous Radwaste System as shown in Table 11.3-3, sheet 2 of 3. This augment does affect the iodine discharged from the plant which accounts for 4.85 person-rem in the thyroid ulation dose. Therefore, it would be impossible to achieve the necessary dose reduction, and this ment is not cost-beneficial.
n Condenser Vacuum Pump Charcoal/HEPA Filtration System TAC for this augment is $7,690. To be cost-beneficial at $1,000 per person-rem, this augment t remove sufficient activity to decrease the population dose by at least 7.69 person-rem (thyroid);
11.3-11 Revision 1
ch accounts for 4.85 person-rem in the thyroid population dose. Therefore, it would be impossible chieve the necessary dose reduction, and this augment is not cost-beneficial.
00 cfm Charcoal/HEPA Filtration System TAC for this augment is $7,580. To be cost-beneficial at $1,000 per person-rem, this augment t remove sufficient activity to decrease the population dose by at least 7.58 person-rem (thyroid);
is, decrease the thyroid dose from an initial level of 9.80 person rem to a final level of person-rem.
servatively assuming that this rather small capacity augment could be placed in the ventilation em at some point that would eliminate all iodine and particulate releases, it would not be effective ducing the noble gas releases, the carbon-14 release, or the airborne tritium release. The noble es, carbon-14, and tritium discharged by the plant account for 4.67 person-rem in the thyroid ulation dose. Therefore, it would be impossible to achieve the necessary dose reduction, and this ment is not cost-beneficial.
ft3 Gas Decay Tank TAC for this augment is $7,460. Thus, to be cost-beneficial at $1,000 per person-rem, this ment must remove at least 7.46 person-rem (thyroid); that is, decrease the thyroid dose from an al level of 9.80 person-rem to a final level of 2.34 person-rem.
odine is released through the AP1000 waste gas system as shown in Table 11.3-3. This augment ld not affect the iodine discharged by the plant which accounts for 4.85 person-rem in the thyroid ulation dose. Therefore, it would be impossible to achieve the necessary dose reduction, and this ment is not cost-beneficial.
am Generator Flash Tank Vent to Main Condenser TAC for this augment is $6,320. Thus, to be cost-beneficial at $1,000 per person-rem, this ment must remove at least 6.32 person-rem (thyroid); that is decrease the thyroid dose from an al level of 9.80 person-rem to a final level of 3.48 person-rem. Addition of this augment presumes the design already includes a steam generator flash tank; the augment being evaluated is the allation of vent piping and instrumentation from the tank to the main condenser. However, the 000 design does not include a steam generator flash tank. Therefore, the TAC of $6,320 for this ment is underestimated. As shown in Figure 10.4.8-1, the AP1000 design includes steam erator blowdown heat exchangers that provide cooling of the blowdown fluid and prevent flashing r to the blowdown flow entering the main condenser. Therefore, this augment would not provide additional dose reduction, and this augment is not cost-beneficial.
nclusion ed on the above evaluation, none of the radwaste augments are cost-beneficial in reducing the ual thyroid dose from gaseous effluents for Lee.
.3.5 Maximum Release Concentrations annual releases of radioactive gases and iodine provided in Table 11.3-3 represent expected ases from the plant and reflect an expected level of fuel cladding defects. If the plant operates the maximum defined fuel defect level, the releases would be substantially greater. The 11.3-12 Revision 1
ration with the maximum defined fuel defect level, and the resulting airborne radionuclide centrations at the site boundary are compared in Table 11.3-4 with the Reference 1 limits for centrations in unrestricted areas. As shown in Table 11.3-4, the overall fraction of the effluent centration limit for operation with the maximum defined fuel defect level is 0.33, which is well w the allowable value of 1.0.
.3.6 Quality Assurance quality assurance program for design, fabrication, procurement, and installation of the gaseous waste system is in accordance with the overall quality assurance program described in pter 17.
e the impact of radwaste systems on safety is limited, the extent of control required by endix B to 10 CFR Part 50 is similarly limited. Thus, a supplemental quality assurance program licable to design, construction, installation, and testing provisions of the gaseous radwaste em is established by procedures that comply with the guidance presented in Regulatory de 1.143.
.4 Inspection and Testing Requirements
.4.1 Preoperational Testing operational tests are performed to verify the proper operation of the WGS. The operational tests ude automatic closure of the discharge control/isolation valve, WGS-PL-V051, upon receipt of a ulated high radiation signal. The discharge line of the gaseous radwaste system includes a ation monitor, WGS-RE017, which detects a high radiation condition and generates an alarm that matically closes the discharge control/isolation valve. By imposing a simulated high radiation m signal, proper operation of the discharge control/isolation valve is confirmed by its closure.
.4.2 Preoperational Inspection proper performance of the gaseous radwaste system depends upon delay of gaseous onuclides by chemical adsorption on activated carbon. As the radionuclides are delayed, they ay and are no longer available for release to the environment. The rate of release and site ndary dose rates have been evaluated based upon the quantity of activated carbon in a delay being at least 80 cubic feet. An inspection of the gaseous radwaste system activated carbon y beds, WGS-MV01A and WGS-MV02B, will confirm that the contained volume of each delay is at least 80 cubic feet.
.5 Combined License Information
.5.1 Cost Benefit Analysis of Population Doses orming cost-benefit analysis for the gaseous radwaste system is addressed in section 11.3.3.4.3.
site specific cost-benefit analysis to demonstrate compliance with 10 CFR 50, Appendix I, arding population doses due to gaseous effluents is addressed in Subsections 11.3.3.4,
.3.4.1, 11.3.3.4.2, and 11.3.3.4.4.
11.3-13 Revision 1
-GW-GLR-008 (Reference 5).
.6 References "Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage," 10 CFR Part 20, Appendix B, Issued by 58 FR 67657, April 28, 1995.
"Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion >As-Low-As-Is-Reasonably-Achievable= for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," 10 CFR Part 50, Appendix I.
"Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," NUREG-0017, Revision 1, March 1985.
"Minimization of Contamination," 10 CFR 20.1406.
APP-GW-GLR-008, "Request for Closure of COL Items in DCD Chapter 11, Identification for Adsorbent Media," Westinghouse Electric Company LLC.
11.3-14 Revision 1
sign operating influent pressure (psig) 2 sign influent flow rate (scfm) 0.5 tivated carbon bed design operating temperature (°F) 77 tivated carbon bed design operating dew point (°F) 45 tivated carbon in delay beds (average) (pounds combined total) 4600 11.3-15 Revision 1
hanical Components mps ample Pumps umber 2 ype Diaphragm t Exchangers Cooler umber 1 ype Dual tube coil Process Side Cooling Side esign pressure (psig) 150 150 esign temperature (°F) 200 200 esign flow 1.0 scfm 0.15 gpm emperature inlet (°F) 125 40 emperature outlet (°F) 40.1 42 aterial Stainless steel Stainless steel ks rd Bed umber 1 ominal volume (ft3) 8 ype Vertical pipe esign pressure (psig) 100 esign temperature (°F) 150 aterial Stainless steel ay Bed umber 2 ominal volume (ft3) 80 ype Vertical serpentine esign pressure (psig) 100 esign temperature (°F) 150 aterial Carbon steel sture Separator umber 1 ominal volume (gal) 3 ype Vertical esign pressure (psig) 150 esign temperature (°F) 200 aterial Stainless steel 11.3-16 Revision 1
Indicate Instrumentation (Note 4) Alarm Cooler inlet temperature X ling water outlet temperature X inlet pressure X X - Hi bon Guard Bed inlet temperature X X - Hi bon Delay Beds inlet temperature X X - Hi outlet temperature X X - Hi annels outlet flow X outlet radiation (Note 3) X X - Hi outlet pressure X bon Bed Vault lt hydrogen (Note 2) X X - Hi lt temperature (Note 1) X X - Hi sture Separator er level X X - Hi pling Subsystem rogen concentration X X gen concentration X X - Hi annels flow X X - Lo s:
Vault temperature monitor common for guard bed and delay bed.
Vault hydrogen monitor common for guard bed and delay bed.
High outlet radiation alarm closes gas outlet isolation valve.
Monitoring of the gaseous radwaste system is performed through the data display and processing system. Control functions are performed by the plant control system. Appropriate alarms and displays are available in the control room. Local indication and control are available on portable displays which may be connected to the data display and processing system. See Chapter 7.
11.3-17 Revision 1
as Determined by the PWR-GALE Code, Revision 1 (Release Rates in Ci/yr)
Building/Area Ventilation Condenser Waste Gas Auxiliary Turbine Air Removal ble Gases(1) System Cont. Building Building System Total Kr-85m 0. 3.0E+01 4.0E+00 0. 2.0E+00 3.6E+01 Kr-85 1.65E+03 2.4E+03 2.9E+01 0. 1.4E+01 4.1E+03 Kr-87 0. 9.0E+00 4.0E+00 0. 2.0E+00 1.5E+01 Kr-88 0. 3.4E+01 8.0E+00 0. 4.0E+00 4.6E+01 Xe-131m 1.42E+02 1.6E+03 2.3E+01 0. 1.1E+01 1.8E+03 Xe-133m 0. 8.5E+01 2.0E+00 0. 0. 8.7E+01 Xe-133 3.0E+01 4.5E+03 7.6E+01 0. 3.6E+01 4.6E+03 Xe-135m 0. 2.0E+00 3.0E+00 0. 2.0E+00 7.0E+00 Xe-135 0. 3.0E+02 2.3E+01 0. 1.1E+01 3.3E+02 Xe-138 0. 1.0E+00. 3.0E+00 0. 2.0E+00 6.0E+00 Total 1.1E+04 itionally:
released via gaseous pathway 350 4 released via gaseous pathway 7.3 41 released via containment vent 34 11.3-18 Revision 1
(Release Rates in Ci/yr)
Building/Area Ventilation Fuel Condenser Handling Auxiliary Turbine Air Removal Iodines(1) Area(2) Cont. Building Building System Total I-131 4.5E-03 2.3E-03 1.1E-01 0. 0. 1.2E-01 I-133 1.6E-02 5.5E-03 3.8E-01 2.0E-04 0. 4.0E-01 11.3-19 Revision 1
(Release Rates in Ci/yr)
Building/Area Ventilation Fuel Waste Gas Auxiliary Handling dionuclide(1) System Cont. Building Area(2) Total Cr-51 1.4E-05 9.2E-05 3.2E-04 1.8E-04 6.1E-04 Mn-54 2.1E-06 5.3E-05 7.8E-05 3.0E-04 4.3E-04 Co-57 0. 8.2E-06 0. 0. 8.2E-06 Co-58 8.7E-06 2.5E-04 1.9E-03 2.1E-02 2.3E-02 C0-60 1.4E-05 2.6E-05 5.1E-04 8.2E-03 8.7E-03 Fe-59 1.8E-06 2.7E-05 5.0E-05 0. 7.9E-05 Sr-89 4.4E-05 1.3E-04 7.5E-04 2.1E-03 3.0E-03 Sr-90 1.7E-05 5.2E-05 2.9E-04 8.0E-04 1.2E-03 Zr-95 4.8E-06 0. 1.0E-03 3.6E-06 1.0E-03 Nb-95 3.7E-06 1.8E-05 3.0E-05 2.4E-03 2.5E-03 Ru-103 3.2E-06 1.6E-05 2.3E-05 3.8E-05 8.0E-05 Ru-106 2.7E-06 0. 6.0E-06 6.9E-05 7.8E-05 Sb-125 0. 0. 3.9E-06 5.7E-05 6.1E-05 Cs-134 3.3E-05 2.5E-05 5.4E-04 1.7E-03 2.3E-03 Cs-136 5.3E-06 3.2E-05 4.8E-05 0. 8.5E-05 Cs-137 7.7E-05 5.5E-05 7.2E-04 2.7E-03 3.6E-03 Ba-140 2.3E-05 0. 4.0E-04 0. 4.2E-04 Ce-141 2.2E-06 1.3E-05 2.6E-05 4.4E-07 4.2E-05 s:
The appearance of 0. in the table indicates less than 1.0 Ci/yr for noble gas or less than 0.0001 Ci/yr for iodine. For particulates, release is not observed and assumed less than 1 percent of the total particulate releases.
The fuel handling area is within the auxiliary building but is considered separately.
11.3-20 Revision 1
Effluent Expected Site Fraction of Maximum Fraction of Concentration Boundary(b) Concentration Site Boundary Concentration Limit Concentration Limit(b) Concentration Limit(c) dionuclide Ci/ml(a) Limit Ci/ml (expected) Limit Ci/ml(c) (maximum)
Kr-85m 1.0E-07 2.9E-11 2.9E-04 1.2E-10 1.2E-03 Kr-85 7.0E-07 3.3E-09 4.6E-03 6.9E-09 9.9E-03 Kr-87 2.0E-08 1.2E-11 5.9E-04 3.0E-11 1.5E-03 Kr-88 9.0E-09 3.6E-11 4.1E-03 1.5E-10 1.7E-02 Xe-131m 2.0E-06 1.4E-09 7.1E-04 1.7E-09 8.7E-04 Xe-133m 6.0E-07 6.9E-11 1.1E-04 1.3E-09 2.1E-03 Xe-133 5.0E-07 3.6E-09 7.3E-03 1.3E-07 2.5E-01 Xe-135m 4.0E-08 5.5E-12 1.4E-04 5.9E-12 1.5E-04 Xe-135 7.0E-08 2.6E-10 3.7E-03 8.5E-10 1.2E-02 Xe-138 2.0E-08 4.8E-12 2.4E-04 7.7E-12 3.8E-04 I-131 2.0E-10 9.5E-14 4.8E-04 2.0E-12 9.8E-03 I-133 1.0E-09 3.2E-13 3.2E-04 3.4E-12 3.4E-03 H-3 1.0E-07 2.8E-10 2.8E-03 2.8E-10 2.8E-03 C-14 3.0E-09 5.8E-12 1.9E-03 5.8E-12 1.9E-03 Ar-41 1.0E-08 2.7E-11 2.7E-03 2.7E-11 2.7E-03 Cr-51 3.0E-08 4.8E-16 1.6E-08 4.8E-16 1.6E-08 Mn-54 1.0E-09 3.4E-16 3.4E-07 3.4E-16 3.4E-07 Co-57 9.0E-10 6.5E-18 7.2E-09 6.5E-18 7.2E-09 Co-58 1.0E-09 1.8E-14 1.8E-05 1.8E-14 1.8E-05 Co-60 5.0E-11 6.9E-15 1.4E-04 6.9E-15 1.4E-04 Fe-59 5.0E-10 6.3E-17 1.3E-07 6.3E-17 1.3E-07 Sr-89 2.0E-10 2.4E-15 1.2E-05 9.9E-14 4.9E-04 Sr-90 6.0E-12 9.5E-16 1.6E-04 2.1E-14 3.5E-03 Zr-95 4.0E-10 7.9E-16 2.0E-06 1.7E-15 4.4E-06 Nb-95 2.0E-09 2.0E-15 9.9E-07 6.1E-15 3.0E-06 Ru-103 9.0E-10 6.3E-17 7.0E-08 6.3E-17 7.0E-08 11.3-21 Revision 1
Concentration Boundary(b) Concentration Site Boundary Concentration Limit Concentration Limit(b) Concentration Limit(c) dionuclide Ci/ml(a) Limit Ci/ml (expected) Limit Ci/ml(c) (maximum)
Ru-106 2.0E-11 6.2E-17 3.1E-06 9.9E-16 4.9E-05 Sb-125 7.0E-10 4.8E-17 6.9E-08 4.8E-16 6.9E-07 Cs-134 2.0E-10 1.8E-15 9.1E-06 9.5E-13 4.7E-03 Cs-136 9.0E-10 6.7E-17 7.5E-08 4.1E-13 4.6E-04 Cs-137 2.0E-10 2.9E-15 1.4E-05 8.1E-13 4.0E-03 Ba-140 2.0E-09 3.3E-16 1.7E-07 3.3E-16 1.7E-07 Ce-141 8.0E-10 3.3E-17 4.2E-08 1.9E-16 2.3E-07 Total = 3.0E-02 Total = 3.3E-01 s:
Effluent concentration limit is from Reference 1.
Expected site boundary concentration based on annual releases predicted by the PWR-GALE code (Table 11.3-3) and an annual average /Q of 2.0 x 10-5 seconds per cubic meter.
Maximum site boundary concentration based on adjusting the releases predicted by the PWR-GALE code (Table 11.3-3) to reflect operation with maximum defined fuel defect level and an annual average /Q of 2.0 x 10-5 seconds per cubic meter.
11.3-22 Revision 1
GASPAR II Input Parameters(a) ut Parameter Value mber of Source Terms 1 tance from site to NE Corner of the US (mi) 1088 urce Term Table 11.3-3 pulation Data Table 2.1-203 and Table 2.1-204, year 2056 ction of the year leafy vegetables are grown 0.58 ction of max individuals vegetable intake from own 0.76 den ction of the year milk cows are on pasture 0.75 ction of milk-cow feed intake from pasture while on 1 ture ction of the year goats are on pasture 0.83 ction of goat feed intake from pasture while on pasture 1 ction of the year beef cattle are on pasture 0.75 ction of beef-cattle feed intake from pasture while on 1 ture al Production Rate for the 50-mile area
-Vegetables (kg/yr) 151,333,289
-Milk (L/yr) 84,765,807
-Meat (kg/yr) 354,508,878 ecial Location Data Section 2.3 teorological Data Section 2.3 Input parameters not specified use default GASPAR II values.
11.3-23 Revision 1
Individual Dose Rates Dose (mrem/yr) hway Total Body GI-Tract Bone Liver Kidney Thyroid Lung Skin ult me 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 8.04E-01 4.90E+00 und 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.98E-01 etable 1.38E-01 1.39E-01 6.09E-01 1.38E-01 1.34E-01 9.08E-01 1.28E-01 1.27E-01 at 3.96E-02 4.36E-02 1.73E-01 3.96E-02 3.92E-02 6.59E-02 3.89E-02 3.88E-02 at Milk 5.72E-02 4.47E-02 1.60E-01 6.28E-02 5.38E-02 9.96E-01 4.49E-02 4.31E-02 w Milk 5.37E-02 4.95E-02 1.98E-01 5.62E-02 5.41E-02 8.13E-01 4.87E-02 4.81E-02 alation 1.23E-01 1.24E-01 1.86E-02 1.26E-01 1.27E-01 1.07E+00 1.59E-01 1.20E-01 al(a) 1.34E+00 1.34E+00 1.98E+00 1.35E+00 1.34E+00 4.02E+00 1.43E+00 5.53E+00 n
me 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 8.04E-01 4.90E+00 und 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.98E-01 etable 2.07E-01 2.09E-01 9.76E-01 2.12E-01 2.06E-01 1.23E+00 1.97E-01 1.96E-01 at 3.21E-02 3.44E-02 1.46E-01 3.23E-02 3.20E-02 5.13E-02 3.17E-02 3.16E-02 at Milk 8.56E-02 7.30E-02 2.91E-01 1.05E-01 8.96E-02 1.58E+00 7.45E-02 7.08E-02 11.3-24 Revision 1
Dose (mrem/yr) hway Total Body GI-Tract Bone Liver Kidney Thyroid Lung Skin w Milk 8.93E-02 8.47E-02 3.63E-01 9.71E-02 9.34E-02 1.29E+00 8.41E-02 8.28E-02 alation 1.24E-01 1.26E-01 2.25E-02 1.29E-01 1.31E-01 1.33E+00 1.80E-01 1.21E-01 al(a) 1.44E+00 1.44E+00 2.49E+00 1.46E+00 1.45E+00 5.18E+00 1.55E+00 5.63E+00 ld me 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 8.04E-01 4.90E+00 und 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.98E-01 etable 4.59E-01 4.52E-01 2.31E+00 4.69E-01 4.59E-01 2.42E+00 4.45E-01 4.43E-01 at 5.81E-02 5.91E-02 2.74E-01 5.85E-02 5.80E-02 8.73E-02 5.77E-02 5.76E-02 at Milk 1.71E-01 1.58E-01 7.07E-01 2.14E-01 1.87E-01 3.15E+00 1.62E-01 1.56E-01 w Milk 1.99E-01 1.93E-01 8.88E-01 2.16E-01 2.09E-01 2.60E+00 1.93E-01 1.91E-01 alation 1.10E-01 1.09E-01 2.73E-02 1.14E-01 1.17E-01 1.54E+00 1.56E-01 1.07E-01 al(a) 1.81E+00 1.80E+00 4.48E+00 1.84E+00 1.83E+00 8.18E+00 1.91E+00 6.00E+00 nt me 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 7.32E-01 8.04E-01 4.90E+00 und 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.53E-01 2.98E-01 etable N/A N/A N/A N/A N/A N/A N/A N/A 11.3-25 Revision 1
Dose (mrem/yr) hway Total Body GI-Tract Bone Liver Kidney Thyroid Lung Skin at N/A N/A N/A N/A N/A N/A N/A N/A at Milk 3.26E-01 3.09E-01 1.34E+00 4.23E-01 3.58E-01 7.58E+00 3.17E-01 3.07E-01 w Milk 3.99E-01 3.89E-01 1.72E+00 4.38E-01 4.17E-01 6.23E+00 3.91E-01 3.88E-01 alation 6.35E-02 6.21E-02 1.36E-02 6.82E-02 6.78E-02 1.38E+00 9.58E-02 6.13E-02 al(a) 1.45E+00 1.44E+00 2.72E+00 1.49E+00 1.47E+00 9.95E+00 1.54E+00 5.65E+00 The milk pathway contribution for the total dose of each receptor is conservatively assumed to be the higher of the two milk pathways, either goat milk or cow milk.
11.3-26 Revision 1
Dose in MilliRads at Special Locations ecial Location Beta Air Dose Gamma Air Dose w (Meat, Milk) 1.09E+00 1.99E-01 at (Milk) 8.25E-01 1.96E-01 B 3.25E+00 7.73E-01 e Boundary 7.32E+00 1.25E+00 rden 1.24E+00 2.94E-01 11.3-27 Revision 1
Population Doses (person-rem) athway Total Body GI-Tract Bone Liver Kidney Thyroid Lung Skin me 1.45E+00 1.45E+00 1.45E+00 1.45E+00 1.45E+00 1.45E+00 1.69E+00 1.48E+01 und 2.75E-01 2.75E-01 2.75E-01 2.75E-01 2.75E-01 2.75E-01 2.75E-01 3.23E-01 lation 4.09E-01 4.10E-01 4.60E-02 4.16E-01 4.21E-01 3.07E+00 4.97E-01 4.01E-01 etable 7.61E-01 7.60E-01 3.34E+00 7.63E-01 7.49E-01 7.75E-01 7.45E-01 7.43E-01 Milk 2.75E-01 2.68E-01 1.15E+00 2.85E-01 2.79E-01 1.82E+00 2.68E-01 2.66E-01 t 1.83E+00 1.90E+00 8.22E+00 1.83E+00 1.82E+00 2.41E+00 1.82E+00 1.81E+00 l 5.00E+00 5.07E+00 1.45E+01 5.02E+00 5.00E+00 9.80E+00 5.29E+00 1.84E+01 11.3-28 Revision 1
Calculated Maximum Individual Doses Compared to 10 CFR Part 50 Appendix I Design Objectives cription Design Objective Calculated Values le Gases(a)
Gamma Dose (mrad) 10 1.25E+00 Beta Dose (mrad) 20 7.32E+00 Total Body Dose (mrem) 5 7.32E-01 Skin Dose (mrem) 15 4.90E+00 ioiodines and Particulates Total Body Dose (mrem) - 1.08E+00 Max to Any Organ (mrem)(b) 15 9.21E+00 imum Doses to Any Organ uding Noble Gas Total Body Dose em)(c) 15 9.95E+00 Doses due to noble gases in the released plume are calculated at the location of maximum dose at the site boundary (location of highest /Q values). This location is 0.27 miles (427 m) northwest of the Effluent Release Boundary.
The maximum dose to any organ is the dose to the thyroid of an infant.
The maximum organ dose listed here includes the dose due to ground exposure, inhalation, food pathways, and the total plume (noble gas) dose given above.
11.3-29 Revision 1
Maximum Individual Doses from Both Units Due to Routine Gaseous Effluents Compared to 10 CFR 20.1301 Limits cription Limit Calculated Values E (mrem)(a) 100 4.11E+00 imum Dose per Hour em/hr) 2 4.70E-04 Consistent with Regulatory Guide 1.183, the TEDE reported here is 3% of the thyroid dose plus the total body dose from Table 11.3-202. The maximum TEDE is to a child.
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Collective Gaseous Doses Compared to 40 CFR Part 190 Limits Calculated Values for Both cription Limit Units l Body Dose(a) Equivalent em) 25 3.62E+00 roid Dose (mrem) 75 1.99E+01 to Any Other Organ em)(b) 25 8.97E+00 The total body dose resulting from plume (noble gas) and radioiodine and particulate exposure pathways due to radiological releases from both units.
Note that the maximum dose to any organ other than the thyroid is the dose to the bone of a child.
The max dose to any other organ listed here includes the dose due to ground exposure, inhalation, food pathways, and the total body plume (noble gas) as given in Table 11.3-202.
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Population Dose by Isotopic Group rce Total Body % of Total Thyroid % of Total (person-rem) Total Body (person-rem) Thyroid le Gases 1.45E+00 29% 1.45E+00 15%
nes 1.00E-02 0% 4.85E+00 49%
iculates 3.16E-01 6% 2.74E-01 3%
4 2.45E+00 49% 2.45E+00 25%
7.70E-01 15% 7.70E-01 8%
l 5.00E+00 100% 9.80E+00 100%
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re represents system functional arrangement.
ils internal to the system may differ as a result of ementation factors such as vendor-specific ponent requirements.
Figure 11.3-1 Gaseous Radwaste System Simplified Sketch 11.3-33 Revision 1
WLS 1&2 - UFSAR Figure represents system functional arrangement. Details internal to the system may differ as a result of implementation factors such as vendor-specific component requirements.
Figure 11.3-2 Gaseous Radwaste System Piping and Instrumentation Diagram (REF) WGS 001 11.3-34 Revision 1
hange resins and deep bed filtration media, spent filter cartridges, dry active wastes, and mixed tes generated as a result of normal plant operation, including anticipated operational urrences. The system is located in the auxiliary and radwaste buildings. Processing and kaging of wastes are by mobile systems in the auxiliary building rail car bay and in the mobile ems facility part of the radwaste building. The packaged waste is stored in the auxiliary and waste buildings until it is shipped offsite to a licensed disposal facility.
use of mobile systems for the processing functions permits the use of the latest technology and ids the equipment obsolescence problems experienced with installed radwaste processing ipment. The most appropriate and efficient systems may be used as they become available.
system does not handle large, radioactive waste materials such as core components or oactive process wastes from the plant's secondary cycle. However, the volumes and activities of secondary cycle wastes are provided in this section.
.1 Design Basis
.1.1 Safety Design Basis solid waste management system performs no function related to the safe shutdown of the plant.
system's failure does not adversely affect any safety-related system or component; therefore, system has no nuclear safety design basis.
re are no safety related systems located near heavy lifts associated with the solid waste agement system. Therefore, a heavy loads analysis is not required.
.1.2 Power Generation Design Basis solid waste management system provides temporary onsite storage for wastes prior to essing and for the packaged wastes. The system has a 60-year design objective and is designed maximum reliability, minimum maintenance, and minimum radiation exposure to operating and ntenance personnel. The system has sufficient temporary waste accumulation capacity based on imum waste generation rates so that maintenance, repair, or replacement of the solid waste agement system equipment does not impact power generation.
.1.3 Functional Design Basis solid waste management system is designed to meet the following objectives:
Provide for the transfer and retention of spent radioactive ion exchange resins and deep bed filtration media from the various ion exchangers and filters in the liquid waste processing, chemical and volume control, and spent fuel cooling systems Provide the means to mix, sample, and transfer spent resins and filtration media to high integrity containers or liners for dewatering or solidification as required Provide the means to change out, transport, sample, and accumulate filter cartridges from liquid systems in a manner that minimizes radiation exposure of personnel and spread of contamination 11.4-1 Revision 1
Provide the means to segregate solid wastes (trash) by radioactivity level and to temporarily store the wastes Provide the means to accumulate radioactive hazardous (mixed) wastes Provide the means to segregate clean wastes originating in the radiologically controlled area (RCA)
Provide the means to store packaged wastes for at least 6 months in the event of delay or disruption of offsite shipping Provide the space and support services required for mobile processing systems that will reduce the volume of and package radioactive solid wastes for offsite shipment and disposal according to applicable regulations, including Department of Transportation regulation 49 CFR 173 (Reference 1) and NRC regulation 10 CFR 71 (Reference 2)
Provide the means to return liquid radwaste to the liquid radwaste system (WLS) for subsequent processing and monitored discharge solid waste management system is designed according to NRC Regulatory Guide 1.143 to meet requirements of General Design Criterion (GDC) 60 as discussed in Sections 1.9 and 3.1. The mic design classifications of the radwaste building and system components are provided in tion 3.2.
visions are made in the auxiliary and radwaste buildings to use mobile radwaste processing ems for processing and packaging each waste stream including concentration and solidification hemical wastes from the liquid waste management system, spent resin dewatering, spent filter ridge encapsulation and dry active waste sorting and compaction.
radioactivities of influents to the solid waste management system are based on estimated onuclide concentrations and volumes. These estimates are based on operating plant experience, sted for the size and design differences of AP1000. The influent source terms are consistent with tion 11.1.
solid waste management system airborne process effluents are released through the monitored t vent as described as part of the 10 CFR 50 (Reference 3), Appendix I, analysis presented in section 11.3.3.
solid waste management system collects and stores radioactive wastes within shielding to ntain radiation exposure to plant operation and maintenance personnel as low as is reasonably ievable (ALARA) according to General Design Criteria 60 as discussed in Section 3.1 and ulatory Guide 8.8. Personnel exposures will be maintained well below the limits of 10 CFR 20 ference 4). Design features incorporated to maintain exposures ALARA include remote and semi-ote operations, automatic resin transport line flushing, and shielding of components, piping and tainers holding radioactive materials. Access to the solid waste storage areas is controlled, to imize inadvertent personnel exposure, by suitable barriers such as heavy storage cask covers locked or key-card-operated doors or gates (see Section 12.1).
solid waste management system conforms to the design criteria of NRC Branch Technical ition ETSB 11-3. Suitable fire protection systems are provided as described in Subsection 9.5.1.
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sportation in 49 CFR 173 and for radioactive waste disposal in 10 CFR 61 (Reference 5) as well pecific disposal facility requirements.
.1.4 Compliance with 10 CFR 20.1406 ccordance with the requirements of 10 CFR 20.1406 (Reference 11), the solid radwaste system esigned to minimize, to the extent practicable, contamination of the facility and the environment, itate decommissioning, and minimize, to the extent practicable, the generation of radioactive te. This is done through appropriate selection of design technology for the system, plus rporating the ability to update the system to use the best available technology throughout the life e plant.
.2 System Description
.2.1 General Description solid waste management system includes the spent resin system. The flows of wastes through solid waste management system are shown on Figure 11.4-1. The radioactivity of influents to the em are dependent on reactor coolant activities and the decontamination factors of the processes e chemical and volume control system, spent fuel cooling system, and the liquid waste essing system.
parameters used to calculate the estimated activity of the influents to the solid waste agement system are listed in Table 11.4-1. The estimated expected isotopic curie content of the ary spent resin and filter cartridge wastes to be processed on an annual basis is listed on le 11.4-2. Table 11.4-3 provides the same information for the estimated maximum annual vities. The AP1000 has sufficient radwaste storage capacity to accommodate the maximum eration rate.
radioactivity of the dry active waste is expected to normally range from 0.1 curies per year to ries per year with a maximum of about 16 curies per year. This waste includes spent HVAC rs, compressible trash, non-compressible components, mixed wastes and solidified chemical tes. These activities are produced by relatively long lived radionuclides (such as Cr-51, Fe-55, 58, Co-60, Nb-95, Cs-134 and Cs-137), and therefore, radioactivity decay during processing and age is minimal. These activities thus apply to the waste as generated and to the waste as ped.
estimated expected and maximum annual quantities of waste influents by source and form are d in Table 11.4-1 with disposal volumes. The annual radwaste influent rates are derived by tiplying the average influent rate (e.g. volume per month, volume per refueling cycle) by one year me. The annual disposal rate is determined by applying the radwaste packaging efficiency to the ual influent rate. The influent volumes are conservatively based on an 18-month refueling cycle.
ual quantities based on a 24-month refueling cycle are less than those for an 18-month cycle.
estimated expected isotopic curie content of the primary spent resin and filter cartridge wastes to hipped offsite are presented in Table 11.4-4 based on 90 days of decay before shipment. The e information is presented in Table 11.4-5 for the estimated maximum activities based on ays of decay before shipment.
tion 11.1 provides the bases for determination of liquid source terms used to calculate several of solid waste management system influent source terms. The influent data presented in 11.4-3 Revision 1
adwaste which is packaged and stored by AP1000 will be shipped for disposal. The AP1000 has rovisions for permanent storage of radwaste. Radwaste is stored ready for shipment. Shipped mes of radwaste for disposal are estimated in Table 11.4-1 from the estimated expected or imum influent volumes by making adjustments for volume reduction processing by mobile ems and the expected container filling efficiencies. For drum compaction, the overall volume uction factor, including packaging efficiency, is 3.6. For box compaction, the overall volume uction factor is 5.4. These adjustments result in a packaged internal waste volume for each waste rce, and the number of containers required to hold this volume is based on the container's internal me. The disposal volume is based on the number of containers and the external (disposal) me of the containers.
expected disposal volumes of wet and dry wastes are approximately 547 and 1417 cubic feet year, respectively as shown in Table 11.4-1. The wet wastes shipping volumes include 510 cubic per year of spent ion exchange resins and deep bed filter activated carbon, 20 cubic feet of me reduced liquid chemical wastes and 17 cubic feet of mixed liquid wastes. The spent resins activated carbon are initially stored in the spent resin storage tanks located in the rail car bay of auxiliary building. When a sufficient quantity has accumulated, the resin is sluiced into two cubic feet high-integrity containers in anticipation of transport for offsite disposal. Liquid chemical tes are reduced in volume and packaged into three 55-gallon drums per year (about 20 cubic
) and are stored in the packaged waste storage room of the radwaste building. The mixed liquid tes fill less than three drums per year (about 17 cubic feet per year) and are stored on tainment pallets in the waste accumulation room of the radwaste building until shipped offsite for essing.
two spent resin storage tanks (275 cubic feet usable, each) and one high integrity container in spent resin waste container fill station at the west end of the rail car bay of the auxiliary building ide more than a year of spent resin storage at the expected rate, and several months of storage e maximum generation rate. The expected radwaste generation rate is based upon the following:
All ion exchange resin beds are disposed and replaced every refueling cycle.
The WGS activated carbon guard bed is replaced every refueling cycle.
The WGS delay beds are replaced every ten years.
All wet filters are replaced every refueling cycle.
Rates of compactible and non-compactible radwaste, chemical waste, and mixed wastes are estimated using historical operating plant data.
maximum radwaste generation rate is based upon the following:
The ion exchange resin beds are disposed based upon operation with 0.25% fuel defects.
The WGS activated carbon guard bed is replaced twice every refueling cycle.
The WGS delay beds are replaced every five years.
All wet filters are replaced based upon operation with 0.25% fuel defects.
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Primary to secondary system leakage contaminates the condensate polishing system and blowdown system resins and membranes which are replaced.
dry solid radwaste includes 1383 cubic feet per year of compactible and non-compactible waste ked into about 14 boxes (90 cubic feet each) and ten drums per year. Drums are used for higher vity compactible and non-compactible wastes. Compactible waste includes HVAC exhaust filter, und sheets, boot covers, hair nets, etc. Non-compactible waste includes about 60 cubic feet per r of dry activated carbon and other solids such as broken tools and wood. Solid mixed wastes will upy 7.5 cubic feet per year (one drum). The low activity spent filter cartridges may be compacted ll about 0.40 drums per year (3 ft3/year) and are stored in the packaged waste storage room.
paction is performed by mobile equipment or is performed offsite. High activity filter cartridges fill e drums per year (22.5 cubic feet per year) and are stored in portable processing or storage ks in the rail car of the auxiliary building.
total volume of radwaste to be stored in the radwaste building packaged waste storage room is 7 cubic feet per year at the expected rate and 2544 cubic feet per year at the maximum rate. The pactible and non-compactible dry wastes, packaged in drums or steel boxes, are stored with the ed liquid and mixed solid, volume reduced liquid chemical wastes, and the lower activity filter ridges. The quantities of liquid radwaste stored in the packaged waste storage room of the waste building consist of 20 cubic feet of chemical waste and 17 cubic feet of mixed liquid waste.
useful storage volume in the packaged waste storage room is approximately 3900 cubic feet feet deep, 30 feet long, and 13 feet high), which accommodates more than one full offsite waste ment using a tractor-trailer truck. The packaged waste storage room provides storage for more two years at the expected rate of generation and more than a year at the maximum rate of eration. One four-drum containment pallet provides more than 8 months of storage capacity for liquid mixed wastes and the volume reduced liquid chemical wastes at the expected rate of eration and more than 4 months at the maximum rate.
nservative estimate of solid wet waste includes blowdown material based on continuous ration of the steam generator blowdown purification system, with leakage from the primary to ondary system. The volume of radioactively contaminated material from this source is estimated e 540 cubic feet per year. Provisions for processing and disposal of radioactive steam generator down resins and membranes are described in Subsection 10.4.8. Note that, although included e for conservatism, this volume of contaminated resin will be removed from the plant within the taminated electrodeionization unit and not stored as wet waste.
condensate polishing system includes mixed bed ion exchanger vessels for purification of the densate as described in Subsection 10.4.6. Should the resins become radioactive, the resins are sferred from the condensate polishing vessel directly to a temporary processing unit or to the porary processing unit via the spent resin tank. The processing unit, located outside of the turbine ding, dewaters and processes the resins as required for offsite disposal. Radioactive condensate shing resin will have very low activity. It will be disposed in containers as permitted by DOT ulations. After packaging, the resins may be stored in the radwaste building. Based on a typical densate polishing system operation of 30 days per refueling cycle with leakage from the primary em to the secondary system, the volume of radioactively contaminated resin is estimated to be cubic feet per year (one 309 cubic foot bed per refueling cycle). Normal disposal of radioactive condensate polishing system resins is described in Subsection 10.4.6.
parameters used to calculate the activities of the steam generator blowdown solid waste and densate polishing resins are given in Table 11.4-1. Based on the above volumes, the disposal me is estimated to be 939 cubic feet per year. The expected and maximum activities of the resins 11.4-5 Revision 1
.2.2 Component Description seismic design classification and safety classification for the solid waste management system ponents are listed in Section 3.2. The components listed are located in the seismic Category I lear Island. Table 11.4-10 lists the solid waste management system equipment design ameters. The following subsections provide a functional description of the major system ponents.
.2.2.1 Spent Resin Tanks spent resin tanks provide holdup capacity for spent resin and filter bed media decay before essing. High- and low-activity resins may be mixed to limit the radioactivity concentration in the te containers to 10 Ci/ft3 in accordance with the USNRC Technical Position on Waste Form ference 6).
in mixing capability is provided by mixing eductors in each tank, and resin dewatering, air rging and complete draining capabilities are also provided. The ultrasonic level sensors and atering screens are arranged for remote removal. The vent and overflow connections have ens to prevent the inadvertent discharge of spent resin, and they are routed to the radioactive te drain system (WRS).
.2.2.2 Resin Mixing Pump resin mixing pump provides the motive force to fluidize and mix the resins in the spent resin s, to transfer water between spent resin tanks, to discharge excess water from the spent resin s to the liquid waste processing system, and to flush the resin transfer lines.
.2.2.3 Resin Fines Filter resin fines filter minimizes the spread of high-activity resin fines and dislodged crud particles by ring the water used for line flushing or discharged from the spent resin tanks to the liquid waste essing system.
.2.2.4 Resin Transfer Pump resin transfer pump provides the motive force for recirculation of spent resins via either one of spent resin tanks for mixing and sampling, for transferring spent resin between tanks, and for ding high- and low-activity resins to meet the specific activity limit for disposal. The resin transfer p is also used to transfer spent resins to a waste container in the fill station or in its shipping cask ted in the auxiliary building rail car bay.
.2.2.5 Resin Sampling Device resin sampling device collects a representative sample of the spent resin either during spent n recirculation or during spent resin waste container filling operations. A portable shielded cask is ided for sample jar transfer.
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a in the auxiliary building, transfer of the filter cartridges into and out of the filter storage, and ing of the filter cartridges into disposal containers.
.2.3 System Operation
.2.3.1 Spent Resin Handling Operations ineralized water is used to transfer spent resins from the various ion exchangers to the spent n tanks. A demineralized water transfer pump provides the pressurized water flow to transfer the nt resins as described in Subsection 9.2.4. Before the transfer operation, it is verified that the cted spent resin tank is aligned as a receiver and has the capacity to accept the bed. It is also fied that the resin mixing pump is aligned to discharge excess transfer water through the resin s filter to the liquid waste processing system.
ing the transfer operation the tank level is monitored and the resin mixing pump is operated, if uired, to limit tank water level. The operator stops the transfer when the CCTV camera viewing the t flow glass indicates on a control panel monitor that the sluice water is clear and the transfer line herefore, flushed of resins.
r the bed transfer, the tank solids level can be checked by operating the resin mixing pump to er the water level below the solids level. The solids level can be determined by the ultrasonic ace detector.
ween bed transfer operations the water level in the spent resin tanks is maintained above the ds level. Demineralized water is supplied for water level adjustment as well as a backup water rce for flushing resin handling lines after resin recirculation and waste disposal container filling rations.
solids bed can be agitated and mixed at any time by using compressed air or by operating the n mixing pump in the resin mixing mode. In the resin mixing mode, water is drawn from the spent n tank via resin retention screens. The water is returned via tank mixing eductors that generate a n slurry recirculation within the tank equivalent to about four times the flow rate generated by the n mixing pump. The solids bed is locally fluidized during this operation.
resin mixing mode is established to fluidize and mix the solids bed in the spent resin tank before te disposal container filling. The resin transfer pump is then started in the recirculation mode. A n slurry is drawn from the spent resin tank and returned to the same tank. A representative resin ple may be obtained during recirculation or container filling modes by operating the sampling ice.
portable system's container fill valve is opened to initiate the filling operation. The resin atering pump of the portable dewatering system is started to dewater the resin as it accumulates e container. The resin dewatering pump discharges the water to the recirculation line. The water s back to the spent resin tank, thereby preserving the water inventory in the system and retaining resin fines or dislodged crud within the system.
resin mixing pump can be stopped at any time during the filling operation. When the solids level rs the top of the container, as detected by level sensors and observed by a television camera, the alve is closed and cycled to top off the container. Excessive water or solids level automatically es the fill valve.
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te container. The resin mixing pump supplies filtered flush water from the spent resin tank. The able dewatering system's dewatering pump is operated periodically until no further dewatering is detected by the pump discharge pressure indicator and/or audible indications from the pump.
.2.3.2 Spent Filter Processing Operations ter transfer cask is used to change the higher-activity filters of the chemical and volume control em and spent fuel cooling system. The filter vessel is drained, and the filter cover is opened otely. The shield plug of the port over the filter is removed and the transfer cask, without its om shield cover, is lifted and positioned on the port directly over the cartridge in the filter vessel.
apple inside the transfer cask is remotely lowered and connected to the filter cartridge. The ridge is lifted into the transfer cask, and the cask is transferred over plastic sheeting to the bottom ld cover. The dose rate of the cartridge is measured with a long probe, and the cask is lowered and connected to the bottom shield cover. The transfer cask is then moved to the auxiliary ding rail car bay.
cent applicable sample analysis results are available, the filter cartridge can be loaded directly a disposal container as described in the following paragraph. If analysis is required, a sample of filter media is obtained through a port in the transfer cask. The filter cartridge is placed in one of high-activity filter storage tubes until sample analysis results are available. The transfer cask om cover is disconnected, the transfer cask is lifted by the crane and transferred to a position r one of the temporary storage tubes, and the spent filter cartridge is lowered into the tube. After ing the transfer cask away, the crane is used to install a shield plug onto the storage tube. Any er draining from the filter during storage collects in the storage tube which may be drained to a r drain for subsequent transfer to the liquid radwaste system.
en sample analysis is complete and packaging requirements are established, the transfer cask is d to retrieve the spent cartridges from storage and deposit them into a waste container via a port e top of a portable processing and storage cask. Plastic coverings are removed and the tainer is capped, smear-surveyed, and decontaminated as required, using reach rod tools ugh a cask port. The dose rate survey is also made through a cask port. Transfer of the filled te container to the shipping cask, including cask cover handling, is then performed using the rail bay crane under remote control.
rs with dose rates less than 15 R/hr on contact may be changed from outside of filter vessel lding by using reach rod tools. The filter vessel is drained, and the cover is removed. Then the nt filter cartridge is grappled and lifted out and into a filter transfer cask.
he radwaste building, low and moderate activity filter cartridges are deposited into disposal or age drums. The drums are stored within portable shield casks in the shielded accumulation room, ch is serviced by the mobile systems facility crane. Depending on dose rates and analysis results, ilization may or may not be required. Cartridges not requiring stabilization are loaded into dard, 55 gallon shipping drums with absorbent and may be compacted using a mobile system.
en stabilization is required, the cartridges may be loaded into either high integrity containers or dard drums. If standard drums are used, mobile equipment is used to encapsulate the contents e drums.
drum covers are manually installed, and the drums are smear surveyed, decontaminated by ng, if required, weighed, stacked on pallets, and placed in the packaged waste storage room.
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hielded shipment, the drums are loaded onto a cask pallet and into a shielded shipping cask g the mobile systems facility crane.
ioactive filters from ventilation exhaust filtration units are bagged and transported to the radwaste ding, where they are temporarily stored. The filters are compacted along with other dry active tes by a mobile system as described in the following subsection.
.2.3.3 Dry Waste Processing Operations wastes are segregated by measuring the contact dose rate of the wastes to determine the ropriate processing method. The contact dose rates for initial waste segregation are as follows:
Low activity <5 mR/hr Moderate activity 5 mR/hr to 100 mR/hr High activity >100 mR/hr se activity levels may be adjusted by the operator to minimize exposures while maximizing essing efficiency.
stes from surface contamination areas in the radiologically controlled area are placed in bags or tainers and tagged at the point of origin with information on radiation levels, waste type, and tination. The bags or containers are transported to the radwaste building, where they are placed low-, moderate-, or high-activity storage, segregated by portable shielding as appropriate.
high-activity wastes (greater than 100 mR/hr) are normally expected to be compacted in drums g a mobile compactor system in the same manner as lower-activity filter cartridges.
erate-activity wastes (5 mR/hr to 100 mR/hr) are expected to be sorted in a mobile system to ove reusable items such as protective clothing articles and tools, hazardous wastes, and larger compressible items. The remaining wastes are normally compacted by mobile equipment. The kaged wastes may be loaded directly onto a truck for shipment or may be stored in the packaged te storage room until a truck load quantity accumulates.
-activity, dry active waste (less than 5 mR/hr) generally contains a large amount of radioactive material. It is expected that these wastes normally will be processed through a mobile ation monitoring and sorting system to remove non-radioactive items for reuse or local disposal.
diation survey allows identification and removal of potentially clean items for the clean waste fication. The remaining radioactive wastes are normally compacted or packaged for disposal as ropriate.
erials that enter the radiologically controlled area are verified as nonradioactive before being ased for reuse or disposal. Tools and equipment belonging to personnel and contractors are eyed at the radiologically controlled area exit in the annex building. If these items cannot be ased or decontaminated, they become plant inventory or dry active waste and are handled as cribed previously.
er wastes generated in the radiologically controlled area but outside of surface contamination as are collected in bags or containers and are delivered to the temporary storage location in the 11.4-9 Revision 1
.2.3.4 Mixed Waste Processing Operations ed wastes from the radiologically controlled area are collected in suitable containers and brought e radwaste building, where separate containment pallets and accumulation drums are provided olid and liquid mixed wastes. Mixed wastes are normally sent to an offsite facility having mixed-te processing and disposal capabilities.
.2.4 Waste Processing and Disposal Alternatives
.2.4.1 Portable and Mobile Radwaste Systems Capabilities able or mobile processing and packaging systems can be located in the auxiliary building rail car or the radwaste building mobile systems facility. Chemical wastes are normally processed in the waste building by a mobile concentration and/or solidification system when a batch accumulates e chemical waste tank. Mobile systems are also used to encapsulate high-activity filters, to sort, ontaminate and compact dry active wastes, and to verify nonradioactive wastes.
spent resin system includes connections in the fill station and rail car bay to allow spent resins to elivered to a disposal container in either location for dewatering using portable equipment.
nch Technical Position ETSB 11-3 provides guidance for portable solid waste systems in tion IV. Compliance with the four guidance items is achieved as follows:
IV.I The spent resin tanks are the only tanks that contain a significant volume of wet wastes, and these tanks are permanently installed. Concentrates that may be produced by mobile evaporation systems will be produced and stored by the mobile systems only in small batches prior to being solidified by the mobile systems. As described in Subsection 1.2.7, the radwaste building is designed to retain spillage from mobile or portable systems.
IV.2 Permanently installed piping for transport of radioactive wastes to mobile or portable systems is routed close to the mobile or portable systems thereby minimizing the use of flexible interfacing hose. The hydrostatic test requirements of Regulatory Guide 1.143 will be applied to the flexible interfacing hose.
IV.3 Portable or mobile systems will be located in either the rail car bay of the auxiliary building or in the mobile systems facility in the radwaste building. The spent resin waste container fill station or the shipping cask in the auxiliary building collects spillage of spent resin during waste container filling operations. The radwaste and auxiliary buildings contain and drain spillage to the liquid radwaste system via the radioactive waste drain system as described in Subsection 1.2.7 and Section 11.2. Portable or mobile systems will, when required, have their own HEPA filtered exhaust ventilation system. HEPA filtered exhaust is required when airborne radioactivity would exceed 10 CFR 20 derived air concentration limits for radiation workers. The mobile systems facility has connections on the exhaust ventilation ducts for connecting exhaust duct from mobile or portable processing systems to the building's exhaust ventilation system.
IV.4 Although the seismic criteria of Regulatory Guide 1.143 are not applicable to structures housing mobile or portable solid radwaste systems, the portable equipment used for spent resin container filling and dewatering and high-activity filter cartridge packaging will be housed within the Seismic Category I auxiliary building. The radwaste building, which 11.4-10 Revision 1
.2.4.2 Central Radwaste Processing Facility an alternative to the mobile or portable processes for lower-activity wastes, the wastes may be t to a licensed central radwaste processing facility for processing and disposal. This option uires minimal onsite processing to remove radioactive materials from the waste streams. The tes are loaded into a cargo container. The mobile systems facility includes a designated laydown a, and the mobile systems facility crane may be used to handle a cargo container.
.2.5 Facilities
.2.5.1 Auxiliary Building in and filtration media transfer lines from the various ion exchangers are routed to the spent resin s on elevation 100 - 0 in the southwest corner of the auxiliary building. The spent resin system ps, valves, and piping are located in shielded rooms near the spent resin tanks.
id radwaste system transfer lines to and from the radwaste building are routed to the south wall e auxiliary building where they penetrate and enter into a shielded pipe pit in the base mat of the waste building.
essways in the auxiliary building are used to move the filter transfer casks. This includes filter sfer cask handling from the containment, where the chemical and volume control filters are ted, to the auxiliary building rail car bay, where the filter cartridges are stored and subsequently kaged using mobile equipment. These accessways are also used to move dry active waste from ous collection locations to the radwaste building. Enclosed access is provided between the iliary building and the radwaste building on elevation 100-0 (grade level).
.2.5.2 Radwaste Building radwaste building, described in Section 1.2, houses the mobile systems facility. It also includes waste accumulation room and the packaged waste storage room. These rooms are serviced by mobile systems facility crane.
e mobile systems facility, three truck bays provide for mobile or portable processing systems and waste disposal container shipping and receiving. A shielded pipe trench to each of the truck bays sed to route liquid radwaste supply and return lines from the connections in the shielded pipe pit e auxiliary building wall. Separate areas are reserved for empty (new) waste disposal container age, container laydown, and forklift charging. An area is available near the door to the annex ding for protective clothing dropoff and frisking.
waste accumulation room (pre-processing) is divided as needed, using partitions and portable lding to adjust the storage areas for different waste categories as needed to complement the oactivity levels and volumes of generated wastes. The accumulation room has lockable doors to imize unauthorized entry and inadvertent exposure.
packaged waste storage room may be separated into high- and low-activity areas, using portable lding to minimize exposure while providing operational flexibility. A lockable door is provided to imize unauthorized entry and radiation exposure.
heating and ventilating system for the radwaste building is described in Subsection 9.4.8.
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lear safety evaluation.
.4 Tests and Inspections operational tests are conducted as described in Subsection 14.2.9. Tests are performed to onstrate the capability to transfer ion exchange resins and deep bed filtration media from the ion hangers and filters to the spent resin tanks or directly to a waste disposal container.
operational tests of the solid waste management system components are performed to prepare system for operation.
r plant operations begin, the operability and functional performance of the solid waste agement system is periodically evaluated according to Regulatory Guide 1.143 by monitoring for ormal or deteriorating performance during routine operations. Instruments and setpoints are also brated on a scheduled basis. The preventive maintenance program includes periodic inspection maintenance of active components.
.5 Quality Assurance quality assurance program for design, installation, procurement, and fabrication issues of the d waste management system is in accordance with the overall quality assurance program cribed in Chapter 17.
e the impact of radwaste systems on safety is limited, the extent of control required by endix B to 10 CFR Part 50 is similarly limited. Thus, a supplemental quality assurance program licable to design, construction, installation and testing provisions of the solid radwaste system is blished by procedures that complies with the guidance presented in Regulatory Guide 1.143.
.6 Combined License Information for Solid Waste Management System Process Control Program rocess Control Program (PCP) is developed and implemented in accordance with the mmendations and guidance of NEI 07-10A (Reference 201). The PCP describes the inistrative and operational controls used for the solidification of liquid or wet solid waste and the atering of wet solid waste. Its purpose is to provide the necessary controls such that the final osal waste product meets applicable federal regulations (10 CFR Parts 20, 50, 61, 71, and CFR Part 173), state regulations, and disposal site waste form requirements for burial at a low l waste disposal site that is licensed in accordance with 10 CFR Part 61.
en the disposable media is removed from mobile radwaste processing system, the process trol program is utilized to move the media from the system and place the media into a package able for shipping. The mobile radwaste processing system is not placed back into service until the ia that has been removed is packaged and ready for shipment.
ste processing (solidification or dewatering) equipment and services may be provided by the plant y third-party vendors. Each process used meets the applicable requirements of the PCP.
additional onsite radwaste storage is required beyond that described in this subsection.
le 13.4-201 provides milestones for PCP implementation.
11.4-12 Revision 1
te acceptance criteria (WAC) of the disposal site, 10 CFR 61.55 and 61.56, and the requirements ird party waste processors.
h waste stream process is controlled by procedures that specify the process for packaging, ment, material properties, destination (for disposal or further processing), testing to verify pliance, the process to address non-conforming materials, and required documentation.
ere materials are to be disposed of as non-radioactive waste (as described in section 11.4.2.3.3), final measurements of each package are performed to verify there has not n an accumulation of licensed material resulting from a buildup of multiple, non-detectable ntities. These measurements are obtained using sensitive scintillation detectors, or instruments qual sensitivity, in a low-background area.
cedures document maintenance activities, spill abatement, upset condition recovery, and training.
cedures document the periodic review and revision, as necessary, of the PCP based on changes e disposal site, WAC regulations, and third party PCPs.
.6.2 Third Party Vendors d party equipment suppliers and/or waste processors are required to supply approved PCPs.
d party vendor PCPs describe compliance with Regulatory Guide 1.143 (Reference 7), Generic er 80-09 (Reference 8), and Generic Letter 81-39 (Reference 9). Third party vendor PCPs are renced appropriately in the plant PCP before commencement of waste processing.
.7 References "Shippers-General Requirements for Shipments and Packagings," 49 CFR 173.
"Packaging and Transportation of Radioactive Material," 10 CFR 71.
"Domestic Licensing of Production and Utilization Facilities," 10 CFR 50.
"Standards for Protection Against Radiation," 10 CFR 20.
"Licensing Requirements for Land Disposal of Radioactive Waste," 10 CFR 61.
"USNRC Technical Position on Waste Form," Rev. 1, January 1991.
Regulatory Guide 1.143, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants."
USNRC Generic Letter GL-80-009, "Low Level Radioactive Waste Disposal," dated January 29, 1980.
USNRC Generic Letter GL-81-039, "NRC Volume Reduction Policy (Generic Letter No. 81-39)," dated November 30, 1981.
11.4-13 Revision 1
USNRC, "Minimization of Contamination," 10 CFR 20.1406.
. NEI 07-10A, Generic FSAR Template Guidance for Process Control Program (PCP),
Revision 0, March 2009 (ML091460627).
11.4-14 Revision 1
Expected Expected Maximum Maximum Generation Shipped Solid Generation Shipped Solid Source (ft3/yr) (ft3/yr) (ft3/yr) (ft3/yr) t Wastes ary Resins (includes spent 400(2) 510 1700(4) 2160 ns and wet activated carbon) mical 350 20 700 40 ed Liquid 15 17 30 34 densate Polishing Resin(1) 0 0 206(5) 259 (1)(6) 0 0 540(5) 680 am Generator Blowdown erial (Resin and Membrane) t Waste Subtotals 765 547 3176 3173 Wastes mpactible Dry Waste 4750 1010 7260 1550
-Compactible Solid Waste 234 373 567 910 ed Solid 5 7.5 10 15 (3) 26 9.4(3) 69 ary Filters (includes high 5.2 vity and low activity cartridges)
Waste Subtotals 4994 1417 7846 2544 TAL WET & DRY WASTES 5759 1964 11,020 5717 s:
Radioactive secondary resins and membranes result from primary to secondary systems leakage (e.g., SG tube leak).
Estimated activity basis is ANSI 18.1 source terms in reactor coolant.
Estimated activity basis is breakdown and transfer of 10% of resin from upstream ion exchangers.
Reactor coolant source terms corresponding to 0.25% fuel defects.
Estimated activity basis from Tables 11.1-5, 11.1-7 and 11.1-8 and a typical 30 day process run time, once per refueling cycle.
Estimated volume and activity used for conservatism. Resin and membrane will be removed with the electrodeionization units and not stored as wet waste. See Subsection 10.4.8.
11.4-15 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Br-83 Br-84 1.98E-01 1.98E-02 Br-85 I-129 I-130 I-131 1.42E+02 1.42E+01 I-132 1.04E+01 1.04E+00 I-133 5.29E+01 5.29E+00 I-134 6.89E+00 6.89E-01 I-135 3.49E+01 3.49E+00 Rb-86 Rb-88 9.72E-01 9.72E-02 Rb-89 Cs-134 3.06E+02 3.06E+01 Cs-136 3.16E+00 3.16E-01 Cs-137 4.64E+02 4.64E+01 Cs-138 Ba-137m 4.44E+02 4.44E+01 Cr-51 3.21E+01 3.21E+00 Mn-54 1.04E+02 1.04E+01 Mn-56 Fe-55 1.04E+02 1.04E+01 Fe-59 5.00E+00 5.00E-01 Co-58 2.05E+02 2.05E+01 Co-60 9.59E+01 9.59E+00 Zn-65 3.02E+01 3.02E+00 Sr-89 2.67E+00 2.67E-01 Sr-90 1.13E+00 1.13E-01 Sr-91 1.72E-01 1.72E-02 Sr-92 Ba-140 6.29E+01 6.29E+00 Y-90 Y-91m Y-91 3.74E-06 3.74E-07 11.4-16 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Y-92 Y-93 La-140 Zr-95 2.80E-04 2.80E-05 Nb-95 Mo-99 Tc-99m Ru-103 5.35E-03 5.35E-04 Ru-106 6.37E-02 6.37E-03 Rh-103m Rh-106 Te-132 Te-125m Te-127m Te-127 Te-129m 1.36E-04 1.36E-05 Te-129 Te-131m Total: 2.11E+03 2.11E+02 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-17 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Br-83 7.03E+00 7.03E-01 Br-84 3.42E-01 3.42E-02 Br-85 3.74E-03 3.74E-04 I-129 3.44E-03 3.44E-04 I-130 9.00E+00 9.00E-01 I-131 5.45E+03 5.45E+02 I-132 1.97E+02 1.97E+01 I-133 1.66E+03 1.66E+02 I-134 7.31E+00 7.31E-01 I-135 3.81E+02 3.81E+01 Rb-86 2.97E+01 2.97E+00 Rb-88 2.52E+01 2.52E+00 Rb-89 9.83E-01 9.83E-02 Cs-134 9.57E+03 9.57E+02 Cs-136 1.72E+03 1.72E+02 Cs-137 9.14E+03 9.14E+02 Cs-138 1.06E+01 1.06E+00 Ba-137m 8.66E+03 8.66E+02 Cr-51 3.95E+01 3.95E+00 Mn-54 1.18E+02 1.18E+01 Mn-56 4.75E+01 4.75E+00 Fe-55 1.14E+02 1.14E+01 Fe-59 5.84E+00 5.84E-01 Co-58 3.03E+02 3.03E+01 Co-60 2.45E+02 2.45E+01 Zn-65 Sr-89 4.56E+01 4.56E+00 Sr-90 1.09E+01 1.09E+00 Sr-91 1.16E+00 1.16E-01 Sr-92 9.96E-02 9.96E-03 Ba-140 1.19E+01 1.19E+00 Y-90 1.07E+01 1.07E+00 Y-91m 3.48E-01 3.48E-02 Y-91 5.48E-01 5.48E-02 11.4-18 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Y-92 4.19E-02 4.19E-03 Y-93 9.07E-05 9.07E-06 La-140 1.07E+01 1.07E+00 Zr-95 Nb-95 Mo-99 Tc-99m Ru-103 Ru-106 Rh-103m Rh-106 Te-132 Te-125m Te-127m Te-127 Te-129m Te-129 Te-131m Total: 3.78E+04 3.78E+03 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-19 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Br-83 Br-84 Br-85 I-129 I-130 I-131 6.04E-02 6.04E-03 I-132 I-133 I-134 I-135 Rb-86 Rb-88 Rb-89 Cs-134 2.81E+02 2.81E+01 Cs-136 2.61E-02 2.61E-03 Cs-137 4.61E+02 4.61E+01 Cs-138 Ba-137m 4.61E+02 4.61E+01 Cr-51 3.37E+00 3.37E-01 Mn-54 8.50E+01 8.50E+00 Mn-56 Fe-55 9.75E+01 9.75E+00 Fe-59 1.23E+00 1.23E-01 Co-58 8.51E+01 8.51E+00 Co-60 9.29E+01 9.29E+00 Zn-65 2.34E+01 2.34E+00 Sr-89 8.05E-01 8.05E-02 Sr-90 1.13E+00 1.13E-01 Sr-91 Sr-92 Ba-140 4.80E-01 4.80E-02 Y-90 1.13E+00 1.13E-01 Y-91m Y-91 4.03E-04 4.03E-05 11.4-20 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Y-92 Y-93 La-140 5.52E-01 5.52E-02 Zr-95 1.09E-04 1.09E-05 Nb-95 1.31E-04 1.31E-05 Mo-99 Tc-99m Ru-103 1.10E-03 1.10E-04 Ru-106 5.38E-02 5.38E-03 Rh-103m 1.11E-03 1.11E-04 Rh-106 5.38E-02 5.38E-03 Te-132 Te-125m Te-127m Te-127 Te-129m 2.10E-05 2.10E-06 Te-129 1.37E-05 1.37E-06 Te-131m Total: 1.60E+03 1.60E+02 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-21 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Br-83 Br-84 Br-85 I-129 3.44E-03 3.44E-04 I-130 I-131 4.10E+02 4.10E+01 I-132 I-133 6.27E-08 6.27E-09 I-134 I-135 Rb-86 9.76E+00 9.76E-01 Rb-88 Rb-89 Cs-134 9.31E+03 9.31E+02 Cs-136 3.47E+02 3.47E+01 Cs-137 9.13E+03 9.13E+02 Cs-138 Ba-137m 9.13E+03 9.13E+02 Cr-51 1.86E+01 1.86E+00 Mn-54 1.10E+02 1.10E+01 Mn-56 Fe-55 1.12E+02 1.12E+01 Fe-59 3.66E+00 3.66E-01 Co-58 2.26E+02 2.26E+01 Co-60 2.42E+02 2.42E+01 Zn-65 Sr-89 3.06E+01 3.06E+00 Sr-90 1.09E+01 1.09E+00 Sr-91 Sr-92 Ba-140 2.35E+00 2.35E-01 Y-90 1.09E+01 1.09E+00 Y-91m Y-91 3.90E-01 3.90E-02 11.4-22 Revision 1
Primary Resin Primary Filter Isotope Total Ci/yr Total Ci/yr Y-92 Y-93 La-140 2.70E+00 2.70E-01 Zr-95 Nb-95 Mo-99 Tc-99m Ru-103 Ru-106 Rh-103m Rh-106 Te-132 Te-125m Te-127m Te-127 Te-129m Te-129 Te-131m Total: 2.91E+04 2.91E+03 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-23 Revision 1
Secondary Resin Isotope Total Ci/yr Na-24 1.83E-02 Cr-51 4.29E-02 Mn-54 2.95E-02 Fe-55 2.35E-02 Fe-59 4.49E-03 Co-58 7.78E-02 Co-60 1.03E-02 Zn-65 9.56E-03 Br-84 2.22E-05 Rb-88 8.99E-05 Sr-89 2.24E-03 Sr-90 2.37E-04 Sr-91 2.11E-04 Y-90 2.06E-04 Y-91 2.53E-04 Y-91m 1.82E-04 Y-93 9.80E-04 Zr-95 6.53E-03 Nb-95 5.19E-03 Nb-95m 4.74E-03 Mo-99 1.52E-02 Tc-99m 1.41E-02 Ru-103 1.13E-01 Ru-106 1.65E+00 Rh-103m 1.39E-01 Rh-106 2.11E+00 Ag-110 2.12E-02 Ag-110m 2.45E-02 Te-129 2.29E-03 Te-129m 2.79E-03 Te-131 1.14E-03 Te-131m 1.42E-03 Te-132 4.74E-04 11.4-24 Revision 1
Secondary Resin Isotope Total Ci/yr I-131 1.70E-01 I-132 7.93E-03 I-133 5.23E-02 I-134 1.18E-03 I-135 2.56E-02 Xe-131m Xe-133 Xe-135 Cs-134 2.50E-01 Cs-135 4.70E-10 Cs-136 1.48E-02 Cs-137 3.39E-01 Ba-136m 1.39E-02 Ba-137m 3.42E-01 Ba-140 1.17E-01 La-140 1.47E-01 Ce-141 2.13E-03 Ce-143 2.91E-03 Ce-144 7.35E-02 Pr-143 2.04E-03 Pr-144 6.37E-02 Total: 5.96E+00 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-25 Revision 1
Secondary Resin Isotope Total Ci/yr Na-24 4.62E-04 Cr-51 5.17E-01 Mn-54 3.55E-01 Mn-56 2.24E-01 Fe-55 2.78E-01 Fe-59 5.88E-02 Co-58 9.25E-01 Co-60 1.23E-01 Br-83 3.73E-02 Br-84 1.41E-03 Br-85 1.64E-06 Kr-83m Kr-85 Kr-85m Rb-88 4.56E-02 Rb-89 1.53E-03 Sr-89 9.10E-01 Sr-90 5.00E-02 Sr-91 2.13E-02 Sr-92 7.25E-04 Y-90 4.60E-02 Y-91 4.34E-02 Y-91m 2.11E-02 Y-92 2.66E-03 Y-93 1.04E-03 Zr-95 7.74E-02 Nb-95 8.25E-02 Nb-95m 5.52E-02 Mo-99 1.52E+01 Tc-99m 1.68E+01 Ru-103 6.28E-02 Ru-103m 3.87E-02 Rh-103m 6.29E-02 Rh-106 5.95E-02 11.4-26 Revision 1
Secondary Resin Isotope Total Ci/yr Ag-110 1.34E-02 Ag-110m 2.24E-01 Te-129 1.19E+00 Te-129m 1.10E+00 Te-131 2.35E+00 Te-131m 2.01E-01 Te-132 6.75E+00 Te-134 1.49E-03 I-130 1.19E-01 I-131 1.37E+02 I-132 6.77E+00 I-133 2.51E+01 I-134 4.99E-02 I-135 3.99E+00 Xe-131m Xe-133 Xe-135 Cs-134 6.90E+02 Cs-135 6.16E-08 Cs-136 5.15E+02 Cs-137 5.00E+02 Cs-138 3.41E-02 Ba-136m 6.35E+02 Ba-137m 5.14E+02 Ba-140 2.83E-01 La-140 3.31E-01 Ce-141 6.42E-02 Ce-143 4.94E-03 Ce-144 6.33E-02 Pr-143 4.63E-02 Pr-144 6.33E-02 Total: 3.08E+03 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-27 Revision 1
Secondary Resin Isotope Total Ci/yr Na-24 Cr-51 4.55E-03 Mn-54 2.40E-02 Fe-55 2.19E-02 Fe-59 1.14E-03 Co-58 3.25E-02 Co-60 9.95E-03 Zn-65 7.42E-03 Br-84 Rb-88 Sr-89 6.86E-04 Sr-90 2.36E-04 Sr-91 Y-90 2.31E-04 Y-91 6.71E-09 Y-91m Y-93 Zr-95 2.52E-03 Nb-95 4.06E-03 Nb-95m 2.32E-03 Mo-99 Tc-99m Ru-103 2.34E-02 Ru-106 1.38E+00 Rh-103m 2.87E-02 Rh-106 1.77E+00 Ag-110 1.66E-02 Ag-110m 1.92E-02 Te-129 3.44E-04 Te-129m 4.48E-04 Te-131 Te-131m 11.4-28 Revision 1
Secondary Resin Isotope Total Ci/yr Te-132 I-131 7.32E-05 I-132 I-133 I-134 I-135 Xe-131m Xe-133 Xe-135 Cs-134 2.31E-01 Cs-135 4.86E-10 Cs-136 1.56E-04 Cs-137 3.36E-01 Ba-136m 1.47E-04 Ba-137m 3.40E-01 Ba-140 8.97E-04 La-140 1.05E-03 Ce-141 3.13E-04 Ce-143 Ce-144 5.91E-02 Pr-143 2.38E-05 Pr-144 5.12E-02 Total: 4.38E+00 es shown as Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-29 Revision 1
Secondary Resin Isotope Total Ci/yr Na-24 Cr-51 5.47E-02 Mn-54 2.89E-01 Mn-56 Fe-55 2.60E-01 Fe-59 1.50E-02 Co-58 3.87E-01 Co-60 1.19E-01 Br-83 Br-84 Br-85 Kr-83m Kr-85 Kr-85m Rb-88 Rb-89 Sr-89 2.79E-01 Sr-90 4.96E-02 Sr-91 Sr-92 Y-90 5.12E-02 Y-91 1.12E-06 Y-91m Y-92 Y-93 Zr-95 2.98E-02 Nb-95 5.19E-02 Nb-95m 2.70E-02 Mo-99 2.72E-09 Tc-99m 3.04E-09 Ru-103 1.30E-02 Ru103m 3.27E-02 11.4-30 Revision 1
Isotope Total Ci/yr Rh-103m 1.30E-02 Rh-106 5.03E-02 Ag-110 1.05E-02 Ag-110m 1.76E-01 Te-129 1.92E-01 Te-129m 1.77E-01 Te-131 Te-131m Te-132 2.90E-08 Te-134 I-130 I-131 5.94E-02 I-132 2.36E-08 I-133 I-134 I-135 Xe-131m Xe-133 Xe-135 Cs-134 6.35E+02 Cs-135 6.36E-08 Cs-136 5.42E+00 Cs-137 4.98E+02 Cs-138 Ba-136m 6.69E+00 Ba-137m 5.11E+02 Ba-140 2.18E-03 La-140 2.87E-03 Ce-141 9.41E-03 Ce-143 Ce-144 5.08E-02 Pr-143 4.75E-04 Pr-144 5.08E-02 Total: 1.66E+03 es shown as "" Ci/yr are those calculated to be lower than 1.0E-10 Ci/yr, and thus considered to have insignificant ibutions to total.
11.4-31 Revision 1
ks nt resin tank Number 2 otal volume (ft3) 300 ype Vertical, conical bottom, dished top Design pressure (psig) 15 Design temperature (°F) 150 Material Stainless steel mps sin mixing pump Number 1 ype Pneumatic diaphragm Design pressure (psig) 125 Design temperature (°F) 150 Design flow rate (gpm) 120 Design head (ft) 160 Air supply pressure (psig) 100 Air consumption (scfm) 130 Material Stainless steel housing, Buna N diaphragms sin transfer pump Number 1 ype Material handling positive displacement Design pressure (psig) 125 Design temperature (°F) 150 Design flow rate (gpm) 100 Material Stainless steel housing, Buna N flexible parts 11.4-32 Revision 1
ers sin fines filter Number 1 ype Filter cartridge for inside to outside flow Design pressure (psig) 150 Design temperature (°F) 150 Design flowrate (gpm) 120 iltration rating 10 microns Material Stainless steel housing and pleated polypropylene cartridge with stainless steel screen outer jacket mpler sin sampling device Number 1 ype Inline sampler, positive displacement sample collection and portable pig for sample jar Material Stainless steel and EPDM wetted parts 11.4-33 Revision 1
WLS 1&2 - UFSAR Figure 11.4-1 Waste Processing System Flow Diagram 11.4-34 Revision 1
orne monitoring, and continuous indication of the radiation environment in plant areas where h information is needed. Radiation monitors that have a safety-related function are qualified ironmentally, seismically, or both. Class 1E radiation monitors conform to the separation criteria cribed in Subsection 8.3.2 and to the fire protection criteria described in Subsection 9.5.1.
ipment qualification requirements, including seismic qualification requirements, and general tion information for radiation monitors are listed in Section 3.11. Seismic Categories for the dings housing radiation monitors are listed in Section 3.2.
radiation monitoring system is installed permanently and operates in conjunction with regular special radiation survey programs to assist in meeting applicable regulatory requirements. The ation monitoring system is designed in accordance with ANSI N13.1-1969. The process monitors designed in accordance with ANSI-N42.18-1980.
radiation monitoring system is divided functionally into two subsystems:
Process, airborne, and effluent radiological monitoring and sampling Area radiation monitoring
.1 Design Basis
.1.1 Safety Design Basis le the radiation monitoring system is primarily a surveillance system, certain detector channels orm safety-related functions. The components used in these channels meet the qualification uirements for safety-related equipment as described in Subsection 7.1.4.
nnel and equipment redundancy is provided for safety-related monitors to maintain the safety-ted function in case of a single failure.
design objectives of the radiation monitoring system during postulated accidents are:
Initiate containment air filtration isolation in the event of abnormally high radiation inside the containment (High-1)
Initiate normal residual heat removal system suction line containment isolation in the event of abnormally high radiation inside the containment (High-2)
Initiate main control room supplemental filtration in the event of abnormally high particulate, iodine, or gaseous radioactivity in the main control room supply air (High-1)
Initiate main control room ventilation isolation and actuate the main control room emergency habitability system in the event of abnormally high particulate or iodine radioactivity in the main control room supply air (High-2)
Provide long-term post-accident monitoring (using both safety-related and nonsafety-related monitors) scope of the radiation monitoring system for post-accident monitoring is set forth in General ign Criterion 64 and in the provisions of Regulatory Guide 1.97.
11.5-1 Revision 1
ide:
Equipment to meet the applicable regulatory requirements for both normal operation and transient events Data to aid plant health physics personnel in limiting release of radioactivity to the environment and limiting exposure of operation and maintenance personnel to meet ALARA (as-low-as-reasonably-achievable) guidance Early indication of a system or equipment malfunction that could result in excessive radiation dose to plant personnel or lead to plant damage Data collection and data storage to support compliance reporting for the applicable NRC requirements and guidelines, such as General Design Criterion 64 and Regulatory Guide 1.21 and Regulatory Guide 4.15, Revision 1.
Exhausts to the environment from the personnel areas in the annex building, electrical and mechanical equipment rooms in the annex and auxiliary buildings, and the diesel generator rooms will not be radioactive because they contain no radioactive materials. These ventilation exhausts are not monitored.
.2 System Description
.2.1 Radiation Monitoring System radiation monitoring system uses distributed radiation monitors, where each radiation monitor sists of one or more radiation detectors and a dedicated radiation processor.
h radiation processor receives, averages and stores radiation data and transmits alarms and data e plant control system (protection and safety monitoring system for safety-related monitors) for trol (as required), display and recording. These alarms include: low (fail), alert, and high. Selected nnels have a rate-of-rise alarm. Storage of radiation readings is provided.
h radiation detector, except the in-duct radiation detectors and the containment high range ion mbers, has a check source that is actuated from the associated local radiation processor. The ck source is used to verify detector and monitor operation. The check source is shielded to meet RA requirements, and returns to its fully retracted/shielded position upon loss of actuator power.
ck sources on detectors can be actuated from the main control room. The in-duct radiation ctor operation may be checked using an internal LED to simulate light pulses emitted in onse to radiation. The containment high range monitors have an internal source that provides a imum reading; loss of signal from the detector indicates detector inoperability.
iation monitoring data, including alarm status, are provided to AP1000 operators via the plant trol system (and the protection and safety monitoring system for Class 1E monitors). The rmation is available in either counts per minute (count rate), microCuries/cc (activity centration), or R/hr (radiation dose rate).
ety-related channels are environmentally qualified and are powered from the Class 1E dc and terruptible power supply system. Nonsafety-related channels are powered from the non-ss 1E dc and uninterruptible power supply system.
11.5-2 Revision 1
itoring for the four functional classifications listed below. Individual monitors may provide tionality in more than one of these classifications.
Fluid process monitors determine concentrations of radioactive material in plant fluid systems Airborne monitors provide operators with information on concentrations of radioactivity at various points in the ventilation system, providing information on airborne concentrations in the plant Liquid and gaseous effluent monitors measure radioactive materials discharged to the environs Post-accident monitors monitor potential pathways for release of radioactive materials during accident conditions area radiation monitoring subsystem provides plant personnel information on radiation at fixed tions in AP1000. Post-accident monitoring functions are also performed by certain area monitors.
.2.3 Monitor Descriptions offline gaseous monitors, the radiation monitor includes a low pressure drop flow sensor suitable measuring the sample flow. The radiation processor receives an analog signal input from this flow sor. This signal is used by the radiation processor to control sample flow. The analog signal is smitted to the plant control system (protection and safety monitoring system for safety-related itors). For offline liquid monitors, a flow indicator is provided for manual adjustment of the flow.
se airborne radiation monitors which monitor plant areas which may be occupied by plant onnel will be capable of detecting 10 DAC-hours. The specific radiation monitors which are uded in this category are identified in Table 11.5-1.
.2.3.1 Fluid Process Monitors am Generator Blowdown Radiation Monitors steam generator blowdown radiation monitors (BDS-JE-RE010, RE011) measure the centration of radioactive material in the blowdown from the steam generators. One measures ation in the purification process effluent before it is returned to the condensate system. The other sures radioactivity in the blowdown system electrodeionization waste brine before it is harged to the waste water system. The presence of radioactive material in the steam generator down indicates a leak between the primary side and the secondary side of the steam generator.
er to Subsection 5.2.5 for details of leakage monitoring and to Subsections 10.4.8 and 11.2 for ess system details. The steam generator blowdown radiation monitors meet the guidelines of ulatory Guide 1.97 as discussed in Appendix 1A and Section 7.5.
000 has two steam generators, each of which has a blowdown line. Each blowdown line has a t exchanger upstream of the blowdown flow control valve. The steam generator blowdown ation detectors are located in the lines downstream of these heat exchangers. Therefore, the ation monitors do not require a sample cooler.
en its predetermined setpoint is exceeded, each steam generator blowdown radiation monitor ates an alarm in the main control room, initiates closure of the steam generator blowdown 11.5-3 Revision 1
steam generator blowdown radiation monitors use inline gamma-sensitive, thallium-activated, ium iodide scintillation detectors. The steam generator blowdown radiation monitor detector ge and principal isotopes are listed in Table 11.5-1.
arrangement for the steam generator blowdown radiation monitor is shown in Figure 11.5-1.
mponent Cooling Water System Radiation Monitor component cooling water system radiation monitor (CCS-JE-RE001) measures the centration of radioactive material in the component cooling water system. Radioactive material in component cooling water system provides indication of leakage. Refer to Subsection 5.2.5 for ils of leakage monitoring and to Subsection 9.2.2 for process system details.
e concentration of radioactive materials exceeds a predetermined setpoint, the component ling water system radiation monitor initiates an alarm in the main control room.
component cooling water system radiation monitor is an offline monitor that uses a gamma-sitive, thallium-activated, sodium iodide scintillation detector. The range and principal isotopes listed in Table 11.5-1.
arrangement for the component cooling water system radiation monitor is shown in re 11.5-7.
n Steam Line Radiation Monitors main steam line radiation monitors (SGS-JE-RE026A/B and SGS-JE-RE027A/B) measure the centration of radioactive materials in the two main steam lines. Additionally, the main steam line oisotope concentration data are used to calculate releases to the environment if the steam erator safety relief or power operated relief valves release steam to the atmosphere. Each main m line radiation monitor meets the guidelines of Regulatory Guide 1.97 as discussed in endix 1A and Section 7.5. If the concentration of radioactive materials exceeds a predetermined oint, the main steam line radiation monitors initiate alarms in the main control room.
main steam line radiation monitors are positioned adjacent to the steam lines. Each monitor ctor shield is arranged so that the detector sensitive volume is exposed to the radiation inating inside the steam line on which it is located, and is shielded from radiation originating in the r steam line. Radioactive material in the main steam line provides early indication of leakage in form of a steam generator tube leak. Refer to Subsection 5.2.5 for details of leakage monitoring to Section 10.3 for process system details.
main steam line radiation monitor detectors use gamma-sensitive detectors.
h main steam line radiation monitor range and principal isotopes are listed in Table 11.5-1.
arrangement for a main steam line radiation monitor is shown in Figure 11.5-8.
vice Water Blowdown Radiation Monitor service water blowdown radiation monitor (SWS-JE-RE008) measures the concentration of oactive materials in the blowdown flow from the service water system. Upstream of the radiation itor, local grab sampling is available.
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service water blowdown monitor is an inline monitor using a gamma-sensitive, thallium-vated, sodium iodide scintillation detector. The range and principal isotopes are listed in le 11.5-1.
arrangement for the service water blowdown radiation monitor is shown in Figure 11.5-1.
mary Sampling System Liquid Sample Radiation Monitor primary sampling system (PSS) liquid sample radiation monitor (PSS-JE-RE050) measures and cates the concentration of radioactive materials in the samples from the reactor coolant system.
liquid sample radiation monitor's primary function is to indicate elevated sample radiation levels wing a design basis or severe accident. High radiation levels show the need for sample dilution mit operator exposure during sampling and sample transport for analysis. The monitor may also sed to provide early indication of a significant increase in the radioactivity of the reactor coolant cating a possible fuel cladding breach. When a predetermined setpoint is exceeded, the primary pling system liquid sample radiation monitor isolates the sample flow by closing the outside tainment isolation valve and initiates an alarm in the main control room and locally to alert the rator. Refer to Subsection 9.3.3 for system details.
primary sampling system liquid sample radiation monitor utilizes a gamma-sensitive radiation ctor that is adjacent to the sampling line immediately downstream of the sample cooler. The ge and principal isotopes are listed in Table 11.5-1.
arrangement for the primary sampling system liquid sample radiation monitor is shown in re 11.5-8.
mary Sampling System Gaseous Sample Radiation Monitor primary sampling system gaseous sample radiation monitor (PSS-JE-RE052) measures the centration of radioactive materials in the gaseous samples taken from containment atmosphere.
gaseous sample radiation monitor is used to provide indication of significant radioactivity in the eous sample being taken and the need for dilution of the sample to limit operator exposure during pling and transport for analysis. When a predetermined setpoint is exceeded, the primary pling system gaseous sample radiation monitor initiates an alarm locally and in the main control m to alert the operator. Refer to Subsection 9.3.3 for system details.
primary sampling system gaseous sample radiation monitor utilizes a gamma-sensitive radiation ctor that is adjacent to the sampling line immediately upstream of the sample bottle. The range principal isotopes are listed in Table 11.5-1.
arrangement for the primary sampling system gaseous sample radiation monitor is shown in re 11.5-8.
n Control Room Supply Air Duct Radiation Monitors main control room supply air duct radiation monitors (particulate detectors VBS-JE-RE001A and
-JE-RE001B, iodine detectors VBS-JE-RE002A and VBS-JE-RE002B, and noble gas ctors VBS-JE-RE003A and VBS-JE-RE003B) are offline monitors that continuously measure the centration of radioactive materials in the air that is supplied to the main control room by the lear island nonradioactive ventilation system air handling units. The control support area tilation is also part of this air supply system. The air supply is partially outside air. Refer to section 9.4.1 for system details. The main control room supply air duct radiation monitors receive 11.5-5 Revision 1
n control room emergency habitability system on High-2 particulate or iodine concentrations.
ms are also provided in the main control room for these high concentrations.
main control room supply air duct radiation monitor components are qualified environmentally seismically in accordance with the guidelines of Regulatory Guides 1.89 and 1.100, respectively.
h monitor meets the guidelines of Regulatory Guide 1.97 as discussed in Appendix 1A and tion 7.5.
particulate detectors are beta-sensitive scintillation detectors that view a fixed filter. The iodine ctors are gamma-sensitive, thallium-activated, sodium iodide scintillation detectors that view a d charcoal filter. The gas detectors are beta-sensitive scintillation detectors. The range and cipal radioisotopes are listed in Table 11.5-1.
arrangement for a main control room supply air duct radiation monitor is shown in Figure 11.5-6.
ntainment Air Filtration Exhaust Radiation Monitor containment air filtration exhaust radiation monitor (VFS-JE-RE001) measures the concentration adioactive materials in the containment purge exhaust air.
monitor provides an alarm in the main control room when the concentration of radioactive gases e exhaust exceeds a predetermined setpoint. Refer to Subsection 9.4.7 for system details.
containment air filtration exhaust radiation monitor is an inline monitor that uses a beta-sensitive tillation detector. It is located downstream of the containment air filtration units with its sensitive me inside the duct. The detector range and principal radioisotopes are listed in Table 11.5-1.
arrangement of the containment air filtration exhaust radiation monitor is shown in Figure 11.5-5.
eous Radwaste Discharge Radiation Monitor gaseous radwaste discharge radiation monitor (WGS-JE-RE017) measures the concentration of oactive materials in the releases from the gaseous radwaste system to the plant vent. The surement is made before the discharge reaches the plant vent or is diluted by any other flows.
gaseous radwaste discharge radiation monitor provides an alarm in the main control room and inates the release of radioactive gas to the plant vent by closing the discharge isolation valve n a predetermined setpoint is exceeded. Refer to Section 11.3 for system details.
monitor is an inline monitor using a beta-sensitive scintillation detector with its sensitive volume de the piping. The range and principal isotopes are listed in Table 11.5-1.
arrangement for the gaseous radwaste discharge radiation monitor is shown in Figure 11.5-1.
ntainment Atmosphere Radiation Monitor containment atmosphere radiation monitor measures the radioactive gaseous (PSS-JE-RE026)
F18 particulate (PSS-JE-RE027) concentrations in the containment atmosphere. The tainment atmosphere radiation monitor is a part of the reactor coolant pressure boundary leak ction system described in Subsection 5.2.5. The presence of gaseous or F18 radioactivity in the tainment atmosphere is an indication of reactor coolant pressure boundary leakage. Refer to section 5.2.5 for further details. Conformance with Regulatory Guide 1.45 is discussed in endix 1A.
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radiogas detector is a beta-sensitive scintillation detector. The F18 particulate detector is also a
-sensitive scintillation detector. The ranges and principal isotopes are listed in Table 11.5-1.
arrangement for the containment atmosphere radiation monitor is shown in Figure 11.5-3.
.2.3.2 Airborne Monitors l Handling Area Exhaust Radiation Monitor fuel handling area exhaust radiation monitor (VAS-JE-RE001) measures the concentration of oactive materials in the exhaust air from the fuel handling area. This radiation monitor is located tream of the exhaust air isolation damper.
en a predetermined setpoint is exceeded, the fuel handling area exhaust radiation monitor ides signals to alarm in the main control room, to initiate closure of the fuel handling area supply exhaust air isolation dampers, to open the fuel handling area exhaust air isolation damper to the tainment air filtration exhaust units, and to start a containment air filtration exhaust unit. These ons provide a filtered air path from the fuel handling area to the plant vent. Refer to section 9.4.3 for system details.
fuel handling area exhaust radiation monitor is an inline monitor that uses a beta-sensitive tillation detector. It is located with the sensitive volume inside the exhaust duct. The range and cipal isotopes are listed in Table 11.5-1.
arrangement for the fuel handling area exhaust radiation monitor is shown in Figure 11.5-5.
iliary Building Exhaust Radiation Monitor auxiliary building exhaust radiation monitor (VAS-JE-RE002) measures the concentration of oactive materials in the radiologically controlled area ventilation system exhaust air from the iliary building. The auxiliary building radiation monitor detector is upstream of the exhaust air ation damper.
en a predetermined setpoint is exceeded, indicating abnormal airborne radiation, the auxiliary ding exhaust radiation monitor provides signals to alarm in the main control room, to initiate ure of the auxiliary building supply and exhaust air isolation dampers, to open the auxiliary ding exhaust air isolation damper to the containment air filtration exhaust units, and to start a tainment air filtration exhaust unit. These actions provide a filtered air path from the auxiliary ding to the plant vent. Refer to Subsection 9.4.3 for system details.
auxiliary building exhaust radiation monitor is an inline monitor that uses a beta-sensitive tillation detector. It is located with the sensitive volume inside the exhaust duct. The range and cipal isotopes are listed in Table 11.5-1.
arrangement for the auxiliary building exhaust radiation monitor is shown in Figure 11.5-5.
nex Building Exhaust Radiation Monitor annex building exhaust radiation monitor (VAS-JE-RE003) measures the concentration of oactive materials in the radiologically controlled area ventilation system exhaust air from the ex building. The annex building exhaust radiation monitor is located upstream of the annex ding exhaust air isolation damper.
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aust air isolation damper to the containment air filtration units, and to start a containment air tion exhaust unit. These actions provide a filtered air path from the annex building to the plant
- t. Refer to Subsection 9.4.3 for system details.
annex building monitor is an inline monitor that uses a beta-sensitive scintillation detector. It is ted with the sensitive volume inside the exhaust duct. The range and principal isotopes are listed able 11.5-1.
arrangement for the annex building exhaust radiation monitor is shown in Figure 11.5-5.
lth Physics and Hot Machine Shop Exhaust Radiation Monitor health physics and hot machine shop exhaust radiation monitor (detector VHS-JE-RE001) sures the concentration of radioactive materials in the exhaust air from the health physics area the hot machine shop. The monitor provides an alarm in the main control room when the centration of radioactive gases in the exhaust exceeds a predetermined setpoint. Refer to section 9.4.11 for system details.
monitor is an offline monitor, located downstream of the exhaust fans, that uses a beta-sensitive tillation detector viewing a fixed particulate filter. The range and principal isotopes are listed in le 11.5-1.
arrangement for the health physics and hot machine shop exhaust radiation monitor is shown in re 11.5-9.
waste Building Exhaust Radiation Monitor radwaste building exhaust radiation monitor (VRS-JE-RE023) measures the concentration of oactive materials in the exhaust air from the radwaste building. The monitor provides an alarm in main control room when radioactive material concentrations in the exhaust duct exceed a determined setpoint. Refer to Subsection 9.4.8 for system details.
monitor is an offline monitor, located downstream of the exhaust fans, that uses a beta-sensitive tillation detector viewing a fixed particulate filter. The range and principal isotopes are listed in le 11.5-1.
arrangement for the radwaste building exhaust radiation monitor is shown in Figure 11.5-9.
.2.3.3 Liquid and Gaseous Effluent Monitors nt Vent Radiation Monitor plant vent radiation monitor measures the concentration of radioactive airborne contamination g released through the plant vent, which is the only design pathway for the release of radioactive erials to the atmosphere. The plant vent radiation monitor sample is provided using an isokinetic pling nozzle assembly that has flow sensors. Heat tracing is provided for the sample line. The itor also provides particulate, iodine, and gaseous grab sampling capability.
plant vent is sampled continuously for the full range of concentrations between normal ditions and those postulated in Regulatory Guide 1.97. The plant vent radiation monitor is a post-dent monitor and meets the guidelines of Regulatory Guide 1.97 and NUREG-0737 as discussed ppendix 1A and Section 7.5. Alarms are provided in the main control room if radioactivity centrations exceed predetermined setpoints. The plant vent radiation monitor also provides data 11.5-8 Revision 1
normal range particulate detector, VFS-JE-RE101, uses a beta-sensitive scintillation detector views a fixed filter. The accident range particulate filter is fixed and identical to the normal range
. The accident range particulate filter is analyzed in an onsite laboratory.
normal range iodine detector, VFS-JE-RE102, is a gamma-sensitive, thallium-activated, sodium de, scintillation detector that views a fixed charcoal filter. The accident range iodine filter is a fixed er zeolite filter. The accident range iodine filter is analyzed in an onsite laboratory.
three radiogas channels measure the entire specified range, with overlap in the detector ranges.
normal range radiogas detector, VFS-JE-RE103, is a beta-sensitive scintillation detector. The dent range radiogas detectors, VFS-JE-RE104A (mid-range) and VFS-JE-RE104B (high-range),
beta/gamma-sensitive detectors with small sensitive volumes compared to the normal range ogas detector.
plant vent radiation monitor detector ranges and principal radioisotopes are listed in le 11.5-1. The arrangement for the plant vent radiation monitor is shown in Figure 11.5-4.
plant vent radiation monitor accepts analog signal inputs from process and sample sensors for t vent effluent flow and temperature. These signals are used to control the sample flow to ntain isokinetic extraction at the sample nozzles, and to calculate concentrations, releases and rates at standard conditions. These analog signals are also used to calculate total process flow, l sample flow, and total discharge for an operator-selected period.
normal range particulate, iodine, and radiogas detectors are deactivated automatically when the channel concentration exceeds the normal range. The sample flow bypasses the normal range ctors and a small portion is extracted for the accident range particulate and iodine sample filters radiogas detectors. This prevents normal range detector damage and allows these detectors to sed to measure the concentrations after they decrease again to within the normal range detector ges.
following design criteria for particulate and iodine collection are applied to the design of the plant t and vent sampling system:
The sample extraction point is located at a sufficient distance downstream of perturbations or flow entry points to provide fully developed flow in the turbulent regime.
The sample extraction point is located between the discharge plane of a fan and the stack exit plane, and is not located close the to the stack exit plane where wind effects significantly influence the velocity profile at the sampling location.
The sample nozzles provide high efficiency transmission ratios (80 to 130%) and an aspiration ratio of 0.80 to 1.50 over the expected normal and off-normal flow range for 10 micron aerodynamic diameter (AD) particles.
The sample line layout includes features to provide particle transport efficiency, including the following:
- Non-reactive materials are used in the construction of sample lines.
- Sample line deposition analyses are performed.
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- Long horizontal runs are avoided.
- Long radius bends are used.
- Heat tracing is included if needed to avoid condensation of water or iodine.
bine Island Vent Discharge Radiation Monitor turbine island vent discharge radiation monitor (TDS-JE-RE001A/B) measures the concentration adioactive gases in the steam and non-condensable gases that are discharged by the condenser uum pumps and the gland seal steam condenser. This measurement provides early indication of age between the primary and secondary sides of the steam generators. The monitor provides an m in the main control room if concentrations exceed a predetermined setpoint. Refer to section 5.2.5 for leakage monitoring details and to Subsections 10.4.2 and 10.4.3 for process em details. The turbine island vent discharge radiation monitor meets the guidelines of ulatory Guide 1.97 as discussed in Appendix 1A and Section 7.5.
turbine island vent discharge radiation monitor provides data for reports of gaseous releases of oactive materials in accordance with Regulatory Guide 1.21. The monitor is an inline monitor that s two beta/gamma-sensitive Geiger-Mueller tubes with overlap in the detector ranges. The range principal isotopes are listed in Table 11.5-1.
arrangement for the turbine island vent discharge radiation monitor is shown in Figure 11.5-1.
uid Radwaste Discharge Radiation Monitor liquid radwaste discharge radiation monitor (WLS-JE-RE229) measures the concentration of oactive materials in liquids released to the environment. The liquid releases are made in batches are mixed thoroughly and sampled. The samples are analyzed on site before discharge to rmine that the discharge is within allowable concentration limits and within allowable totals.
liquid radwaste discharge radiation monitor provides data for reports of liquid releases of oactive materials in accordance with Regulatory Guide 1.21.
liquid radwaste discharge radiation monitor is an inline monitor that provides signals to isolate discharge of liquid radwaste, stop the liquid radwaste system discharge pumps and alarms in the n control room if the concentrations exceed a predetermined setpoint. For process system details r to Section 11.2.
range and principal isotopes are listed in Table 11.5-1. The detector is a gamma-sensitive, lium-activated, sodium iodide scintillation detector.
arrangement for the liquid radwaste discharge radiation monitoring channel is shown in re 11.5-1.
ste Water Discharge Radiation Monitor waste water discharge radiation monitor (WWS-JE-RE021) measures the concentration of oactive materials in the discharge from the waste water system. The waste water discharge ation monitor provides data for reports of liquid releases of radioactive materials in accordance Regulatory Guide 1.21.
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harge to the liquid radwaste system for processing. For process system details refer to section 9.2.9.
range and principal isotopes are listed in Table 11.5-1. The detector is a gamma-sensitive, lium-activated, sodium iodide scintillation detector.
arrangement for the waste water discharge radiation monitor is shown in Figure 11.5-1.
.2.4 Inservice Inspection, Calibration, and Maintenance operability of each radiation monitoring system channel is checked periodically.
and inspection requirements for safety-related channels and certain nonsafety-related channels provided in the Technical Specifications, Chapter 16.
y checks of effluent monitoring system operability are made by observing channel behavior.
ector response is routinely observed with a remotely-positioned check source in accordance with t procedures. Instrument background count rate is also observed to determine proper functioning e monitors. Any detector whose response cannot be verified by observation during normal ration or by using the remotely-positioned check source can have its response checked with a able check source. A record is maintained showing the background radiation level and the ctor response.
bration of the continuous radiation monitors is done with commercial radionuclide standards that e been standardized using a measurement system traceable to the National Institute of ndards and Technology.
.3 Effluent Monitoring and Sampling primary means of quantitatively evaluating the isotopic activities in effluent paths is a program of pling and onsite laboratory measurements. Gross activity measurements provided by the ation monitors described in Subsection 11.5.2.3 are used to determine the activities released in ent paths by calibrating the monitors against normalized laboratory results.
ple points are located on the gaseous effluent radiation monitor skids.
requirements of General Design Criterion 64 are satisfied by the sampling program and the ent radiation monitors described in Subsection 11.5.2.3.
e Energy is extending the existing Duke Energy program for quality assurance of radiological ent and environmental monitoring that is based on Regulatory Guide 4.15, Revision 1, to apply to Nuclear Station. Regulatory Guide 4.15, Revision 1, is a proven methodology for quality urance of radiological effluent and environmental monitoring programs that is acceptable to the C staff as a method for demonstrating compliance with applicable requirements of 10 CFR s 20, 50, 52, 61, and 72. Use of Revision 2 of Regulatory Guide 4.15 would necessitate ducting two separate programs involving the use of common staff, facilities and equipment, which ld create an undue burden and could lead to an increased possibility for human error. Therefore, e Energy commits to use Regulatory Guide 4.15, Revision 1, methodology for Lee Nuclear ion for optimal consistency, efficiency and practicality.
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eous process systems as described in Subsection 11.5.2.3. These radiation monitors address the uirement of General Design Criterion 60 to suitably control the release of radioactive materials in eous and liquid effluents. The sampling program for liquid and gaseous effluents will conform to ulatory Guide 4.15, Revision 1 (See Appendix 1A).
iation monitors are used in the radioactive waste processing systems as described in section 11.5.2.3. These radiation monitors address the requirement of General Design erion 63 to monitor radiation levels in radioactive waste systems.
iation monitors are used in the ventilation systems as described in Subsection 11.5.2.3 to ensure airborne concentrations within the plant are within the limits of 10 CFR 20.
.4.1 Effluent Sampling ent sampling of potential radioactive liquid and gaseous effluent paths is conducted on a periodic is to verify effluent processing meets the discharge limits to offsite areas. The effluent sampling gram provides the information for the effluent measuring and reporting required by CFR 50.36a and 10 CFR Part 20 and implemented through the Offsite Dose Calculation Manual CM) and plant procedures. The frequency of the periodic sampling and analyses described ein are nominal and may be increased as permitted by procedure. Tables 11.5-201 and 11.5-202 marize the sample and analysis schedules and sensitivities, respectively. The information tained in Tables 11.5-201 and 11.5-202 are derived from Regulatory Guide 1.21.
oratory isotopic analyses are performed on continuous and batch effluent releases in accordance the ODCM. Results of these analyses are compiled and appropriate portions are utilized to duce the Radioactive Effluent Release Report.
.4.2 Representative Sampling resentative samples are obtained from well-mixed streams or volumes of effluent liquid through use of proper sampling equipment, proper location of sampling points, and the development and of sampling procedures. The recommendations of ANSI N 42.18 (Reference 203) are considered he selection of instrumentation specific to the continuous monitoring of radioactivity in liquid ents.
pling of effluent liquids is consistent with guidance in Regulatory Guide 1.21. When practical, ent releases are batch-controlled, and prior to sampling, large volumes of liquid waste are mixed, s short a time span as practicable, so that solid particulates are uniformly distributed in the liquid me. Sampling and analysis is performed, and release conditions set, before release. Sample ts are located to minimize flow disturbance due to fittings and other characteristics of equipment components. Sample lines are flushed consistent with plant procedures to remove sediment osits.
resentative sampling of process effluents is attained through sample and monitor locations and hods and criteria detailed in plant procedures.
posite sampling is employed to analyze for hard to measure radionuclides and to monitor ent streams that normally are not expected to contain significant amounts of radioactive tamination. Composite liquid samples are collected in proportion to the volume of each batch of ent release. The composite is thoroughly mixed prior to analysis. Collection periods for posites are as short as practicable and periodic checks are performed to identify changes in 11.5-12 Revision 1
pressure head of the fluid, if available, is used for taking samples. If sufficient pressure head is available to take samples, then sample pumps are used to draw the sample from the process to the detector panels and back to the process.
ing and obtaining representative samples using the radiation monitors described in Section 11.5 be performed in accordance with ANSI N13.1 (Reference 201).
obtaining representative samples in unfiltered ducts, isokinetic probes are tested and used in ordance with ANSI N13.1 (Reference 201).
lytical Procedures ically, samples of process and effluent gases and liquids are analyzed in the station laboratory or n outside laboratory via the following techniques:
Gross alpha/beta counting Gamma spectrometry Liquid scintillation counting ailable" instrumentation and counting techniques change as other instruments and techniques ome available. For this reason, the frequency of sampling and the analysis of samples are eralized in this subsection.
ss alpha/beta analysis may be performed directly on unprocessed samples (e.g., air filters) or on essed samples (e.g., evaporated liquid samples). Sample volume, counting geometry, and nting time are chosen to match measurement capability with sample activity. Correction factors ample-detector geometry, self-absorption and counter resolving time are applied to provide the uired accuracy.
id effluent samples are prepared for alpha/beta counting by evaporation onto steel planchets.
mma analysis may be done on any type of sample (gas, solid or liquid) in a gamma spectrometer.
ated water vapor samples are collected by condensation or adsorption, and the resultant liquid is lyzed by liquid scintillation counting techniques.
iochemical separations are used for the routine analysis of Sr-89 and Sr-90.
id samples are collected in polyethylene bottles to minimize absorption of nuclides onto container s.
.5 Post-Accident Radiation Monitoring radiation monitors listed below meet the guidelines of Regulatory Guide 1.97 and are described ubsections 11.5.2.3 and 11.5.6.2. For further Regulatory Guide 1.97 information refer to endix 1A and Section 7.5.
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Main control room supply air duct radiation monitors Plant vent radiation monitor Turbine island vent discharge radiation monitor Containment high range radiation monitors Primary sampling room area monitor CSA area monitor post-accident sampling system is described in Subsection 9.3.3 and is used to obtain samples onsite laboratory analysis, including radioisotopic analysis, after a postulated accident.
.6 Area Radiation Monitors area radiation monitors are provided to supplement the personnel and area radiation survey isions of the AP1000 health physics program described in Section 12.5 and to comply with the onnel radiation protection guidelines of 10 CFR 20, 10 CFR 50, and 10 CFR 70; and Regulatory des 1.97, 8.2, and 8.8.
ing refueling operations in containment and the fuel handling area, criticality monitoring functions, tated in 10 CFR 70.24, are performed by the area radiation monitors in combination with portable ge monitors.
.6.1 Design Objectives design objectives of the area radiation monitors during normal operating plant conditions and cipated operational occurrences are to:
Measure the radiation intensities in specific areas of AP1000 Warn of uncontrolled or inadvertent movement of radioactive material in AP1000 Provide local and remote indication of ambient gamma radiation and local and remote alarms at key points where substantial changes in radiation flux might be of immediate importance to personnel Annunciate and warn of possible equipment malfunctions and leaks in specific areas of AP1000 Furnish information for radiation surveys Minimize the time, effort, and radiation received by operating personnel during routine maintenance and calibration Incorporate modular design concepts throughout, to provide easy maintenance 11.5-14 Revision 1
ations of area monitor detectors are based on the following criteria:
Area monitors are located in areas that are normally accessible and where changes in normal plant operating conditions can cause significant increases in exposure rates above those expected for the areas.
Area monitors are located in areas that are normally or occasionally accessible where significant increases in exposure rates might occur because of operational transients or maintenance activities.
Area monitors are located to best measure the increase in exposure rates within a specific area and to avoid shielding of the detector by equipment or structural materials.
In the selection of area monitors, consideration is given to the environmental conditions under which the monitor operates.
Area monitors are located to provide access so that minimal maintenance equipment is required and to provide an uncluttered area near the detector and local processing electronics to allow for field alignment and calibration.
area radiation monitors are listed in Table 11.5-2.
.6.2 Post-Accident Area Monitors following area monitors are provided to meet Regulatory Guide 1.97 guidelines as discussed in endix 1A and Section 7.5.
ntainment High Range Radiation Monitor containment high range radiation monitors (PXS-JE-RE160, PXS-JE-RE161, PXS-JE-RE162, PXS-JE-RE163) measure the radiation from the radioactive gases in the containment osphere. The monitors receive safety-related power. The detectors are ion chambers, designed easure the radiation from the radioactive gases inside the containment in accordance with ulatory Guide 1.97 and NUREG-0737. The monitors are qualified environmentally and mically in accordance with the guidelines of Regulatory Guides 1.89 and 1.100, respectively.
containment high range radiation data are displayed in the main control room. When determined setpoints are exceeded, the containment high range radiation monitors provide main trol room alarms and signals to the protection and safety monitoring system for containment air tion isolation and normal residual heat removal system valve closure (refer to Section 7.3 for her details). The containment high range radiation monitors provide data for maintaining a record e gamma radiation intensities after a postulated accident as a function of time, so that the ntory of radioactive materials in the containment volume can be estimated.
range and principal isotopes are listed in Table 11.5-1.
high range radiation detectors are mounted inside the containment on the containment wall in ely separated locations. The locations allow the detectors to be exposed to a significant volume of tainment atmosphere without obstruction so that the readouts are representative of the tainment atmosphere. The arrangement for a containment high range monitor is shown in re 11.5-2.
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hat its readout is representative of the radiation to which the operating personnel are exposed. A l readout, an audible alarm, and visual alarms are provided in the primary sampling room to alert rating personnel to increasing exposure rates. A local readout, an audible alarm, and visual ms are provided outside of the primary sampling room and are visible to operating personnel prior ntry. Indication and alarms are also provided in the main control room.
monitor is an extended range monitor that uses a gamma-sensitive ion chamber. The monitor ge and principal isotopes are listed in Table 11.5-2.
ntrol Support Area (CSA) Area Monitor control support area is the location from which engineering support will be provided to the rators following a postulated accident. The CSA area radiation monitor (RMS-JE-RE016) is ted so that its readout is representative of the radiation to which the support personnel are osed. A local readout, an audible alarm, and visual alarms are provided locally to alert personnel creasing exposure rates. A local readout, an audible alarm, and visual alarms are provided ide of the room and are visible to personnel prior to entry. Indication and alarms are also ided in the main control room.
monitor is a normal range monitor that uses a gamma-sensitive Geiger-Mueller tube. The itor range and principal isotopes are listed in Table 11.5-2.
.6.3 Normal Range Area Monitors mal range area radiation monitors are located in accordance with the location criteria given in section 11.5.6.1. A local readout, an audible alarm, and visual alarms are provided in each itored area to alert operating personnel to increasing exposure rates. Visual alarms are provided ide of each monitored area so that they are visible to operating personnel prior to entry.
cation and alarms are also provided in the main control room.
monitor detectors are gamma-sensitive Geiger-Mueller tubes. The monitors and their ranges are d in Table 11.5-2.
.6.4 Fuel Handling Area Criticality Monitors cality monitoring of the fuel handling and storage areas is performed in accordance with CFR 70.24 by radiation monitors RMS-JE-RE012 and RMS-JE-RE020. The area radiation itoring is augmented during fuel handling operations by a portable radiation monitor on the hine handling fuel. The fuel handling area radiation monitor parameters are provided in le 11.5-2.
permanent criticality monitors are physically separated by a large distance and have overlapping s of view. Each detector's field of view can detect radiation from a fuel criticality accident in the as occupied by personnel where fuel is stored and handled. The criticality monitors do not have a ct line of sight in the new fuel storage pit because the arrangement of new fuel prevents dental criticality. The alarm set points of the radiation monitors are below the sensitivity needed to ct the 10 CFR 70.24 specified 20 rads/minute dose rate in soft tissue of combined gamma and tron radiation from an unshielded source at two meters distance. A criticality excursion will duce an audible local alarm and an alarm in the plant MCR.
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itoring system and radiation monitors from other systems is in accordance with the overall quality urance program described in Chapter 17.
sampling program and the associated monitors conform to Regulatory Guide 4.15, Revision 1 e Appendix 1A).
.7 Preoperational Testing firmation testing on the plant vent will be performed during plant startup to qualify the sample action location.
city profile mapping at the sample extraction point will confirm that the velocity profile, including onic flow, does not substantially affect flow mixing or sample nozzle performance, and is eptable for obtaining a representative sample.
ormance testing with tracer gas and particulates will be performed over normal and selected ormal flow conditions. Tracer gas and particulates testing will confirm an acceptably esentative sample is obtained.
quantitative test acceptance criteria are dependent on the final design of the sampling system.
acceptance criteria will be established prior to testing and will be defined in the test procedures.
set of confirmation tests will be performed for the first plant. For subsequent units, either these s may be performed, or documentation may be used to justify that the plant vent geometry and effluent flow conditions are the same or similar, and that these test results remain applicable.
.8 Combined License Information Offsite Dose Calculation Manual (ODCM) is developed and implemented in accordance with the mmendations and guidance of NEI 07-09A (Reference 202). The ODCM contains the hodology and parameters used for calculating doses resulting from liquid and gaseous effluents.
ODCM addresses operational setpoints, including planned discharge rates, for radiation itors and monitoring programs (process and effluent monitoring and environmental monitoring) he control and assessment of the release of radioactive material to the environment. The ODCM ides the limitations on operation of the radwaste systems, including functional capability of itoring instruments, concentrations of effluents, sampling, analysis, 10 CFR Part 50, Appendix I e and dose commitments, and reporting. The ODCM will be finalized prior to fuel load with site-cific information.
le 13.4-201 provides milestones for ODCM implementation.
process and effluent monitoring and sampling per ANSI N13.1 and Regulatory Guides 1.21 4.15 is addressed in Subsections 11.5.1.2, 11.5.2.4, 11.5.4, 11.5.4.1, 11.5.4.2, and 11.5.6.5.
iological effluent and environmental monitoring is addressed in Subsection 11.5.3 10 CFR Part 50, Appendix I guidelines for maximally exposed offsite individual doses and ulation doses via liquid and gaseous effluents are addressed in Subsections 11.2.3.5 and
.3.4 for liquid and gaseous effluents, respectively.
11.5-17 Revision 1
Facilities.
. NEI 07-09A, Generic FSAR Template Guidance for Offsite Dose Calculation Manual (ODCM) Program Description, Revision 0, March 2009 (ML091050234).
. ANSI N42.18-2004, Specification and Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents.
11.5-18 Revision 1
Detector Type Service Isotopes Nominal Range S-JE-RE010 Steam Generator Blowdown Cs-137 1.0E-7 to 1.0E-2 Ci/cc Electrodeionization Effluent S-JE-RE011 Steam Generator Blowdown Cs-137 1.0E-7 to 1.0E-2 Ci/cc Electrodeionization Brine S-JE-RE001 Component Cooling Water Cs-137 1.0E-7 to 1.0E-2 Ci/cc System S-JE-RE101 Plant Vent Particulate Sr-90 Cs-137 1.0E-12 to 1.0E-7 Ci/cc S-JE-RE102 Plant Vent Iodine I-131 1.0E-11 to 1.0E-6 Ci/cc S-JE-RE103 Plant Vent Gas (Normal Range) Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc S-JE-RE104A / P.V. Extended Range Gas Kr-85 Xe-133 1.0E-4 to 1.0E+2 Ci/cc (Accident Mid Range)
S-JE-RE104B / P.V. Extended Range Gas Kr-85 Xe-133 1.0E-1 to 1.0E+5 Ci/cc (Accident High Range)
S-JE-RE026 Containment Atmosphere Gas Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc (Note 2) Ar-41 N-13 S-JE-RE027 Containment Atmosphere beta- F18 1.0E-10 to 1.0E-5 Ci/cc sensitive scintillation detector (Note 2)
S-JE-050 Primary Sampling Liquid I-131 Cs-137 1.0E-4 to 1.0E+2 Ci/cc S-JE-052 Primary Sampling Gaseous Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc S-JE-RE026A Main Steam Line Kr, Xe, I 1.0E-1 to 1.0E+3 Ci/cc S-JE-RE026B Main Steam Line N-16 30 to 200 gallons per day S-JE-RE027A Main Steam Line Kr, Xe, I 1.0E-1 to 1.0E+3 Ci/cc S-JE-RE027B Main Steam Line N-16 30 to 200 gallons per day S-JE-RE008 Service Water Blowdown Cs-137 1.0E-7 to 1.0E-2 Ci/cc S-JE-RE001A/B / Turbine Island Vent Discharge Kr-85 Xe-133 1.0E-6 to 1.0E+5 Ci/cc (Note 3) (Note 4)
S-JE-RE001 Fuel Handling Area Exhaust Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc (Note 5)
S-JE-RE002 Auxiliary Building Exhaust Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc (Note 5)
S-JE-RE003 Annex Building Exhaust Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc (Note 5) 11.5-19 Revision 1
Detector Type Service Isotopes Nominal Range S-JE-RE001A Main Control Room Supply Air Sr-90 Cs-137 1.0E-12 to 1.0E-7 Ci/cc Duct (Particulate)
(Note 1) (Note 5)
S-JE-RE001B Main Control Room Supply Air Sr-90 Cs-137 1.0E-12 to 1.0E-7 Ci/cc Duct (Particulate)
(Note 1) (Note 5)
S-JE-RE002A MCR Supply Air Duct (Iodine) I-131 1.0E-11 to 1.0E-5 Ci/cc (Note 1) (Note 5)
S-JE-RE002B MCR Supply Air Duct (Iodine) I-131 1.0E-11 to 1.0E-5 Ci/cc (Note 1) (Note 5)
S-JE-RE003A MCR Supply Air Duct (Gas) Kr-85 Xe-133 1.0E-7 to 1.0E-1 Ci/cc (Note 1) (Note 5)
S-JE-RE003B MCR Supply Air Duct (Gas) Kr-85 Xe-133 1.0E-7 to 1.0E-1 Ci/cc (Note 1) (Note 5)
S-JE-RE001 Containment Air Filtration Kr-85 Xe-133 1.0E-7 to 1.0E-2 Ci/cc Exhaust (Note 5)
S-JE-RE001 H.P. & Hot Machine Shop Sr-90 Cs-137 1.0E-12 to 1.0E-7 Ci/cc Exhaust (Note 5)
S-JE-RE023 Radwaste Building Exhaust Sr-90 Cs-137 1.0E-12 to 1.0E-7 Ci/cc (Note 5)
S-JE-RE017 Gaseous Radwaste Discharge Kr-85 Xe-133 1.0E-4 to 1.0E+2 Ci/cc S-JE-RE229 Liquid Radwaste Discharge Cs-137 1.0E-6 to 1.0E-1 Ci/cc S-JE-RE021 Waste Water Discharge Cs-137 1.0E-7 to 1.0E-2 Ci/cc s:
Safety-related Seismic Category I The condenser air removal system (CMS) and the gland seal system (GSS) discharge into the turbine island vents, drains and relief system (TDS). The exhaust from the TDS into the turbine island vent is continuously monitored for radiation.
Turbine island vent radiation monitor includes two G-M tubes with nominal ranges of 1.0E-6 to 1.0E+0 ci/cc and 1.0E-1 to 1.0E+5 ci/cc.
Monitor is sensitive enough to detect 10 Derived Air Concentration (DAC)-hours.
11.5-20 Revision 1
Detector Type Service Nominal Range S-JE-RE160 Containment High Range (Note 3) 1.0E-0 to 1.0E+7 R/hr S-JE-RE161 Containment High Range (Note 3) 1.0E-0 to 1.0E+7 R/hr S-JE-RE162 Containment High Range (Note 3) 1.0E-0 to 1.0E+7 R/hr S-JE-RE163 Containment High Range (Note 3) 1.0E-0 to 1.0E+7 R/hr S-JE-RE008 Primary Sampling Room 1.0E-1 to 1.0E+7 mR/hr S-JE-RE009 Containment Area Personnel Hatch - 1.0E-1 to 1.0E+4 mR/hr (Note 1)
Operating Deck - 135'-3" Elevation S-JE-RE010 Main Control Room 1.0E-1 to 1.0E+4 mR/hr S-JE-RE011 Chemistry Laboratory Area 1.0E-1 to 1.0E+4 mR/hr S-JE-RE012 Fuel Handling Area 1.0E-1 to 1.0E+4 mR/hr (Note 2)
S-JE-RE013 Rail Car Bay/Filter Storage Area 1.0E-1 to 1.0E+4 mR/hr (Note 4)
S-JE-RE014A Liquid and Gaseous Radwaste Area 1 1.0E-1 to 1.0E+4 mR/hr S-JE-RE014B Liquid and Gaseous Radwaste Area 2 1.0E-1 to 1.0E+4 mR/hr S-JE-RE016 CSA Area 1.0E-1 to 1.0E+4 mR/hr S-JE-RE017 Radwaste Bldg. Mobile Systems Facility 1.0E-1 to 1.0E+4 mR/hr (Note 4)
S-JE-RE018 Hot Machine Shop 1.0E-1 to 1.0E+4 mR/hr S-JE-RE019 Annex Staging & Storage Area 1.0E-1 to 1.0E+4 mR.hr S-JE-RE020 Fuel Handling Area 1.0E-1 to 1.0E+4 mR/hr (Note 2)
S-JE-RE021 Containment Area Personnel Hatch - 1.0E-1 to 1.0E+4 mR/hr (Note 1)
Maintenance Level - 100'-0" Elevation s:
Radiation levels are monitored by the permanent containment area radiation monitor and by a portable bridge monitor during refueling operations. The containment area radiation monitor is located to best measure the increase in exposure rates for this area and to provide an alarm locally and in the main control room.
Radiation levels are monitored by the permanent fuel handling area radiation monitors and by a portable bridge monitor during fuel handling operations. The fuel handling area radiation monitors are located to best measure the increase in exposure rates for this area and to provide an alarm locally and in the main control room.
Safety-related Monitors areas used for storage of wet wastes (including processed and packaged spent resins) and dry wastes.
11.5-21 Revision 1
Minimum Sampling Frequency Sampled Stream Medium Frequency eous Continuous A sample is taken within one month of initial criticality, and at least weekly Release thereafter to determine the identity and quantity for principal nuclides being released. A similar analysis of samples is performed following each refueling, process change, or other occurrence that could alter the mixture of radionuclides.
When continuous monitoring shows an unexplained variance from an established norm.
Monthly for tritium.
Batch Release Prior to release to determine the identity and quantity of the principal radionuclides (including tritium).
Filters Weekly.
(particulates)
Quarterly for Sr-89 and Sr-90.
Monthly for gross alpha.
id Continuous Weekly for principal gamma-emitting radionuclides.
Releases Monthly, a composite sample for tritium and gross alpha.
Monthly, a representative sample for dissolved and entrained fission and activation gases.
Quarterly, a composite sample for Sr-89 and Sr-90.
Batch Releases Prior to release for principal gamma-emitting radionuclides.
Monthly, a composite sample for tritium and gross alpha.
Monthly, a representative sample from at least one representative batch for dissolved and entrained fission and activation gases.
Quarterly, a composite sample for Sr-89 and Sr-90.
11.5-22 Revision 1
Minimum Sensitivities Stream Nuclide Sensitivity seous Fission & Activation Gases 1.0E-04 µCi/cc Tritium 1.0E-06 µCi/cc Iodines & Particulates Sufficient to permit measurement of a small fraction of the activity that would result in annual exposures of 15 mrem to thyroid for iodines, and 15 mrem to any organ for particulates, to an individual in an unrestricted area.
Gross Radioactivity Sufficient to permit measurement of a small fraction of the activity that would result in annual air dose of 1) 10 mrad due to gamma, and 2) 20 mrad of beta at any location near ground level at or beyond the site boundary.
uid Gross Radioactivity 1.0E-07 µCi/ml Gamma-emitters 5.0E-07 µCi/ml Dissolved & Entrained Gases 1.0E-05 µCi/ml Gross Alpha 1.0E-07 µCi/ml Tritium 1.0E-05 µCi/ml Sr-89 & Sr-90 5.0E-08 µCi/ml 11.5-23 Revision 1
Figure 11.5-1 Process In-Line Radiation Monitor 11.5-24 Revision 1
Figure 11.5-2 Safety-Related Containment High Range Radiation Monitor 11.5-25 Revision 1
Figure 11.5-3 Containment Atmosphere Radiation Monitor 11.5-26 Revision 1
Figure 11.5-4 Plant Vent Radiation Monitor 11.5-27 Revision 1
Figure 11.5-5 In-Line HVAC Duct Radiation Monitor 11.5-28 Revision 1
Figure 11.5-6 Safety-Related Main Control Room Supply Duct Radiation Monitor 11.5-29 Revision 1
Figure 11.5-7 Liquid Offline Radiation Monitor 11.5-30 Revision 1
Figure 11.5-8 Adjacent to Line Radiation Monitor 11.5-31 Revision 1
Figure 11.5-9 HVAC Duct Particulate Radiation Monitor 11.5-32 Revision 1