ML17212A052
| ML17212A052 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/2017 |
| From: | Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17212A038 | List:
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| References | |
| 17-208 | |
| Download: ML17212A052 (151) | |
Text
Millstone Power Station Unit 2 Safety Analysis Report Chapter 11 MPS2 UFSAR 11-i Rev. 35 CHAPTER 11-RADIOACTIVE WASTE PROCESSING AND RADIATION PROTECTION SYSTEMS Table of ContentsSection Title Page11.1RADIOACTIVE WASTE PROCESSING SYSTEMS.....................................11.1-111.1.1General......................................................................................................11.1-111.1.2Design Bases.............................................................................................11.1-111.1.2.1Functional Requirements..........................................................................11.1-1 11.1.2.2Design Criter ia..........................................................................................11.1-111.1.2.3System Compone nts.................................................................................11.1-311.1.3Liquid Waste Pro cessing System..............................................................11.1-311.1.3.1System Descri ptions.................................................................................11.1-311.1.3.2System Operat ion......................................................................................11.1-711.1.4Gaseous Waste Pro cessing System...........................................................11.1-911.1.4.1System Descri ptions.................................................................................11.1-911.1.4.2System Operat ion....................................................................................11.1-1011.1.4.3Gaseous Release Radiological Consequences........................................11.1-1011.1.4.4Waste Gas System Failure......................................................................11.1-1011.1.4.4.1General....................................................................................................11.1-1011.1.4.4.2Method of An alysis.................................................................................11.1-1111.1.4.4.3Results of An alysis.................................................................................11.1-1111.1.4.4.4Conclusions.............................................................................................11.1-11 11.1.5Solid Waste Proces sing System..............................................................11.1-1111.1.5.1System Descri ptions...............................................................................11.1-1111.1.6System Reliability and Availability........................................................11.1-1411.1.6.1Special Features......................................................................................11.1-14 11.1.6.2Tests and Inspections..............................................................................11.1-1511.1.7References...............................................................................................
11.1-1511.2RADIATION PRO TECTION...........................................................................11.2-111.2.1Final Safety Analysis Report Update........................................................11.2-111.2.2Design Bases.............................................................................................11.2-111.2.3Description................................................................................................11.2-311.2.3.1Containment Shielding.............................................................................11.2-3 11.2.3.2Auxiliary Building Shielding....................................................................11.2-411.2.3.3Control Room Shielding...........................................................................11.2-511.2.3.4Spent Fuel Pool Shielding and Fuel Handling Shielding.........................11.2-511.2.3.5Piping Systems Shielding.........................................................................11.2-511.2.4Health Physics Program............................................................................11.2-611.2.5References.................................................................................................
11.2-6 MPS2 UFSAR Table of Contents (Continued)
Section Title Page 11-ii Rev. 3511.ASOURCE TERMS FOR RADIOACTIVE WASTE PROCESSING AND RELEASES TO THE ENVIRONMENT.........................................................11.A-111.A.1Reactor Coolant Design Basis Radionuclide Activities..........................11.A-111.A.1.1Development of Reactor Core Radionuclide Activities..........................11.A-111.A.1.2Corrosion Pr oducts..................................................................................11.A-211.A.1.3Tritium Production...................................................................................11.A-2 11.A.1.4Fuel Experience.......................................................................................11.A-311.A.2Reactor Coolant Expected Radionuclide Activities.................................11.A-411.A.3Calculation of Liquid and Ga seous Effluent Releases.............................11.A-411.A.3.1Expected Liquid and Gaseous Ra dioactive Effluent Releases................11.A-411.A.3.2Design Basis Liquid and Gaseous Ra dioactive Effluent Releases..........11.A-511.A.4Solid Waste Pro cessing System...............................................................11.A-511.A.4.1Spent Resins.............................................................................................11.A-6 11.A.4.1.1Spent Resins from CV CS Ion Exchanger................................................11.A-611.A.4.1.2Spent Resins from Clean Liquid Was te Processing System Demineralizers
......11.A-611.A.4.1.3Spent Resins from Aerated Liquid Waste Processing System Demineralizer
.....11.A-611.A.4.1.4Spent Resins from Spent Fuel Pool Demineralizer.................................11.A-611.A.4.1.5Contaminated Filter Cartridges................................................................11.A-711.BRADIOACTIVE WASTE PROCESSING OF RELEASES TO ENVIRONMENT
...11.B-111.B.1Bases.........................................................................................................11.B-111.B.2Liquid Waste Pro cessing System..............................................................11.B-111.B.2.1Processing of Clean Liquid Waste............................................................11.B-111.B.2.2Processing of Aera ted Liquid Waste........................................................11.B-211.B.2.3Processing of Secondary Side Liquid Waste............................................11.B-211.B.3Gaseous Waste Processing System...........................................................11.B-311.CDOSES FROM RADIOACTIVE RELEA SES AND COST-BENEFIT ANALYSIS
.11.C-111.C.1Doses to Humans......................................................................................11.C-111.C.2Methods for Calculating Do ses From Liquid Releases............................11.C-111.C.2.1Generalized Equation for Calculating Radiation Doses to Humans via Liquid Pathways...................................................................................................11.C-111.C.2.2Doses from Aquatic Foods.......................................................................11.C-211.C.2.3Doses from Shoreline Deposits.................................................................
11.C-2 MPS2 UFSAR Table of Contents (Continued)
Section Title Page 11-iii Rev. 3511.C.2.4Doses from Swimming and Boating.........................................................11.C-311.C.3Method for Calculating Doses From Gaseous Releases...........................11.C-311.C.3.1Gamma and Beta Doses from Noble Ga s Discharged to th e Atmosphere.11.C-411.C.3.1.1Annual Air Doses from Noble Ga s Releases (Non-Elevated)..................11.C-411.C.3.1.2Annual Total Body Dose from Noble Gas Releases (Non-Elevated).......11.C-511.C.3.1.3Annual Skin Dose from Noble Gas Releases (Non-Elevated).................11.C-511.C.3.1.4Annual Gamma Air Dose and Annual Total Body Dose Due to Noble Gas Releases from Free-Standing Stack s More Than 80 Meters Tall.............11.C-611.C.3.2Doses from Radioiodines and Other Ra dionuclides, Exclusive of Noble Gases, Released to the Atmosphere.....................................................................11.C-611.C.3.2.1Annual Organ Dose Due to External Irradiation from Gr ound Deposition of Radionuclides............................................................................................11.C-611.C.3.2.2Annual Organ Dose from Inhalation of Radionuclides in Air..................11.C-611.C.3.2.3Annual Organ Dose from Ingestion of Atmospherically Released Radionuclides in Food......................................................................................................11.C-711.C.4Comparison of Calculated Annual Ma ximum Individual Doses with Appendix I Design Objectives.....................................................................................11.C-711.C.5General Expression for Population Doses................................................11.C-811.C.6Cost-Benefit An alysis...............................................................................11.C-811.C.6.1Procedure Used for Performing Cost-Benefit Analysis............................11.C-911.C.6.2Augments to the Liquid Radioact ive Waste Processing System............11.C-1011.C.6.3Augments to the Gaseous Radioactive Waste Processing System.........11.C-1011.DEXPECTED ANNUAL INHALATION DOSES AND ESTIMATED AIR CONCENTRATIONS OF RADIOACTIVE ISOTOPES FOR MP2 FACILITIES
....11.D-111.EAIRBORNE ACTIVITY SAMPLING SY STEM FOR CONTAINMENT, SPENT FUEL AND RADWASTE ATMOSPHERES...................................................11.E-1 MPS2 UFSAR 11-iv Rev. 35 CHAPTER 11-RADIOACT IVE WASTE PROCESSI NG AND RADIATION PROTECTION SYSTEMS List of Tables Number Title11.1-1Radioactive Waste Processing System Component Description11.1-2Sources and Expected Volumes of Solid Wastes11.1-3Radioactivity levels of solid wastes (See Note) 11.1-4Curie Inventory of Solid Waste Shi pped from Millstone Unit 2 (See Note)11.1-5Assumptions for Waste Gas Decay Tank Accident 11.2-1Source Terms for Shielding Design 11.A-1Design-Basis and Expected Prim ary Coolant Activity Concentrations11.A-2Calculated Reactor Core Activities 11.A-3Expected Annual Effluent Releases (Cur ies Per Year), by Ra dionuclide, from Each Release Point11.A-4Expected Annual Liquid Effluent Ac tivity Releases (Curies/Year), by Radionuclide, from Each Waste Stream11.A-5Expected Annual Liquid Effluent Concen trations (Diluted and Undiluted), by Radionuclide, from Each Waste System11.A-6Design Basis Radionuclide Concentrations in Liquid E ffluent, in Fractions of 10 CFR Part 20 Concentration Limits11.A-7Design-Basis Radionuclide Ai rborne Concentrations at th e Site Boundary From All Gaseous Effluent Release Points Combined, in Fractions of 10 CFR Part 20 Concentration Limits11.A-8Total Annual Design-Basis and Expected Releases of Radioactive Liquid Waste to the Environment From All Sources Combined, in Curies Per Year11.A-9Total Annual Design Basis and Expected Releases of Airborne Radioactive Waste to the Environment From All Release Points Combined, in Curies Per Year11.A-10Basis for Reactor Coolant System Activity NUREG-0017 gale Code Input11.B-1Inputs to PWR-GALE Code11.C-1Comparison of Calculated Annual Maximum Individual Doses with 10 CFR Part 50 Appendix I Design Objectives11.C-2Annual Total Body and Thyroid Doses to the Population Within 50 Miles of the Millstone Site, In Man-Rem, From Expected Liquid and Airborne Effluent Releases MPS2 UFSAR List of Tables (Continued)
Number Title 11-v Rev. 3511.D-1Containment Building Airborne Concentrations11.D-2Auxiliary Building Airborne Concentrations11.D-3Turbine Building Airborne Concentrations MPS2 UFSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
11-vi Rev. 35 CHAPTER 11-RADIOACT IVE WASTE PROCESSI NG AND RADIATION PROTECTION SYSTEMS List of Figures Figure Title11.1-1P&ID Clean Liquid Radwaste System11.1-2P&ID Clean Liquid Radwaste System 11.1-3P&ID Drains (Containmen t & Auxiliary Building and Auxiliary Yard Sump) (Sheet 1)11.1-4P&ID Aerated Liquid Radwaste System 11.1-5P&ID Diagram Gaseous Radwaste System 11.1-6Degasifier Performance Curve 11.1-7P&ID Spent Resin Radwaste System 11.2-1P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Buil ding - Elevation (-) 45 Feet 6 Inches11.2-2P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Buil ding - Elevation (-) 29 Feet 6 Inches11.2-3P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Buil ding - Elevation (-) 5 Feet 0 Inches11.2-4P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment Building - Elevation 14 Feet 6 Inches and 38 Feet 6 Inches11.2-5P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Auxiliary Building - Elevation 14 Feet 6 Inches and 25 Feet 6 Inches11.2-6P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Turbine Building - Elevation 14 Feet 6 Inches11.2-7P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Auxiliary Building -
Elevation 36 Feet 6 Inches11.2-8P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Section A-A11.2-9P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Section B-B11.2-10P&ID, Radiation Zones and Access Contro l Normal Operation With 1.0% Failed Fuel Containment and Auxiliary11.2-11Neutron Shield Segment 11.2-12Neutron Shielding - Sectional View MPS2 UFSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 11-vii Rev. 3511.2-13Not Used11.2-14Not Used 11.2-15Neutron Shield - Thermal Load ings (Typical Annular Section) 11.2-16Neutron Shield - Thermal Loadings (Plug Shield Section) 11.A-1Escape Rate Coefficients 11.B-1NUREG-0017 Gale Code Input Diagram - Liquid Waste System11.B-2NUREG-0017 Gale Code Input Diagram - Airborne MPS2 UFSAR11.1-1Rev. 35CHAPTER 11 - RADIOACTIVE WAST E PROCESSING AND RADIATION PROTECTION SYSTEMS11.1 RADIOACTIVE WASTE PROCESSING SYSTEMS11.1.1 GENERAL The purpose of the radioactive wa ste processing systems is to provi de controlled handling of all radioactive waste, proper dischar g ing of radioactivity in liquid and gaseous efflue nts, and proper packaging for shipment of solid waste cont aining radioactivity fr om Millstone Unit 2.
On occasion, the Unit will genera te liquid radioactive waste that cannot practicably be processed in the liquid radwaste system. Th e station may process this waste outside the Unit in compliance with state and federal regulations, and in accordance with the Radiological Effluent Control Program outlined in the Administrative Section of the Technical Specifications (e.g., Unit 1 evaporator or shipped off site for processing).11.1.2 DESIGN BASES11.1.2.1 Functional Requirements These systems are required to ensure that the gene ral public and plant pe rsonnel are protected against exposure to radioactive material in accordance with th e regulations of 10 CFR Part 20, and the recommended guidelines of 10 CFR Part 50, Appendix I. Under normal plant operating conditions, the radioactive waste processing systems are required to limit radionuclide release concentrations to unrestricted areas to less than the maximum permissible concentrations as specified in 10 CFR Part 20, Section 1302, and Appendix B.Interim onsite storage facilities accept waste from Millstone Units 1, 2 and 3. Information regarding facility design criter ia is presented in Section 11.4 of the Millstone Unit 3 Final Sa fety Analysis Report.
11.1.2.2 Design CriteriaThe radioactive waste processing system is designed in accordance with the following criteria:a.To protect the general public and plan t operating personnel agai nst radiation from materials in accordance with 10 CFR Part 20 and 10 CFR Part 50, Appendix I.b.To limit the levels of radioactivity of the ef fluents in unrestricted areas to as low as reasonably achievable (ALARA).c.To provide suitable control of the releas e of radioactive materials in gaseous and liquid ef fluents during normal plant ope ration including anticipated operational occurrences.
MPS2 UFSAR11.1-2Rev. 35d.To provide sufficient holdup capacity to retain gaseous and liquid wastes containing radioactive materials.e.To ensure adequate safety under normal and postulated accident conditions.f.To provide the capability to permit inspection and testing of appropriate components.g.To provide suitable shield ing for radiation protection.h.To provide appropriate containment a nd confinement of radioactive materials.i.To provide appropriate monitoring capabil ity to detect excessive radiation levels and to monitor effluent discharge paths.j.To provide suitable processing of li quid and gaseous radwaste generated in accordance with the following operating and design criteria:1.Normal operation with expected pr imary and secondary side activities calculated in accordance with NUREG-0017 Rev. 1 (Reference 11.1-1) using the PWR-GALE code.2.Normal operation with design basis reactor coolant activ ities based on 1%
failed fuel in accordance with Technical Specifications.
The only portions of the radioact ive waste processing system de signated as Seismic Category I are the containment penetration piping and isolation valves. However, the high pressure (150 psig) portions of the gaseous waste system, consisting of th e compressors, decay tanks, interconnecting pipi ng and valves (see Figure 11.1-5), have been designed and analyzed for Seismic Category I requir ements as given in Section 5.8.An analysis had been performed, at the time of initial plant li censing, to determine the site boundary doses due to simultaneous failure of th e entire radioactive waste processing system, excluding those portions of gaseous waste processing system th at were analyzed to Seismic Category I requirements. That anal ysis was predicated on the orig inal design and operation of the radioactive waste processing system. More recently, this chapter was updated to reflect changes to the radioactive waste processing system design and operation. The result s of this update are bounded by those of the original analysis discussed in this subsection.
Failure the of the radioactive waste processing system may release gaseous and liquid wastes into the auxiliary building. The auxiliary building, a Seismic Category I struct ure, is designed to contain all liquids within the building.
Therefore, the total inventory of radioactivity within the gases c ontained in the radioactive waste processing system, excluding high pressure portion of gaseous waste system , are assumed to be released.
MPS2 UFSAR11.1-3Rev. 35The total activity is cons ervatively assumed to be released to the auxiliary build ing at an in stant followed by puff release at ground level to the turbulent wake do wnwind of plant structures.
Under these conditions, th e X/Q value of 9.6 x 10
-5 sec/m 3 is applicable. The doses at 625 meters were calculated by the method of Safety Guides 3 and 4. Based on th e results of the original plant licensing analysis, the site boundary doses due to simultaneous failure of en tire radioactive waste processing system, excluding portions of gaseous waste system, are as follows:
Normal Operation Based on 0.1% Failed Fuel Thyroid dose (rem) - 3.9 x 10
-4 Whole body dose (rem) - 5.04 x 10
-2 Normal Operation Based on 1.0% Failed Fuel Thyroid dose (rem) - 4.35 x 10
-3 Whole body dose (rem) - 5.78 x 10
-1 Since the doses are approxima tely equal to or less th an the limits given in 10 CFR Part 20, Sections 105 an d 106 and Appendix B (version prior to January 1, 1994), a Seismic Noncategory 1 waste processing system is acceptable.11.1.2.3 System Components Descriptions of the radioactive waste pr ocessing system components are given in Table 11.1-1.11.1.3 LIQUID WASTE PROCESSING SYSTEM11.1.3.1 System DescriptionsClean Liquid Waste Processing System:
Clean liquid waste is normally tritiated, non-aer ated, low conductivity liquid waste consisting primarily of reactor coolant let down and liquid waste collected fr om equipment leaks and drains and certain valve and pump seal leaks. The clean liquid radioact ive waste processing system is shown schematically in Figures 11.1-1 and 11.1-2. The design of the cl ean liquid radioactive waste processing system is based upon processing of radioactive liquids postu lated to be released from the reactor coolant system (RCS) during normal reactor opera tion with design basis reactor coolant activities. Operating experience of nuclea r power plants indicate s that Millstone Unit 2 can expect to continue operating with a percentage of fuel failure much less than the postulated design basis of one percent. Nevertheless, the performance of the clea n liquid waste processing system during normal operation is based on both expected and design basis primary side activities.
Discussed in Appendix 11.A are the methodologies used to dete rmine the expected and design basis radionuclide activity concentrations in the reactor coolant.
The clean liquid waste processi ng system is designed to s upport the processing of 14 RCS volumes per year of react or coolant waste. However, the quantity of wastes to be generated and MPS2 UFSAR11.1-4Rev. 35 processed annually by this system is approxi mately 1,200,000 gallons, base d on the assumptions used in the 10 CFR 50 Appendix I analysis. The liquid waste input streams are shown in Figure 11.B-1.The clean liquid waste processing sy stem is designed for the proces sing of reactor coolant wastes concurrently with the letdown fl ow from the chemical and volum e control system (CVCS). This mode of operation, at the maxi mum clean liquid waste processi ng flow of 132 gpm, sets the maximum system flow rate.Reactor coolant is diverted to the clean liqui d radioactive waste processing system from the CVCS when changes in RCS inventory or bor on concentration are n ecessitated by startups, shutdowns, fuel depletion, etc. R eactor coolant at a ra te of 44 gpm to 132 gpm is let down from the CVCS through a filter to reduc e insoluble particulate, after which it flow s to the clean liquid radioactive waste processing sy stem for further processing.
Sources of clean liquid waste in the containment are co llected in the primar y drain tank. A heat exchanger and pump are provided for cooling the primary drain tank content as well as the content of the pressurizer quench tank. The contents of these tanks are maintained or cooled to 120°F to minimize the carryover of ra dioactive moisture to the ga seous waste processing system due to tank venting.
Equipment drains, valve leak-offs, and relief valve dischar ges fr om components that are located in the auxiliary building and that contain liquids with dissolved fission gases are collected in the equipment drain sump tank via the cl osed drain system, as shown in Figure 11.1-3. This design minimizes the uncontrolled release of gaseous radioactivity to the atmosphere.
The liquid contents of both the primary drain tank and the equipm ent drain sump tank are pumped via the demineralizers to the coolant waste receiver ta nk, bypassing the degasifier. The demineralizers and degasifier are discussed below:1.Degasifier Reactor coolant degassing is accomplished by diverting the letdown flow in the CVCS to the degasifier in the clean liquid waste system.
Interconnecting piping and valves are provided for de gassin g the reactor coolant prior to cold shutdown of the reactor for refueling opera tions. The degasifier is pl aced in service, prior to cold shutdown, to remove hydrogen, fissi on product gases, and other dissolved gases from the reactor coolant system liquid and to discharge these gases to the waste gas surge tank in the gase ous waste proce ssing system. (See Section 11.1.4) The degassed liquid is pumped through the de gasifier effluent cooler, which is used to lower the temperature of this liquid to 120
°F before it is passed through one of two primary de mineralizers. This is done to protect liquid radwaste system demineralizer resins against the damage caused by high liquid temperatures. The degassed reactor coolant is re turned to either the volume control tank inlet or to the clean liquid radwaste system primary demineralizers for processing and discharge to the environment.
MPS2 UFSAR11.1-5Rev. 35The packed columnar type degasifier em ploys internally generated stripping steam for removal of dissolved gases. A nitrogen cover gas, slightly above atmospheric pressure, is maintained within the degasi fier during the shutdown mode to prevent air in-leakage and the formation of a potentially explosive hydrogen/oxygen mixture.The degasifier system is provided with cascading steam controls to minimize the adverse effects of varying feed flow rates and temperatures on the dissolved gas removal performance of the uni
- t. The degasifier operates at a pressure of 5 psig and a temperature of 228
°F, and its pumps (both operating at full capacity) are controlled automatically by the degasifier level controls.2.Demineralizers The two (2) primary and two (2) secondary demineralizers are of the mixed-bed non-regenerating type, the de signs of which are based on an expected operating ion exchange capacity of 12,000 grains of CaCO
- 3. The resin beds for mixed-bed non-regenerating demine ralizers are mixtures of cation and anion resins in the H-OH forms. These demineralizers are bypa ssed automatically upon detection of liquid waste temperatures above 135
°F. An additional secondary demineralizer provides the capability to further polish th e waste stream, with the capability of placing the two (2) secondary demi neralizers in series operation.
The demineralizers of the mixed-bed nonregenerable type are sized for one-year operation between resin replacement. Resin replacement is accomplished remotely by N 2 pressure and/or water sl uicing to the spent resin ta nk. All valves required for resin removal are located either outside the demineralizer compartments if non-radioactive, or inside shielded vaults with extension operators to provide protection of plant pe rsonnel if radioactive. The demineralizer effluent flows to one of two coolant wa ste receiver tanks.
The two receiver tanks provide storage for approximately two RCS volumes (120,000 gal.
nominal) of liquid wastes. A nitroge n gas blanket in each of the tanks is automatically maintained abo ve atmospheric pressure to prevent air in-lea kage. The nitrogen cover gas is vented to the gaseous waste processing system or displaced into the other receiver tank or monitor tanks as liquid fills the first receiver tank. No flashing occurs in the rece iver tanks, and any transfer of hydrogen or fission gases from the liquid to the cover gas occurs via the slow process of molecular diffusion.
The content of the coolant waste receiver tank is sampled prior to processing. The content of the coolant waste receiver tank is then pumped through the secondary deminerali zer(s) to one of two coolant waste monitor tanks, which of fers a final check on the liquid waste to be released. The total storage capacity of the monitor tanks is approximately one RCS volume (60,000 gal.
nominal).
MPS2 UFSAR11.1-6Rev. 35 All liquid to be released will be sampled to ensure that the limits set forth in 10 CFR Part 20, Sections 1301 and 1302, and Appendix B are not exceeded. If the ac tivity level is unacceptable after sampling the monitor tank contents, then the liquid is reprocessed through the demineralizers. If the activity le vel is within dischar ge limits, the liquid is pumped through a final filter, at a rate selectable over a range of 10 gpm to 132 gpm, to the circulating water system where it is diluted with the water in the discharge conduit. The proper rate of liquid release to the discharge conduit is determined by sampling the liqui d to be released. The concentrations of the limiting isotopes for release ar e determined. Based on the circ ulating water flow rate, the discharge rate is selected such that the releases to unrestricted areas are within permissible concentrations as specified in 10 CFR Part 20, Appendix B. The second isolation valve in the waste dischar ge header to the ci rculating water system is provi ded with a panel mounted manual controller for setting the desire d flow rate. The clean liquid waste processing system effluent enters the circulating water system discharge box that runs along the south wall of the auxiliary building, as shown in Figure 1.2-10. The routing of the discharge st ructure to the quarry is shown on the plot plan, Figure 1.2-2. The circulating water is furthe r diluted by the Long Island Sound tidal flow.The only liquid discharge path from the clean liq uid waste processing syst em to the environment is through the discharge header, which contains a radiation monitor. Th e radiation monitor and redundant isolation valves are in stalled between the wa ste processing system and the circulating water system. This radiation monitor annunciates in the control room on high radioactivity level and instrument failure, and wi ll automatically close the isolat ion valves to prevent further discharge. The single radiation monitor serves as a backup system to sampli ng of the liquid to be released. All liquid to be released is samp led to confirm that the regulations of 10 CFR Part 20, Sections 1301 and 1302, and Appendix B and 10 CFR Part 50, Appendix I are met. The radiation monitor is described in Section 7.5.6.To prevent the addition of waste to a monitor tank, the content of wh ich is being disc harged to the circulating water system, a valve interlock system is provided. Th e pneumatic valves, located in the discharge piping from each tank, can be ope ned and kept open only if the corresponding valve in the inlet piping to each tank is closed. This arrangement prohibits the addition of wastes to a previously sampled tank whose content is being released.Aerated Liquid Waste Processing System:
The design of the aerated liquid waste processing system, shown in Figure 11.1-4 , is based upon the processing of radioactive liquids (other than those ha ndled by the clean liquid waste processing system) postulated to be generated annually. However, the expected quantity of wastes to be generated and processed annually by this system is approximately 600,000 gallons, based on the assumptions used in the 10 CFR 50 Appendix I analysis.
The aerated liquid waste system design is based upon the pr ocessing of wastes with a radioactivity level resulting fr om normal reactor operation with design basis reactor coolant activity concentrations. The performance of the aerated li quid waste processing system during normal operation is evaluated base d upon processing of wastes with radioactiv ity levels resulting MPS2 UFSAR11.1-7Rev. 35from expected reactor coolant source terms as well as from design basis reactor coolant source terms.Aerated liquid wastes are collected by the open drains system (shown in Figure 11.1-3) that empties into one of two drain ta nks. The system is de signed for batch proces sing. The contents of the drain tanks are monitored by sampling and processed through a sy stem that contains a filter , one hard-piped demineralizer, a nd three portable demineralizers.The demineralizer effluent is collected in the aerated wast e monitor tank, which of fers a final check on the liquid to be release
- d. If the radioactivity level is within discharge limits, as determined by sampling, the liquid is then pumped, at a rate sele cted over a range of 10 gpm to 100 gpm, to the circulating water sy stem, where it is diluted with water in the discharge conduit.
The second isolation valve in the waste discharge header to th e circulating water system is provided with a panel mounted ma nual controller for setting the desired flow rate. The liquid released from the aerated liqui d radioactive waste processing sy stem is monitored continuously for radioactivity. The radiation monitor annunciates in the contro l room on high radioactivity and instrument failure, and will automatically close two isolation valves in the discharge piping to prevent further discharge. The ra diation monitor is discussed in Section 7.5.6. Provisions are furnished for recycling the monitor tank's content for further processing.The only liquid discharge path from the aerated liquid wast e processing system to the environment is through the dischar ge pipe containing the radiation monitor. All system leakage, drains, and relief valve flows are collected in the drains system and return ed to the aerated liquid waste processing system.
11.1.3.2 System OperationClean Liquid Waste Processing Systems:
1.Normal Operation
Letdown from the RCS is automatically dive rted to the radioactive waste processing system on detection of a high volume control tank liquid level.
The flow of wastes into the radioactive waste processing system, normally at a rate of 44 gpm, is intermittent due to changes in RCS inventory or boron concentration necessitated by startups, shutdowns, fuel depletion, etc.
All other liquid wastes input in to the clean liquid waste proces sing system are intermittent due to plant operation. Liquid wastes are collected in either the equipment drain sump tank (EDST) or the primary drain tank. On high tank level, the equipment drain sump tank contents are pumped to the coolant waste receiver tank. During reactor shutdown for refueling, a greater quantity of waste is anti cipated due to draining of system equipment for maintenance or inspection. The operati on of the primary drain tank system, for collection of liquid wastes in the containment, is on a batch-type ba sis. On primary drain tank high tank liquid level, an alarm is annunc iated in the main control room to alert the plant operator. The operation of the prim ary drain tank pumps and opening of the MPS2 UFSAR11.1-8Rev. 35containment isolation valves are initiated manually from th e control room. The initiation of the operation from the main control r oom is required to maintain the tank normal operating temperature of 120
°F.The processing of EDST and le tdown liquid wastes up to the coolant waste receiver tanks is on a fully automatic, unattended basis. Processing up to the coolant waste receiver tanks consists of filtration, degasi fication (when needed), and i on exchange. Degassing the RCS prior to shutdown facilitates s ubsequent opening of the system.
The processing of wastes downstream of the coolant waste receiver tanks occurs on a batch-type basis. The receiver tank content is sampled to determine the amount of further processing required. A remotely actuated reci rculation system is provided for each tank for the purpose of taking a representative sample. Norm al processing consists of demineralization.The processed wastes are collected in the m onitor tanks, which of fer a final check on the liquid to be released. Sampling is employed to determine the required discharge rate to ensure releases are within 10 CFR Part 20, Sections 1301 and 1302 and Appendix B and 10 CFR Part 50, Appendix I limits.2.Abnormal Operation Abnormal operation of the clean liquid waste processing system may involve reactor operation with greater than expected failed fuel and larger than anticipate d volume of liquid waste generated, such as for cold-shutdow ns and startups late in the core cycle. Since the clean liquid radioactive waste processing system is designed for reactor operation with design basis reactor coolant activities for processing 14 RCS volumes of waste per year and with sufficient storage capacity, system operation for these a bnormal occurrences will be the same as for normal operation described above.Aerated Liquid Waste Processing Systems:
1.Normal Operation The aerated liquid waste processing system is operated on a batch type basis. System processing will vary with the volume of wast e generated. The typical processing cycle is expected to be 5 to 7 days.
Normal processing consists of filtration and demineralizat ion. The aerated waste monitor tank provides a final check on the liquid to be released to ensure compliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR 50 Appendix I limits. All potential releas es are sampled prior to discharge to the environment.
MPS2 UFSAR11.1-9Rev. 352.Abnormal Operation Abnormal operating occurrences considered are lar ger than expected volumes of wastes, such as for steam generator blowdown processi ng. If the waste genera tion rate exceeds the capacity of the system, interconnecting pipi ng is provided for pumping the wastes to the clean liquid radioactive waste pr ocessing system for treatment.11.1.4 GASEOUS WASTE PROCESSING SYSTEM11.1.4.1 System Descriptions The gaseous waste processing system, shown in Figure 11.1-5 , processes potentially radioactive hydrogenated waste gases. The system design is based upon the processing of radioactive gases postulated to be releas ed from the RCS during normal operati on with design basis reactor coolant activities. The gaseous waste processing system is shown as one of the streams in Figure 11.B-2. Waste gases flow to the waste ga s header and are coll ected in the waste gas surge tank. When the surge tank pressure increases to approximately 3 psig, one of the two 25 scfm compressors is automatically started by pressure instrumentation located on the tank. The surge tank gases are compressed into one of the six waste gas decay tanks, where gases are stored at a maximum pressure of 150 psig.
The storage capacity of each wast e gas decay tank is based on the st orage of waste gases expected to enter the system during any two months of normal operation. Fo r all the waste gases entering the system, including those contributed by operatio nal occurrences, the six decay tanks provide adequate storage capacity for a decay time of 90 days. Gases are held in the decay tanks until the radioactivity level has been reduc ed by decay and the gases are su itable for release. Prior to release, the gases in the decay tanks are sampled to determine compliance with the regulations of 10 CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR Part 50, Appendix I.The decay tanks discharge through an absolute (i.e., HEPA) filter to th e Millstone stack. The discharge pipe contains a radiation monitor and redundant automatic isolat ion valves. A radiation monitor annunciates in the contro l room on high radioactivity leve l and instrument failure and will automatically close the isolation valves to prevent any further release.The release rate for waste gases is selectable over a range of 10 scfm to 50 scfm. The rate is determined by sampling the cont ent of the decay tank to be dischar ged. The radionuclide concentrations are correla ted with the potential site boundary dose and the 10 CFR Part 20, Appendix B Effluent Concentrations, and th e releas e rate is determined.
The radiation monitor in the wast e gas discharge header is provided as a ba ckup system to the sampling of the decay tank contents. The radiation monitor is described in FSAR Section 7.5.6.The pneumatic valve in the inlet to each decay tank is interlocked with the corresponding valve in the tank discharge piping. The discharge valve can be opened and rema ins open only when the MPS2 UFSAR11.1-10Rev. 35 inlet valve is closed. This feat ure precludes the inadvertent rel ease of waste gases that are not sampled.11.1.4.2 System Operation1.Normal Operation Waste gases from the sources shown in Figure 11.B-2 are collected in the waste gas surge tank through the waste gas header. As the pressu re in the sur ge tank increases to 3 psig, one compressor is automatically started by pressure instrumentation mounted on the surge tank. If the surge tank pressure continue s to increase, the second compressor is automatically started at 5 psig. The compressor discharges the waste gases into one of the six decay tanks selected by the operator.
When the decay tank being filled reaches approximately 140 psig pressure, an alarm is annunciated in the control room to alert the operating personnel. The inlet valve to the decay tank is closed remotely by the operator, and one of the five remaining decay ta nks is selected to receive the gases.
After an appropriate storage period, a nd after sampling has confirmed that 10 CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR 50 Appendix I limits are met, the gases are released on a controlled batch basis to the Millstone stack.2.Abnormal Operation The gaseous waste processing system, unde r abnormal conditions (e.g., unexpectedly lar ge volumes of gases) will function in a fashion similar to that for normal conditions.
Suitable storage capability is available to store the cover gas from the coolant waste receiver and monitor tanks resulting from back-to-back cold shutdowns and startups late in core life.11.1.4.3 Gaseous Release Radiological ConsequencesAnnual design basis and expected releases of radioactive ga seous waste are presented in Table 11.A-9. Gaseous release radiological c onsequences are presented in Appendix 11.C.11.1.4.4 Waste Gas System Failure11.1.4.4.1 GeneralThe limiting accident considered is the postulated and uncontrolled release to the auxiliary building of the radioactive xe non and krypton gases stored in one waste gas decay tank. The credibility of such an occurrence is low since th e waste gas system is not subjected to pressures greater than 150 psig, or large stre sses. The result of a rupture of a gas decay tank is analyzed in order that the maximum hazard, which would result from a malfunction in the radioactive waste process system, will be defined.
MPS2 UFSAR11.1-11Rev. 3511.1.4.4.2 Method of Analysis It is assumed that the tank contains the gaseous activity evolved from degassing one system volume of reactor coolant for refueling. The maximum activity would exist prior to cold shutdown at the end of an operating cycle dur ing which extended operation with one percent defective fuel had occurred. Base d on this and neglecting decay af ter degasification, the noble gas activity in the tank is given in Table 11.1-5.11.1.4.4.3 Results of Analysis(1) The current waste gas system failure analysis is based on updated reactor coolant design activity contained in Table 11.A-1. This analysis results in a whole body dose of 3.0E-01 rem at the EAB and 4.0E-02 rem at the LPZ. The current results are bounded by the licensed values listed above.11.1.4.4.4 ConclusionsIf a waste gas decay tank rupture did occur, the dose would be substantially below 10 CFR Part 100 guidelines.11.1.5 SOLID WASTE PROCESSING SYSTEM11.1.5.1 System DescriptionsThe solid waste processi ng system is designed to provide c ontrolled handling of spent resins, contaminated filter cartridges, and miscellaneous solid waste. The system is designed for handling solid waste with radioactivity levels resulting from re actor operation with design basis reactor coolant source terms. Th e sources and expected annual volumes of so lid wastes are given in Table 11.1-2.The estimated isotopic curie inventory for each source of solid waste is given in Table 11.1-3. The activity data in Table 11.1-3 are based on analyses done for the original licensing of Millstone Unit 2 and are not derive d from the updated radionuclide activ ities shown in th e Appendices to this FSAR chapter.Design of the solid waste processi ng system for handling and disposal of each type of solid waste is as follows:Dose (rems)
OrganExclusion Area Boundary (EAB)
LPZ Thyroid----Whole Body 6.4E-01 (1) 6.6E-02 (1)
MPS2 UFSAR11.1-12Rev. 35A.Spent Resins Spent resins from the radioactive waste pr ocessing system demineralizers, CVCS ion exchangers, and spent fuel pool demin eralizer are replaced in accordance with plant procedures. Resin replacement is accomplished by sluicing the resins from the hard-piped demineralizers and ion exchangers with nitr ogen and/or demineralized water to the spent resin tank. The portable demineralizers may be sluiced with an air/demineralized water mixture to the spent resin tank (SRT), or sluiced directly to a shipping cask, bypassing the SRT.Spent resins are accumulated and stored in th e spent resin tank for radioactive decay prior to filling the disposable container located in a shipping cask at elevation (-)45 feet, 6 inches in the auxiliary building. The spent re sin tank is sized for storing the total volume of resins resulting fr om one resin replacement per year per deminerali zer. With the storage capacity available, minimum storage time of about six months is expected for resins in the tank. The resins are dewatered by the use of a pump after placement in the disposable container.Solid waste containers, shipping casks, methods of packaging, and transportation meet applicable federa l regulations 10 CFR Part 71and 10 CFR 171-178, and wastes are buried at a licen sed burial site in accordance with applicable NRC 10 CFR Part 61. Solid waste treatment design is in complian ce with the requirements of 10 CFR Part 20, Sections 1301 and 1302 as it relates to radioactivity in ef fluents to unrestricted areas.
The spent resin radwaste system is shown in Figure 11.1-7.B.Contaminated Filter Cartridges Filter cartridges for all radioa ctive service filters are normally replaced when the pressure drop across the filter unit exceeds 40 psi. All filters are located in concrete shielded compartments with access provided by a hatch located in the roof of the compartment. (See FSAR Section 11.2 for shielding design.) For the removal of contaminated cartridges, the concrete hatch plug is removed first. Then contaminated cartridges are removed in accordance with plant procedures.The contaminated cartridges are transferred to shielded cont ainers, which are then capped and stored in the drumming area for ultimate off-site disposal.C.Miscellaneous Solid Wastes Contaminated metallic materials and solid objects are placed in disposable shipping containers for transportation to an of f site waste-processi ng facility or disposal site.
Miscellaneous compressible wastes, such as contaminated clothing, rags, paper, etc., are transported to the Millstone Radwas te Reduction Facility for compacting.
MPS2 UFSAR11.1-13Rev. 35 A temporary storage facility for two high-integrity contai ners is provided in the container storage area of the auxiliary building. Temporary storag e for various containe rs prior to further disposition can also be provided depending on curie content and container quantity.
Casks used for shipment of filte r cartridges and miscellaneous soli d wastes are rented as needed.
The containers are designed to prevent loss or di sp ersal of the contents and to maintain self-shielding properties under normal conditions of transport. Where surface dose rates exceed those allowed by the applicable Department of Trans portation (DOT) regulations , the containers are provided with additional shielding. Where the curie content of the containers might exceed the regulatory limit applicable to this type of shipment, special casks lice nsed by DOT and by NRC, if required, would be used.
The shipment of radioactive waste materials is governed by NRC regulations as set forth in 10 CFR Part 71 and regulations of th e U. S. DOT contained in 49 CFR Parts 170 through 178.
The packaging and shipping of all waste from the Millston e site is in accordance with these regulations and any other applicable regulations that may come into effect. All loading of radioactive waste containers is performed by qualified person nel and monitored by radiation protection personnel. The transpor tation is provided by a carrier au thorized by the disposal site operator in accordance with NRC and DOT regulations. The material to be shipped from the site is sent to a licensed waste disposal site or to a licensed waste-processing facility.
Table 11.1-4 gives the total estimated annua l curie inventory of solid wast es to be shipped from the station for off-site burial. This es timated inventory is based on analys es done for the original licensing of Millstone Unit 2 and is not derived from the updated radi onuclide activities shown in the Appendices to FSAR Chapter 11.System Operation 1.Normal Operation
Spent resins from the sources listed in Table 11.1-3 are collected in the spent resin tank for radioactive decay. The spent resins are subs equently loaded into the shipping container and are dewatered by the spent resin sh ipping cask dewatering pump and portable dewatering pumps. After dewatering, the shi pping container and cask are sealed and prepared for shipment.
Prior to a differential pressure of 40 psi being exceeded across any of the filters listed in Table 11.1-3 , the filter is taken out of service, and allowed to decay. To change the filter cartridge, the concrete shield plug above the filter is re moved. The cartridge is removed in accordance with plant procedures, placed in a shielded area and stored for shipment.2.Abnormal Operation No abnormal operations are anticipated.
MPS2 UFSAR11.1-14Rev. 3511.1.6 SYSTEM RELIABILITY AND AVAILABILITY11.1.6.1 Special Features The radioactive waste processing system design is based upon the pr ocessing of wastes postulated to be generated from reactor oper ation with design basis reactor coolant activi ties and the release of these wastes in accorda nce with the requirements 10 CFR Part 20, Sections 1301 and 1302 and Appendix B. For expected radioactivit y levels, sufficient processi ng capability is available to ensure all releases from the radioactive wast e processing system are in accordance with 10 CFR Part 50, Appendix I.The radioactive waste processi ng system is designed for as low as reasonably achievable (ALARA) radioactivity releases to the environment. All radioactive waste processing system vents, equipment drains, leakage, valve stem leadoffs, and relief valve discharges are collected by the drains system and reproces sed by the radioactive waste pr ocessing system.
Those sources containing dissolved fission gases are collected in the equipment drain sump tank and processed as clean liquid waste. All othe r sources are collected and processed by the aerated liquid waste system. All vent gases containing fission gases are discharged to the gaseous waste system for storage and decay.
Each liquid and gaseous waste proc essing system is provided with a single discharge path to the environment. Each discharge header is provided with a radiation monito r and redundant isolation valves which are closed on high radiation level or instrument failure to prevent releases not in compliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B.All releases from the radioactive waste pro cessing system to the environment are to be accomplished on a batch basis for suitable control. All wastes are first sampled to ensure compliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B and 10 CFR Part 50, Appendix I. Adequate sample points are provided in the radioactive waste processing system for suitable control of the processi ng and to ensure that processing equipment performs as designed.
All sampling points in the system are piped to the sampling ro om for analysis. Some local sampling points are utilized in the aerated and clean liquid waste systems.To avoid inadvertent releases prior to sampling, the waste gas decay and coolant waste monitor tanks are provided with inlet a nd outlet valve interlocks. This control provision precludes the draining and filling of th e tanks simultaneously
.All storage tanks in the liquid wa ste system are provided with pipi ng and valves fo r recirculating the tank contents to obtain a re presentative sample. The valves are designed for remote operation to protect station operating personnel from radiation.Sufficient storage capacity is provided for the retention of wastes in the radioactive waste processing system. The three RCS volume storage ca pacity in the clean liquid radioactive waste processing system is sufficient to allow for b ack-to-back cold shutdowns and startups up to 75 percent of the equilibrium core cycle. The six decay tanks allow for 60-day storage of all waste gases, including cover gases from the clean liq uid radioactive waste pr ocessing system tanks.
MPS2 UFSAR11.1-15Rev. 35The radioactive waste processing sy stem is provided with suitable interconnections and sufficient flexibility to allow recirculat ion of waste for additional pr ocessing, if required. Suitable component redundancy is provided to ensure adequate processing.11.1.6.2 Tests and InspectionsAll components of the radioactive waste processing system are nonde structive tested in accordance with the applicable c odes and standards as listed in Table 11.1-1. In addition to code requirements, additional testi ng is required for important co mponents to ensure component integrity and performance.
All pumps in the radioactive waste processing system are manufacturer shop performance tested to demonstrate compliance with desi gn head and capacity requirements.
Prototype filter cartridge test ing was conducted by the filter ma nufacturer confir ming a filter ef ficiency of 92 weight percent when tested with Fine Arizona Air Dust.
The assembled degasifier package was shop gas le ak tested to ensure leak-tightness of all components. The degasifier pack age was subject to a shop perfor mance test to confirm design calculations and performance. T e sting to confirm the performanc e in the removing of dissolved gases was performed with the use of oxygen at feed rate of 60 gpm. Sufficient data were generated with oxygen testing to predict actual pe rformance with resp ect to hydrogen, krypton and xenon removal. The results of these tests are shown in Figure 11.1-6. The testing also confirmed the capability of the degasifier to perform ef fective gas st ripping under conditions of instantaneous feed rate changes and varying feed temperatures.
All system components are visually inspected an d manually adjusted if necessary to ensure correct installation and arrangement. The completely installed system is subject to an acceptance test in accordance with design re quirements. The acceptance test is to check and/or calibrate pumps, valves, instrumentation, interlocks, and system operation. In addition, the completely installed system will be checked for environmental considerat ions, such as co rrect routing of drains and vent conne ctions, leakage, etc.
Individual system component s are located so as to allow acces s for periodic inspection and testing after component decontamination.
Major components are located in individual shielded rooms or compartments to maintain access control.11.
1.7 REFERENCES
11.1-1NUREG-0017 Rev. 1, "Calcula tion of Releases of Radioac tive Materials in Gaseous and Liquid Ef fluents from Pressurized Water Reactors, PWR-GALE Code".
MPS2 UFSAR11.1-16Rev. 35TABLE 11.1-1 RADIOACTIVE WASTE PROCE SSING SYSTEM COMPONENT DESCRIPTIONCLEAN LIQUID WASTE PROCESSING SYSTEMPrimary Drain Tank Pumps Type Inline horizontal centrifu gal with mechanical seals Quantity 2 Design temperature (°F) 250 Design head (TDH) (feet) 150 Design capacity (gpm) 50 NPSH available (feet) 3 to 33 Minimum NPSH required (feet) 3 Material:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Motor 7.5 hp Codes and Standards ASME Section III Class 3 (1971)
Seismic design class 2 Design integrated radiation dosage (rad) 10 6 Degasifier Pumps Type Inline horizontal centrifu gal with mechanical seals Quantity 2 Design temperature (°F) 250 Design head (TDH) (feet) 115 Design capacity (gpm) 132 NPSH available (feet) 2 Minimum NPSH required (feet) 1.5 Material:
MPS2 UFSAR11.1-17Rev. 35 Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Motor (hp) 7.5 Code ASME Section III Class 3 (1971)
Seismic design class 2 Design integrated radiation dosage (rad) 10 6 Coolant Waste Receiver Tank and Monitor Tank Pumps Type Inline horizontal centrifu gal with mechanical seals Quantity 2 Design temperature (°F) 175 Design head (TDH) (feet) 115 Design capacity (gpm) 132 NPSH available (feet) 30 Minimum NPSH required (feet) 10 Material:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Motor (hp) 7.5 Code ASME Section III Class 3 (1971)
Seismic design class 2 Design integrated radiation dosage (rad) 10 6 Equipment Drain Sump Tank Pumps Type Vertical wet pit Quantity 2 Design temperature (°F) 150 Design head (feet) 125 Design capacity (gpm) 50 MPS2 UFSAR11.1-18Rev. 35Material:
Case Type 316 stainless steel Impeller Type 316 stainless steel Shaft Type 316 stainless steel Motor (hp) 10 Code Hydraulic Pump Institute Seismic design class 2 Design integrated radiation dosage (rad) 10 6 Primary Drain Tank and Quench Tank Cooler Pump Type Inline horizontal centrifu gal with mechanical seals Quantity 1 Design temperature (°F) 325 Design head (TDH) (feet) 125 Design capacity (gpm) 100 NPSH available (feet) 3.3-50 Minimum NPSH required (feet) 3 Material:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Motor (hp) 10 Code ASME Section III Class 3 (1971)
Seismic design class 2 Design integrated radiation dosage (rad) 10 6 Primary Drain Tank Type Horizontal
Quantity 1 Volume (gallons net) 1,500 Design pressure (psig) 20 MPS2 UFSAR11.1-19Rev. 35Design temperature (°F) 250 Operating pressure (psig) 3 Operating temperature (°F) 120-150 Material ASTM A-240 TP 304 Design code ASME Section III Class 3 (1971)
Seismic design classification 2 Coolant Waste Receiver Tanks Type Vertical Quantity 2 Volume (gallons net) 60,000 Design pressure (psig) 15 Design temperature (°F) 150 Operating pressure (psig) 3 Operating temperature (°F) 120 Material ASTM A-240 TP 304 Design code:
ASME Section III Class C (1968 Edition includi ng addenda through Summer 1970)
Seismic design classification 2 Coolant Waste Monitor Tanks Type Vertical
Quantity 2 Volume (gallons net) 30,000 Design pressure (psig) 15 Design temperature (°F) 150 Operating pressure (psig) 3 Operating temperature (°F) 120 Material ASTM A-240 TP 304 Design code:
ASME Section III Class C (1968 Edition including Addenda through Summer 1970)
MPS2 UFSAR11.1-20Rev. 35 Seismic design classification 2 Equipment Drain Sump Tank Type Vertical Quantity 1 Volume (gallons net) 500 Design pressure (psig 20 Design temperature (°F) 150 Operating pressure (psig) 3 Operating temperature (°F) 120 Material ASTM A-240 TP 304 Design code ASME Section III Class 3 (1971)
Seismic design classification 2 Degasifier Type Packed column utilizing inte rnally generated stripping steam.
Quantity 1
Design pressure (psig) 50 Design temperature (°F) 250 Operating pressure (psig)
5.3 Operating
temperature (°F) 228 Design capacity (gpm) 40 to 132 Performance See Figure 11.1-6 Design codes:
Components containing radioactive material:
ASME Section III Class C (1968 Ed ition including addend a through summer 1970), ANSI B31.7 Cla ss III, TEMA R.
Other components:
ASME Section VIII, ANSI B31.1 Column: Packing One inch Rashig rings.
Height (feet) 4.25 MPS2 UFSAR11.1-21Rev. 35 Diameter (feet) 2 Materials:
Components containing radioact ive material:
ASTM A-240 TP 304 Other components:
Carbon steel Seismic design classification 2 Coolant Waste
- Demineralizers (Primary and Secondary)
Type Mixed-bed, non-regenerative Quantity 4 (See Note 3)
Design pressure (psig) 100 Design temperature (°F) 150 Operating pressure (psig) 60 Operating temperature (°F) 120 Design flow (gpm) 132 Design code ASME Section III Class 3 (1971)
Material ASTM A-240 TP 304 Seismic design classification 2 Coolant Waste
-Filters Type Disposable cartridge Quantity 2 Design pressure (psig) 200 Design temperature (°F) 250 Operating pressure (psig) 60 Operating temperature (°F) 120 Design flow rate (gpm) 132 Operating flow (gpm)40-132 Filter rating (micron) 3 Design filter efficiency (%)
80 Design code:
ASME Section III Class C (1968 Edition includi ng Addenda through summer 1970).
MPS2 UFSAR11.1-22Rev. 35 Materials: Vessel ASTM A-312 TP 304 Internals TP 304 SS, Micarta Cartridges Ethylene propylene Seismic design classification 2 Coolant Waste
-Piping and Valves Piping: Material ASTM A-312 TP-304 or 316/316L Design pressure (psig) 50, 100, 150 Design temperature (°F) 150, 250, 300, 350 Joints 2.5 inches and larger Butt welded except at flanged equipment. Joints 2 inches and smaller Socket welded except at flanged equipment **
Codes: Fabrication ANSI B31.7 Class III
- Testing and Installation ASME Section III Class 3 (1971) *, ** Valves: ASTM A-182 F304, F316; ASTM A-351 CF8, CF-8M Ratings: 2.5 inches and larger 150 lb ANSI 2 inches and smaller 600 lb ANSI ***
Code: ASME Draft for Pumps and Valves for Nuclear Service (1968) *, **
- Portions of the Clean Liquid Waste Processing System have been replaced with piping and piping components designed, constructed and te sted to the ANSI B31.1 Power Piping Code with augmented requirements per th e guidance in Regulatory Guide 1.143. ** Portions of the Clean Liquid Radwaste System have been desi gned, constructed and tested to the ANSI B31.1 Piping Code with augmented quality requirements per Regulatory Guide 1.143. *** 600 pound ANSI rating represents minimum requirements. 800 pound ANSI rating valves are utilized on a case-by-case basis.
MPS2 UFSAR11.1-23Rev. 35AERATED LIQUID WASTE PROCESSING SYSTEM Aerated Waste Drain and Monitor Tanks Type Vertical Quantity 3 Volume (gallons net) 5,000 Design pressure Atmospheric Design temperature (°F) 150 Operating pressure Atmospheric Operating temperature (°F) 120 Material ASTM A-240 TP 304 Design codes ASME Section III - ASME Section VIII (Original Fabrication)
Seismic design classification 2 Aerated Waste Demineralizer Type Mixed-bed, non-regenerative Quantity 1
Design pressure (psig) 100 Design temperature (°F) 150 Operating pressure (psig) 60 Operating temperature (°F) 120 Design flow (gpm) 132 Normal operating flow (gpm) 50 Design code ASME Section III Class 3 (1971)
Material ASTM A-240 TP 304 Seismic design classification 2 Aerated Waste Portable Demineralizers Type Various, non-regenerative, sluicible Quantity 3
Design Pressure 150 psig at 180
°F MPS2 UFSAR11.1-24Rev. 35Volume Approximately 15 ft 3 each Filters Type Disposable cartridge Quantity 2 Design pressure (psig) 200 Design temperature (°F) 250 Operating pressure (psig) 60 Operating temperature (°F) 120 Design flow rate (gpm) 132 Operating flow (gpm)40-132 Filter rating (micron) 3 Design filter efficiency (%)
80 Design code:
ASME Section III Class C (1968 Edition including Addenda through summer 1970).
Seismic design 2 "A" Aerated Waste Drain Tank Pump Type Inline horizontal centrifu gal with mechanical seals Quantity 1 Design temperature (°F) 180 Design head (TDH) (feet) 238 Design capacity (gpm) 90 NPSH available (feet) 27 Minimum NPSH required (feet) 9
Horsepower (hp) 15 Material:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Codes and standards ASME Section III Class 3 (1971)
MPS2 UFSAR11.1-25Rev. 35 Seismic design classification 2 Design integrated radiation dosage (rads) 10 6 "B" Aerated Waste Drain Tank and Monitor Tank Pumps Type Inline horizontal centrifu gal with mechanical seals Quality 2 Design temperature (°F) 150 Design head (TDH) (feet) 135 Design capacity (gpm) 50 NPSH available (feet) 27 Minimum NPSH required (feet) 5
Horsepower (hp) 5 Material:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 Codes and standards ASME Section III Class 3 (1971)
Seismic design classification 2 Design integrated radiation dosage (rads) 10 6 Piping and Valves Piping: Material ASTM A-316 TP 304 or 316L Design pressure (psig) 50 and 100 Design temperature (°F) 150 Joints: 2.5 inches and larger Butt weld ed except at flanged equipment 2 inches and smaller Socket welded except at flanged equipment Codes: Fabrication ANSI B31.7 Class III
- Testing and Installation ASME Section III Class 3 (1971)
- MPS2 UFSAR11.1-26Rev. 35Valves Material ASTM A-182 F 304, F316 Ratings: 2.5 inches and larger 150 lb ANSI 2 inches and smaller 600 lb ANSI, 800 lb ANSI (Int ermediate Class)
Code ASME Draft for Pumps and Valv es for Nuclear Service (19
- 68)
- Portions of the Aerated Liquid Waste Processing System have b een replaced with piping and piping components designed, constructed and tested to the ANSI B31.1 Power Piping Code with augmented requirements per the guidance in Regulatory Guide 1.143.
MPS2 UFSAR11.1-27Rev. 35GASEOUS WASTE PROCESSING SYSTEM Waste Gas Compressors Type Diaphragm, single state Quantity 2 Capacity (scfm) 25 at 14.7 psia suction pressure Design discharge pressure (psig) 150
Design pressure (psig) 165 Materials:
Heads - ASTM A-105 Grade 2, with ASTM A-240 TP 304 in contact with waste gases.
Diaphragm TP 301 stainless st eel or 316 stainless steel Support cylinder ASTM A-105 Grade 2 Motor (hp) 40 Seismic design classification See Note 1 Design integrated radiat ion level (rads) 3.5 x 10 5 Code: ASME Section VIII, ASME Draft for Pump and Valv es for Nuclear Service, ANSI B31.7. Waste Gas Surge Tank Quantity 1 Type Vertical
Design pressure (psig) 20 Design temperature (°F) 150 Normal operating pressure (psig) 3 to 5
Normal operating temp (°F) 150 Volume (ft
- 3) 582 Material ASTM A-240 TP 304 Seismic design classification 2 Code: ASME Section III Class C (1968 Edition including Addenda through 1970).
MPS2 UFSAR11.1-28Rev. 35Waste Gas Decay Tank Type Vertical Quantity 6
Design pressure (psig) 165 Design temperature (°F) 150 Normal operating pressure (psig) 5 to 150
Normal operating temp (°F) 120 Volume (ft
- 3) 582 Material ASTM A-515 Grade 70 Seismic design classification See Note 1 Code: ASME Section III Class C (1968 Edition including Addenda through summer 1970). Waste Gas Filter Type Disposable cartridge with HEPA filter and demister Quantity 1
Design flow (scfm) 50 Normal operating flow (scfm) 5 - 25 Design pressure (psig) 165 Design temperature (°F) 150 Normal operating pressure (psig) 4
Normal operating temp (°F) 120 Design efficiency 99.97% of particles 0.3 micron and larger Material:
Filter housing ASTM A-240 TP 304 Internals TP 304 stainless steel Cartridge Glass fiber Code ASME Section III Class 3 (1971) Seismic design classification See Note 1 Piping and Valves
MPS2 UFSAR11.1-29Rev. 35 Piping: Material ASTM A-106 Grade B Design pressure (psig) 50, 150 and 200 Design temperature (°F) 150, 200 and 300 Joints: 2.5 inches and larger Butt welded except at flanged equipment. 2 inches and smaller Socket welded except at flanged equipment.
Codes: Fabrication ANSI B31.7 Class III Testing and Installation ASME Section III Class 3 (1971) Design seismic classification 2 (See Note 1) Valves: Materials ASTM A-216, Grade WCB Ratings: 2.5 inches and larger 150 lb ANSI butt weld ends 2 inches and smaller 3,000 lb ANSI socket weld ends Code: ASME Draft for Pup and Va lves for Nuclear Service (1968).
MPS2 UFSAR11.1-30Rev. 35SOLID WASTE PROC ESSING SYSTEM Spent Resin Storage Tank Type Vertical Quantity 1 Volume (ft
- 3) 380 Design pressure (psig) 75 Operating pressure (psig) 35 Design temperature (°F) 175 Operating temperature (°F) 120 Material ASTM A-240 TP 304 Design code ASME Section III Class 3 (1971)
Seismic design classification 2 Spent Resin Shipping Cask Dewatering Pump Type Horizontal Centrifugal Quantity 1 Capacity (gpm) 10 Head (feet) 92 Material Stainless steel Note 1: The high pressure components of the gaseous waste processing system are designated as Seismic Class 2 equipment. However , the co mponents are designed to meet Seismic Class 1 loadings due to the nature of the equipment service.Note 2: The supply lines to the solidification concentrates pump were cut in accordance with PDCR 2-54-95.Note 3: Secondary Demineralizer may be filled with either ion-specific resin or mixed bed resin.
MPS2 UFSAR11.1-31Rev. 35
- 50 cubic foot containers
- 55 gallon drumsAerated Waste hardpiped demineralizer at 42 cu. ft. x 1 = 42 cubic feet Aerated Waste sluicible portable demineralizers at 15 cubic feet/each x 3 = 45 cubic feet Spent Fuel Pool demineralizer at 42 cubic feet x 1 =
42 cubic feet Primary Liquid Radwaste demineralizers at 42 cubic feet x 2 =
84 cubic feet Secondary Liquid Radwaste demineralizer at 42 cubic feet x 1 =
42 cubic feet Letdown Ion Exchangers at 32 cubic feet x 3 =
96 cubic feet 351 cubic feet Note:The data in this Table is not derived from the updated 10 CFR 50 Appendix I analysis. The best source of Solid Waste Volumes can be found in the Annual Radiological Effluent Release Report.TABLE 11.1-2 SOURCES AND EXPECTED VOLUMES OF SOLID WASTES SourceWaste Generating OperationQuantity Per YearPer YearSpent resins One resin replacement per demineralizer/year 351 cu ft 7 *Contaminated filter cartridgesOne cartridge replacement per filter/year 12 cartridges 12 **Miscellaneous solid wastes One 55 gallon drum/week
-52 **
MPS2 UFSARMPS2 UFSAR11.1-32Rev. 35TABLE 11.1-3 RADIOACTIVITY LEVELS OF SOLID WASTES (SEE NOTE)
ION EXCHANGER AND DEMINERALIZER RESINSChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Deborating Ion Exchanger ResinPrimary Clean Liquid Waste Demineralizer Demineralizer ResinAerated Liquid Waste Demineralizer ResinSpent Fuel Pool Clean-up Demineralizer Resin Isotope (Curies)(Li Removal) (Curies)(Curies)(Curies)(Curies)(Curies)NormalMaximumNormalMaximum NormalMaximumNormal Maximum NormalMaximumNormalMaximumCr-51Mn-54Mn-56 Co-58 Fe-59 Co-604.78 x 10-20.4789.6 x 10-3 9.6 x 10-21.2 x 10-3 1.2 x 10-21.12 x 10-4 1.9 x 10-38.72 x 10-51.01 x 10-21.23 12.3Rb-881.36613.66 0.246 2.46 -- --3.1 x 10-35.22 x 10-2 2.4 x 10-30.27967.2 672Rb-893.06 x 10-20.3065.5 x 10-3 5.5 x 10-2 -- --6.9 x 10-51.17 x 10-35.38 x 10-56.25 x 10-31.69 16.9Sr-8913.8138.0 2.398 23.98 -- --3.11 x 10-20.5252.41 x 10-22.800.1341.34Sr-903.05230.52 0.471 4.71 -- --6.77 x 10-30.1145.25 x 10-30.6106.88 x 10-36.88 x 10-2Y-90 -- --2.24 x 10-2 0.224 -- --4.31 x 10-57.26 x 10-43.34 x 10-53.88 x 10-32.69 x 10-20.269Sr-91 6.9 x 10-20.691 1.26 x 10-2 0.126 -- --1.57 x 10-42.65 x 10-31.22 x 10-41.41 x 10-29.39 x 10-20.939 MPS2 UFSARMPS2 UFSAR11.1-33Rev. 35Y-91 -- --7.71 x 10-2 0.771 -- --1.48 x 10-4 2.5 x 10-31.15 x 10-41.33 x 10-22.93 29.3Zr-95Mo-99 -- --52.46524.6 -- --0.101 1.707.82 x 10-29.0853.5 535Ru-103 7.8478.4 1.57 x 10-2 0.157 -- --1.51 x 10-20.2551.17 x 10-21.360.1091.09Ru-106 2.0220.2 0.363 3.63 -- --4.58 x 10-37.72 x 10-23.55 x 10-30.4126.54 x 10-36.54 x 10-2I-1299.36 x 10-49.36 x 10-3 1.56 x 10-4 1.56 x 10-3 -- --2.1 x 10-63.54 x 10-51.63 x 10-61.89 x 10-41.9 x 10-61.9 x 10-5Te-1295.42 x 10-20.542 1.09 x 10-2 0.1091.3 x 10-3 1.3 x 10-21.28 x 10-42.15 x 10-39.88 x 10-51.15 x 10-20.6626.62I-1311,680.016,800.0 336.03,360.0 19.2192.03.9165.9 3.033521051,050I-132 5.0450.4 1.007 10.07 0.121 1.211.19 x 10-2 0.209.19 x 10-31.0726.9 269Te-13249.3493.0 9.86 98.6 0.944 9.440.116 1.958.96 x 10-2 10.48.7 87.0I-133247.02,470.0 4.94 49.4 5.91 59.10.496 8.350.38444.61491,490Cs-134388.03,800.0 517.05,170.0 -- --1.7429.3 1.351562.64 26.4TABLE 11.1-3 RADIOACTIVIT Y LEVELS OF SOLID WASTES (SEE NOTE) (CONTINUED)ION EXCHANGER AND DEMINERALIZER RESINSChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Deborating Ion Exchanger ResinPrimary Clean Liquid Waste Demineralizer Demineralizer ResinAerated Liquid Waste Demineralizer ResinSpent Fuel Pool Clean-up Demineralizer ResinIsotope(Curies)(Li Removal) (Curies)(Curies)(Curies)(Curies)(Curies)NormalMaximumNormalMaximumNormalMaximumNormal MaximumNormalMaximumNormalMaximum MPS2 UFSARMPS2 UFSAR11.1-34Rev. 35I-134 1.0610.6 0.212 2.12 2.54 x 10-2 0.2542.49 x 10-3 4.2 x 10-21.93 x 10-30.22516.3 163Te-1343.19 x 10-20.3196.4 x 10-3 6.4 x 10-2 7.65 x 10-4 7.65 x 10-47.5 x 10-51.26 x 10-35.82 x 10-36.76 x 10-30.6916.91I-13535.9359.0 7.24 72.4 0.869 8.698.46 x 10-2 1.436.56 x 10-27.6271.2 712Cs-136 -- --18.12181.2 -- --3.48 x 10-20.587 2.7 x 10-23.140.6726.72Cs-1371,695.016,950.0 1,728.017,280.0 -- --6.58 111 5.105938.44 84.4Cs-138 -- --0.14 1.40 -- --2.69 x 10-44.54 x 10-32.09 x 10-42.42 x 10-218.2 182Ba-140 4.0840.8 0.743 7.43 -- --9.27 x 10-30.1567.19 x 10-30.8350.1611.61 La-1400.4694.69 9.37 x 10-2 0.937 -- --1.08 x 10-31.82 x 10-28.38 x 10-49.74 x 10-20.1541.54Pr-143 3.4934.9 0.697 6.97 -- --8.05 x 10-30.1366.24 x 10-30.7250.1541.54Ce-14426.2262.0 5.24 52.4 -- --6.04 x 10-2 1.024.68 x 10-25.440.1091.09TABLE 11.1-3 RADIOACTIVIT Y LEVELS OF SOLID WASTES (SEE NOTE) (CONTINUED)ION EXCHANGER AND DEMINERALIZER RESINSChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Purification Ion Exchanger ResinChemical and Volume Control System Deborating Ion Exchanger ResinPrimary Clean Liquid Waste Demineralizer Demineralizer ResinAerated Liquid Waste Demineralizer ResinSpent Fuel Pool Clean-up Demineralizer ResinIsotope(Curies)(Li Removal) (Curies)(Curies)(Curies)(Curies)(Curies)NormalMaximumNormalMaximumNormalMaximumNormal MaximumNormalMaximumNormalMaximum MPS2 UFSARMPS2 UFSAR11.1-35Rev. 35TABLE 11.1-3 RADIOACTIVITY LEVELS OF SOLID WA STES (S EE NOTE) FILTER CARTRIDGES Chemical and Volume Control System Filter CartridgesClean Liquid Waste Filter CartridgesAerated Liquid Waste Filte r CartridgesSpent Fuel Pool Clean-up Filter CartridgesReactor Vessel Head Decontamination System Filter CartridgesIsotope (Curies/Cartridges) (Curies/Cartridge)(Curies/Cartridges)(Curies/Cartridge)(Curies/Cartridges)Normal MaximumNormal MaximumNormal Maximu mNormal MaximumNormalMaximumCr-51 0.209 13.05 0.001 0.043 --0.4520.003 0.185 0.0390.039Mn-54 0.9350.67 0.002 0.002 0.0040.0240.004 0.003 0.0280.028Mn-56 --0.486 --0.002 --0.018-- -- -- --Co-58 64.0 107.0 0.123 0.347 0.1923.7020.50 0.84 4.784.78 Fe-59 0.1920.069 0.001 0.001 --0.0020.002 0.001 0.0220.022Co-60 0.025 61.0 --0.166 --1.7660.068 0.142 0.5330.533Br-84Rb-88 Rb-89Sr-89Sr-90 Y-90 Sr-91 Y-91 Zr-95 -- -- -- -- -- -- -- -- -- --Mo-99 Ru-103 Ru-106 MPS2 UFSARMPS2 UFSAR11.1-36Rev. 35Note:The activity data in this Table are based on analysis done for the orig inal licensing of Millstone Unit 2 and are not deri ved from the updated radionuclide activities s hown in the Appendices to FSAR Chapter 11.I-129Te-129 I-131 I-132 Te-132I-133Cs-134 I-134 Te-134 I-135 Cs-136 Cs-137 Cs-138 Ba-140 La-140Pr-143 Ce-144TABLE 11.1-3 RADIOACTIVITY LEVELS OF SOLID WASTES (SEE NOTE)
FILTER CARTRIDGES (CONTINUED)Chemical and Volume Control System Filter CartridgesClean Liquid Waste Filter CartridgesAerated Liquid Waste Filter CartridgesSpent Fuel Pool Clean-up Filter CartridgesReactor Vessel Head Decontamination System Filter CartridgesIsotope(Curies/Cartridges) (Curies/Cartridge)(Curies/Cartridges)(Curies/Cartridge)(Curies/Cartridges)NormalMaximumNormalMaximumNormal MaximumNormal MaximumNormalMaximum MPS2 UFSAR11.1-37Rev. 35Note:The activity data in this Table are ba sed on analysis done for the original licensing of Millstone Unit 2 and are not derived from the updated radionuclide activities shown in the Appendices to FSAR Chapter 11. The best source of Solid Waste data can be found in the Annual Radiol ogical Ef fluent Release Report.TABLE 11.1-4 CURIE INVENTORY OF SOLID WASTE SHIPPED FROM MILLSTONE UNIT 2 (SEE NOTE) SourceNormal Curies per YearMaximum Curies per YearResins 4,215 42,871 Filter Cartridges 35 149 ConcentratesLess than 1 5TOTAL 4,250 43,025 MPS2 UFSAR11.1-38Rev. 35TABLE 11.1-5 ASSUMPTIONS FOR WASTE GAS DECAY TANK ACCIDENT Assumption (1) Maximum Noble Gas Activity in Waste Gas Decay Tank.
Basis: Maximum activity in tanks based on 1% degraded fuel, core power = 2700 MWt and degassing one system volume. Assumption (2) 2 Hour Ground Level Release.
Basis: Regulatory Guide 1.24 Assumption (3)
Ground Level X/Q:EAB hr. = 3.66 E -04 (sec/m
- 3) LPZ 4.80 E -05 (sec/m
- 3) Basis: 95% maximum X/Q' s during the years 1974 -1981.
IsotopeActivity, curiesKr-85m 278Kr-85 942Kr-87 160Kr-88 494 Xe-131m 486 Xe-133m 708 Xe-133 45,600 Xe-135m 26.2 Xe-135 1,280 Xe-137 28.9 Xe-138 101 MPS2 UFSAR11.1-39Rev. 35FIGURE 11.1-1 P&ID CLEAN LIQUID RADWASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-40Rev. 35FIGURE 11.1-2 P&ID CLEAN LIQUID RADWASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-41Rev. 35FIGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 1)
The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-42Rev. 35FIGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 2)
The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-43Rev. 35FIGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 3)
The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-44Rev. 35FIGURE 11.1-4 P&ID AERATED LIQUID RADWASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-45Rev. 35FIGURE 11.1-5 P&ID DIAGRAM GASEOUS RADWASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.1-46Rev. 35FIGURE 11.1-6 DEGASIFIER PERFORMANCE CURVE MPS2 UFSAR11.1-47Rev. 35FIGURE 11.1-7 P&ID SPENT RESIN RADWASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS2 UFSAR11.2-1Rev. 3511.2 RADIATION PROTECTION11.2.1 FINAL SAFETY ANALYS IS REPORT UP DATE It should be noted that the info rmation in this section provided the basis and description of the shielding design prior to operation. As the station has operated, dose rates ha ve fluctuated within the various areas as a func tion of time but for the most part have remained well within the zone designation limits presented below. Minor changes in equipment layout from that indicated in Figures 11.2-1 through 11.2-10 have been made, but since the va lues and figures given in this section provided the design basis for the major sh ielding structures, they are not being changed.
Many minor shielding changes (m ostly through the use of porta ble shielding) have been incorporated in the interest of maintaini ng occupational exposures ALARA. The only major shielding change was the in stallation of a reactor cavity neutron shield.
The values and figures given in th is section should not be used to estimate ar ea dose rates. Actual health physics surveys are availabl e and should be consulted to get a more accurate and up-to-date indication of radiological conditions.
Appendices 11.D and 11.E provide some of the design basis considerations on the control and expected levels of airborne radioactivity in the plant. Ag ain, health physics survey data should be consulted for more realistic indi cations of actual air borne radioactivity le vels. The occupational exposure due to airborne radioact ivity has been insignificant in co mparison to the direct radiation exposure.
11.2.2 DESIGN BASES The shielding is designed to perform two primary functions:
to ensure that during normal operation the radiation dose to operating personnel and to the general public is within the radiation exposure li mits set forth in 10 CFR Part 20 and to ensure th at operating personnel are adequately protected in the ev ent of a reactor incident so that the incident can be terminated without undue hazard to the general public. The shielding design is based on operating at design power level of 2700 MWt with reactor coolant system activity levels corresponding to one percent failed fuel. The shielding design is governed by the limits for radiation levels as follows: Zone DescriptionMaximum Dose Rate (mrem/hr)Uncontrolled, unlimited access (Zone A) 0.5 Controlled, unlimited (40 hr/week) (Zone B) 1.0Controlled, limited access (6-2/3 hr/week) (Zone C) 15 Controlled, limited access (1 hr/wk) (Zone D) 100 Normally inaccessible (Zone E)
> 100 MPS2 UFSAR11.2-2Rev. 35The maximum external dose rates within the sta tion are shown in the radiation zone diagrams, Figures 11.2-1 through 11.2-10 , as described in this subsection and later in Section 11.2.2. These maximum external dose rates are based on one percent fa iled fuel. Since the expected amount of failed fuel is much less than this, the average external dose rates will be much lower. The unrestricted areas in the plant where construction workers and pl ant visitors may be are all radiation Zone A areas which ha ve a maximum dose rate of 0.5 mrem/hr. Only plant workers are allowed in radiation zones with dose rates above this level, and their stay times are increasingly limited as the dose rate increases.
The anticipated average dose rate s will vary within ea ch zone and with operating conditions. In general, the dose rate normally around process equipm ent is expected to be at least an order of magnitude lower than indicated by the applicable zone. This is due to the extremely conservative assumptions made concerning the quantity of failed fuel. The dir ect dose at the site boundary is negligible both in the case of normal opera tion and under postulated incident conditions.
Shielding wall thicknesse s, as shown in Figures 11.2-1 through 11.2-10 , were found using the theory of the Reactor Shield ing Design Manual by T. Rockwell, Reactor Physics Constants ANL-5800, Table of Isotopes by Perlman and Lederer, Engineering Compendium on Radiation Shielding and Nuclear Engineer ing Handbook by H. Etherington. Th e geometric considerations were determined by each physical situation, such as source configurations and distances to dose points.Assumptions used included those of TID 14844 and the following:1.If the source was liquid, then self-attenuation by water was used.
If the s ource was gaseous, then no self-attenuation was assume
- d. If the source was solid, the volume percentages of each solid were used for self-attenuation (for example, solidified waste in drums).2.Shields included the tank or pipe walls which were assumed to have an attenuation coefficient similar to that for iron.3.The shielding materials were lead, iron, water , and concrete.4.Sources were given in Mev/cm 3 sec for each of 7 standard energy bins: 0.4, 0.8, 1.3, 1.7, 2.2, 2.5, > 2.5 Mev, and 6.1 and 7.1 Mev for nitrogen 16. One percent failed fuel was assumed.5.Flux to dose conversion factors us ed were those given by Henderson in XDC-59-8-179.6.The shield thicknesses were determined by the amount of shielding required to achieve a dose rate less than the upper limit for the radiation zone in which the dose point was located.
MPS2 UFSAR11.2-3Rev. 35 The shielding design, therefore, is based on cons ervative assumptions ensu ring adequate radiation protection to the general public a nd to operating personnel as well.11.
2.3 DESCRIPTION
Shielding throughout Millstone Un it 2 is designed in accordance wi th the criteria specified in Section 11.2.1. Figures 11.2-1 through 11.2-10 show the equipment plac ement, the quantity of shielding used and the appropriate radiation zone
- s. All components contai ning radioactive fluids are shown on the figures except the refueling water storage tank (RWST), located as shown on Figure 1.2-2. The maximum direct dose rate from th e content of the RWST is 5.6 x 10
-3 mr/hr, based on reactor operation with on e percent failed fuel. The ma ximum total site boundary dose due to the RWST is approximately 1.66 mrem/yr. For normal operating conditions, the site boundary doses will be approximately one-tenth of those values given above.11.2.3.1 Containment ShieldingThe containment shielding consisting of the primary shielding, the reactor cavity neutron shielding, the secondary shielding, and the containment wall, is shown on Figure 1.2-6.Primary shielding is provided to limit radiation emanating from the react or vessel. The primary shielding is designed to:a.Attenuate the neutron flux to limit the activation of components and structural steel.b.Limit the radiation level after shutdown to permit access to the reactor coolant system equipment.c.To reduce, in conjunction with the sec ondary shield, the radiation level from sources within the reactor vessel to allow limited access to th e containment during normal operation.
The primary shield consists of a minimum of five feet of re inforced concrete surrounding the reactor vessel. The cav ity betwee n the primary shield and the reactor vessel insulation is air-cooled to prevent overheating and dehydration of the concrete primary shield wall.A reactor cavity neutron shield reduces the ope rating neutron and gamma dose rates in the containment. A dose reduction factor of 40 on th e operating floor of th e containment was the design goal for the shield. With the postulation of leak-before-break applied to the reactor coolant piping, the neutron shield design does not consider dynamic effects resulting from a design basis LOCA.The permanent reactor cavity neutron shield consis ts of borated concrete blocks contained inside stainless steel containers. The concrete blocks are supported by the shie lding support bars, which span between the reactor vessel flange and th e embedment ring. Th e borated concrete blocks are 15 inches high and 23.25 inches wide, located al l around the reactor vessel beneath the permanent MPS2 UFSAR11.2-4Rev. 35reactor cavity seal. The bottom and the sides of the shieldi ng are ins ulated by NUKON blanket insulation to maintain an acceptable shielding temperature as well as limit heat loss from the reactor vessel (Figures 11.2-11 through 11.2-12). Thermal analysis has shown that the borated concrete blocks inside the stainless steel c ontainers could reach a temperature of up to 455
°F in some sections of the concrete blocks during normal operation.
The shielding is provided with personnel access openings, located direc tly above the neutron detector wells. During reactor operation these openings are plugged with borated concrete shield plugs to further minimize neut ron flux. The shield plugs are re moved through the access openings in the permanent cavity seal for maintenan ce of the neutron dete ctor instrumentation.The shielding in combination with th e reactor cavity se al described in Section 4.3.10 will provide adequate airflow channels to maintain suff icient cavity cooling during reactor operation.
In addition to the borated c oncrete shielding, borated polye thylene layers are installed approximately at the reactor vessel flange elev ation to reduce streaming between the borated concrete shielding and the primary shield wall. The thickness of the polyethylene shielding ranges from 1 to 3 inches contained in stainless steel liner between the ribs of the support members. The borated polyethylene layers are not insulated. This shielding is cooled by cavi ty cooling air passing under the shielding at temperatures le ss than 150
°F. Together, the borat ed concrete blocks and the borated polyethylene layers are capable of reducing the ne utron radiation by a dose factor of approximately 16 to 27.
Secondary shielding is provided to reduce the activity from the reactor coolant system to radiation levels which allow limited access to the cont ainment during normal opera tion and to supplement primary shielding. Nitrogen 16 is the major source of radioactivity in the reactor coolant during operation and controls the thickne ss of the secondary shie ld. The secondary shie lding consists of a minimum of 3.5 feet of reinfor ced concrete surrounding the reactor coolant piping, pumps, steam generators and pressurizer. The N-16 activity concentration at the reactor vessel outlet nozzles is 9.8 x 10-5 Ci/cc at full power.
The containment is a reinforced prestressed concrete structure wi th 3.75 feet thick cylindrical walls and a 3.25 feet thickness dome. In conjunction with the primary and secondary shield, it will limit the radiation level outside the enclosure building due to sources inside the structure to no more than 0.5 mrem/hr at full power operation. Th e structure is also designed to protect plant personnel from radiation sources inside the structure following a postulated incident.11.2.3.2 Auxiliary Building Shielding The function of the auxiliary bui lding shielding is to protect personnel working near various system components, such as those in the CVCS, the radioa ctive waste processing system, sampling system, and the spent fu el pool cooling system. Contro lled access to the auxiliary building is allowed during reac tor operation. Each equipment compartment is individually shielded to reduce the radiation level in it and adjacent compartments as reflected by the zone designations. Source terms used in the design of shielding for major components throughout the MPS2 UFSAR11.2-5Rev. 35 auxiliary building are listed in Table 11.2-1. Similar source terms fo r all components containing radioactive materials were develope d and used in the shielding design.11.2.3.3 Control Room Shielding The layout of the control room is shown in Figure 1.2-7. In conformance with General Criterion Number 19 of 10 CFR Part 50, the control room shielding is designed to ensure that the dose will not exceed 5 rem for the durat ion of the incident (see Subsection 14.18.3.3
). The walls of the control room are two feet thick. Under normal ope rating conditions, the cont rol room is a Zone A region with expected dose rates we ll below the indicated limits
.11.2.3.4 Spent Fuel Pool Shielding and Fuel Handling Shielding Fuel handling shielding is designed to facilitate the removal and transfer of spent fuel assemblies from the reactor vessel to the spent fuel pool.
it is designed to prot ect personnel against the radiation emitted from the spent fuel and control rod assemblies.
The refueling cavity above the reactor vessel is fl ooded to Elevation 36 feet 6 inches to provide a temporary water shield above the components being withdrawn from the reactor vessel.
The water height is approximately 24 feet above the reactor vessel flange. This height assures a minimum of 108 inches of water above the active portion of a wi thdrawn fuel assembly at its highest point of travel. Under thes e conditions, the dose rate from th e spent fuel assembly is less than 1.0 mrem/hr at the water surface.
Upon removal of the fuel assembly from the reactor vessel, it is moved to the spent fuel pool by the fuel transfer mechanism, vi a the fuel transfer tube. Conc rete shielding is provided around reactor internals storage and th e steam generator for personnel protection during refueling. The spent fuel pool in the auxiliary building is permanently flooded to provide a minimum of 108 inches of water abov e the active portion of a fuel assembly when being withdrawn from the fuel transfer tube and raised by the fuel pool platform crane, prior to in sertion in the spent fuel storage rack. The minimum water height above stored fuel a ssemblies is approximately 24.5 feet during operation of the spent fuel pool cooling system, described in Section 9.5.2.1 , to avoid air entrainment at the pump suction intakes. Otherwis e, a minimum height of 23 feet prev ails in the spent fuel pool and also, while fuel is in movement, above the reactor pressure vessel, to satisfy fission product retention assumptions for fu el handling accident calculations (see Section 14.7.4.2). The sides of the spent fuel pool are six feet thick concrete to ensure a dose rate of less than 0.035 mrem/hr on the outer surface of the spent fuel pool.11.2.3.5 Piping Systems Shielding All piping systems containing radioa ctive material are routed and/
or provided with shielding in accordance with the radi ation zones given in Section 11.2.1. Consideration is given to maintenance and inspection requirements for co mponents located in shielded compartments through which these lines are routed.
MPS2 UFSAR11.2-6Rev. 35 Isometrics of field run piping, two inches and sm aller in diameter , were reviewed and approved by Bechtel Corporation En gineering Department.11.2.4 HEALTH PHYSICS PROGRAM Information regarding the heal th physics program or ganizatio n is presented in Section 12.5 of Millstone 3 Final Safety Analysis Report (Reference 11.2-1). That information is contained herein by reference.11.
2.5 REFERENCES
11.2-1Millstone Unit 3, Final Safety Analysis Report, Section 12.5 - Health Physics Program MPS2 UFSAR11.2-7Rev. 35TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGN IsotopeLetdown Pr e-Filter (Curies)Purification Ion Exchanger (C uries)Letdown Post-filter (Curies)Volume Control Tank Demineralizer Degasifier (Curie s)Br-84-1.23-840.6 4.24x10-3Kr-85m---34.0 0.0Kr-85---87.1 0.0Kr-87---18.4 0.0Kr-88---58.2 0.0Rb-88-38.20-4.63 0.23Rb-89-0.82-0.11 4.02x10-3Sr-89-339.80-0.01 5.24x10-4Sr-90-85.90-5.0x10-4 2.63x10-5Y-90-1.79-2.1x10-2 1.05x10-3Sr-91---6.7x10-3 3.32x10-4Y-91---5.22 0.26 Mo-99---40.86 2.06 Ru-103-222.0-8.4x10-3 4.24x10-4 Ru-106-59.4-4.9x10-4 2.49x105Te-129-1.54-5.0x10-2 2.52x10-3 I-129-2.4x10-2-1.44x10 7 7.29x10-9 I-131-4.26x10 4-8.01 4.04x10-1 Xe-131m---50.8 0.0Te-132-1.40x10 3-0.65 3.27x10-2 I-132-129.0-2.03 1.03x10-1 I-133-6.32x10 2-10.90 5.51x10-1 Xe-133---5012.3 0.0Te-134-0.92-4.7x10 2 2.39x10-3 MPS2 UFSAR11.2-8Rev. 35 I-134-27.55-1.13 5.72x10-2 Cs-134---8.19 4.13X10-1 I-135-933.4-5.04 2.54x10-1 Xe-135---172.4 0.0 Cs-136---0.28 1.45x10-2 Cs-137---39.78 2.0 Xe-138---8.16 0.0 Cs-138---1.24 3.27x10-2 Ba-140-104.2-1.23x10-2 6.23x10-4 La-140-13.1-1.19x10-2 6.00x10-4Pr-143---1.08x10-2 5.45x10-3 Ce-143---7.74x10-3 3.91x10-3Co-60 263.0-2.63 1.98x10-4 1.0x10-5Fe-59 0.45-4.5x10-3 3.83x10-6 2.09x10-7Co-58 690.0-6.90 1.42x10-3 7.18x10-6 Mn-56 3.27-3.27x10-2 4.14x10-3 2.09x10-5 Mn-54 3.69-3.69x10-2 4.95x10-6 2.50x10-7Cr-51 88.10-0.88 4.32x10-4 2.18x10-7Zr-95 1.54-0.15 3.42x10-6 1.73x10-7TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGN (CONTINUED)
IsotopeLetdown Pre-Filter (Curies)Purification Ion Exchanger (Curies)Letdown Post-filter (Curies)Volume Control Tank Demineralizer Degasifier (Curies)
MPS2 UFSAR11.2-9Rev. 35TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGNS IsotopeWaste Gas Surge TankWaste Gas Decay T ankAerated Waste Drain TankAerated Waste Demineralizer Aerated Waste Monitor TankBr-84--8.8x10-3 2.78x10-5 8.8x10-4Kr-85m 4.09x10 2 2.80x10 2---Kr-85 3.13.10 2 2.61x10 3---Kr-87 6.97x10 1 1.41x10 1---Kr-88 4.80x10 2 2.10x10 2---Rb-88--0.486 8.63x10-4 0.243Rb-89--1.2x10-2 1.85x10-5 6x10-3Sr-89--1.08x10-3 7.68x10-3 5.4x10-4Sr-90--5.47x10-5 1.94x10-3 2.74x10-5Y-90--2.20x10-4 4.05x10-5 2.2x10-4Sr-91--7.04x10-4-3.52x10-4Y-91--5.49x10-2-5.49x10-2 Mo-99--0.430-0.430 Ru-103--8.84x10-4 5.02x10-3 8.84x10-5 Ru-106--5.19x10-5 1.34x10-3 5.19x10-6Te-129--5.24x10-3 3.48x10-5 5.24x10-4 I-129--1.52x10-8 5.42x10-7 1.52x10-9 I-131--0.842 9.63x10-1 8.42x10-2 Xe-131m 1.80x10 3 1.37x10 4---Te-132--6.81x10-2 3.16x10-2 6.81x10-3 I-132--0.214 2.92x10-3 2.14x10-2 I-133--1.15 1.43x10-2 1.15x10-1 Xe-133 1.71x10 5 1.17x10 6---
MPS2 UFSAR11.2-10Rev. 35Te-134--4.98x10-3 2.08x10-5 4.98x10-4 I-134--0.119 6.23x10-4 1.19x10-2 Cs-134--0.861-0.430 I-135--0.5302.11x10-2 5.3x10-2 Xe-135 3.41x10 2 4.86x10 2---Cs-136--3.03x10-2-1.52x10-2 Cs-137--4.18-2.09 Xe-138 6.75 2.99x10-1---Cs-138--1.31x10-1-6.6x10-2 Ba-140--1.30x10-3 2.35x10-3 1.30x10-4 La-140--1.25x10-3 2.96x10-4 1.25x10-4Pr-143--1.14x10-3-1.14x10-4 Ce-143--8.14x10-4-8.14x10-5Co-60--2.08x10-4-2.08x10-5Fe-59--4.03x10-6-4.03x10-7Co-58--1.50x10-3-1.50x10-4 Mn-56--4.35x10-3-4.35x10-4 Mn-54--5.29x10-6-5.20x10-7Cr-51--4.54x10-4-4.54x10-3Zr-95--3.60x10-6-3.6x10-7TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGNS IsotopeWaste Gas Surge TankWaste Gas Decay TankAerated Waste Drain TankAerated Waste Demineralizer Aerated Waste Monitor Tank MPS2 UFSAR11.2-11Rev. 35TABLE 11.2-1 SOURCE TERM S FOR SHIELDING DESIGNS Isotope Primary Demineralizer Coolant Waste ReceiverSecondary DemineralizerCoolant Waste Monitor T anks Primary Drain T ankBr-84 4.87x10-1 0.106 4.87x10-2 5.30x10-3 2.65x10-1Kr-85m-----Kr-85-----Kr-87-----Kr-88-----Rb-88 4.87x10-1 5.84 4.87x10-2 2.92x10-2 1.50x10+1Rb-89 1.04x10-2 0.145 1.04x10-3 7.24x10-3 3.62x10-1Sr-89 4.34 1.30x10-2 4.34x10-2 6.48x10-4 3.24x10-2Sr-90 1.10 6.56x10-4 1.10x10-1 3.28x10-5 1.64x10-3Y-90 4.71x10-2 2.63x10-2 4.71x10-3 1.32x10-3 6.59x10-3Sr-91 2.28x10-2 8.45x10-3 2.28x10-3 4.22x10-42.11x10-2Y-91 2.50x10 2 6.59 2.50x10-1 3.29x10-1 1.65 Mo-99 9.47x10 1 5.16x10 1 9.47 2.58 1.29x10 1 Ru-103 2.83 1.06x10-2 2.83x10-1 5.30x10-4 2.65x10-2 Ru-106 7.59x10-1 6.22x10-4 7.59x10-23.11x10-4 1.56x10-3Te-129 1.96x10-2 6.29x10-2 1.96x10-3 3.15x10-3 1.57x10-1 I-129 3.07x10-4-3.04x10-5--I-131 5.43x10 2 1.01x10 1 5.43x10 1 5.05x10-1 2.53x10 1 Xe-131m-----Te-132 1.77x10 1 0.818 1.77 4.09x10-2 2.04 I-132 1.64 2.566 1.65x10 1 1.28x10-1 6.42 I-133 8.06x10 1 1.378x10 1 8.06 6.89x10-1 3.45x10 1 Xe-133-----Te-134 1.16x10-2 5.97x10-2 1.16x10-3 2.99x10-3 1.49x10-1 MPS2 UFSAR11.2-12Rev. 35 I-134 3.52x10-10 1.43 3.52x10-2 7.15x10-2 3.58 Cs-134 1.50x10 4 1.03x10 1 1.5x10 3 5.17x10-1 2.58x10 1 I-135 3.52x10-1 6.36 3.52x10-2 3.18x10-1 1.59x10 1 Xe-135-----Cs-136 3.06x10 1 3.63x10-1 3.06 1.82x10-2 9.08x10-1 Cs-137 8.38x10 4 5.0x10 1 8.38x10 3 2.50 1.25 2 Xe-138-----Cs-138 2.34x10-1 8.18x10-1 2.34x10-2 4.09x10-2 3.92 Ba-140 1.33 1.56x10-2 1.33x10-1 7.79x10-4 3.89x10-2 La-140 1.68x10-1 1.5x10-2 1.68x10-2 7.51x10-4 3.75x10-2Pr-143 1.24 1.24x10-1 1.24x10-1 6.81x10-3 3.41x10-2 Ce-143 1.09x10 1 9.77x10-2 1.09 4.88x10-3 2.44x10-2Co-60----6.25x10-3Fe-59----1.21x10-4Co-58----4.49x10-2 Mn-56----1.31x10-1Mr-54----1.56x10-4Cr-51----1.36x10-2Zr-95----1.08x10-4TABLE 11.2-1 SOURCE TERM S FOR SHIELDING DESIGNS Isotope Primary Demineralizer Coolant Waste ReceiverSecondary DemineralizerCoolant Waste Monitor Tanks Primary Drain Tank MPS2 UFSAR11.2-13Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-1 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 45 FEET 6 INCHES
MPS2 UFSAR11.2-14Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-2 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 29 FEET 6 INCHES
MPS2 UFSAR11.2-15Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-3 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 5 FEET 0 INCHES
MPS2 UFSAR11.2-16Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-4 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT BUILDING - ELEVATION 14 FEET 6 INCHES AND 38 FEET 6 INCHES MPS2 UFSAR11.2-17Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-5 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL A UXILIARY BUILDING - ELEVATION 14 FEET 6 INCHES AN D 25 FEET 6 INCHES MPS2 UFSAR11.2-18Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-6 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL TURBINE BUILDING - ELEVATION 14 FEET 6 INCHES MPS2 UFSAR11.2-19Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-7 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL A UXILIARY BUILDING - ELEVATION 36 FEET 6 INCHES MPS2 UFSAR11.2-20Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-8 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - SECTION A-A MPS2 UFSAR11.2-21Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-9 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - SECTION B-B MPS2 UFSAR11.2-22Rev. 35Withheld under 10 CFR 2.390 (d)(1) FIGURE 11.2-10 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY MPS-2 FSAR Rev. 22.6FIGURE 11.2-11 NEUT RON SHIELD SEGMENT Outer C y linder Hanging Beam Access Port Inner C y linder MPS-2 FSAR Rev. 22.6FIGURE 11.2-12 NEUTRON SHIELDING - SECTIONAL VIEW FLOW AREA THROUGH 4 ROUND ACCESS PORTS:
SUPPORT PLATE RIB POLY SHIELDING REACTOR VESSEL FLOW AREA CONCRETE S HIELD WAL L BORATED CONCRETE WALL LINER NUKON INSULATION NUKON INSULATION FLOW AREA REFLECTIVE INSULATION MPS2 UFSAR11.2-25Rev. 35FIGURE 11.2-13 NOT USED MPS2 UFSAR11.2-26Rev. 35FIGURE 11.2-14 NOT USED MPS-2 FSAR Rev. 22.6FIGURE 11.2-15 NEUTRON SHIELD - THERMA L LOADINGS (TYPICAL ANNULAR SECTION)
THERMAL ANALYSIS MODEL: TYPICAL SHIELDING SECTION A DIABATI C FILM COEFFICIENT BULK AIR TEMP = 140 F AIR GAP BORATED CONCRETE TYPE 304 SS PLATE NUKON INSULATION T = 600 F SURFACE TEMPERATURE RV FLANGE FILM COEF.
FILM COEFFICIENT ADIABATIC MPS-2 FSAR Rev. 22.6FIGURE 11.2-16 NEUTRON SHIELD - TH ERMAL LOADINGS (PLUG SHIELD SECTION)
THERMAL ANALYSIS MODEL: PLUG SHIELDING SECTION A DIABATI C RV FLANGE FILM COEFFICIENT BULK AIR TEMP = 140 F AIR GAP (TYP)
BORATED CONCRETE TYPE 304 SS PLATE T = 600 F SURFACE TEMPERATURE FILM COEFFICIENT FILM COEF.
ADIABATIC NUKON INSULATION AIR GAP (TYP) 1 1 MPS2 UFSAR11.A-1Rev. 3511.A SOURCE TERMS FOR RADI OACTIVE WASTE PROCESSING AND RELEASES TO THE ENVIRONMENT11.A.1 REACTOR COOLANT DESIGN BASIS RADIONUCLIDE ACTIVITIESTo calculate reactor coolant de sign-basis radionuclide activities, a calculation of reactor core source term (discussed in Section 11.A.1.1 below) was performed. Reactor coolant activities were calculated taking into account (a) leakage from the fuel pellets to the primary coolan t due to fuel pin cladding failure, (b) removal by radioactive decay , (c) removal by the CVCS purification system, (d) removal of primary co olant for boron reduction, and (e) discharge of primary coolant to the liquid waste processing sy stem. Included in the resultan t reactor coolant design-basis radionuclide inventory are activat ed corrosion products (crud) and tritium (discussed in Sections 11.A.1.2 and 11.A.1.3 , respectively). Table 11.A-1 provides a tabula tion of the design-basis reactor coolant radionuclide activities. Bases for the calcul ation of these activities are presented in Table 11.A-10.11.A.1.1 Development of Reactor Co re Radionuclid e ActivitiesFor conservatism, the model used to calculate reactor core radi onuclide activities employs an end-of-cycle equilibrium fuel cycle condition. Cycle characteristics ar e consistent with nominal two-year fuel cycles (with an 80% cap acity factor included). The ORIGEN-S computer program is used to model the buildup and decay of core fission product radionuclides.
ORIGEN-S calculates the following three cate gories of core radionuclide activities:1.Fission Products2.Light Elements3.Actinides The light elements are neutron activation products of fuel assembly metallic structural components. The ORIGEN-S model used in the calculation of Millstone Unit 2 reactor core radionuclide activities is predicated on a standa rd 14-by-14 fuel pin ar rangement. Actinides simulate the buildup, decay, and loss due to fissioning. Only the 239 Np actinide data are used.
The ORIGEN-S results are calculat ed in terms of gram-atoms per metric tonne of initial heavy metal. These results are converte d, as necessary , to acti vities (in Curies) in the core using the following equation:
Acore N AACF--------------
-M u1B1C1B1C2B1C3
++()M u2B2C1B2C2+()+{}=
MPS2 UFSAR11.A-2Rev. 35 where N A is in atoms per gram-atom, is in seconds
-1 , ACF is the activit y conversion factor (equal to 3.7 x 10 10 disintegrations/second per Curie), M u1 is equal to 19.208 metric tonnes of uranium in Batch 1, M u2 is equal to 13.72 metric tonnes of uranium in Batch 2, and where B1C1, B1C2, B1C3, B2C1, and B2C2 are batch/cycle-specific values (in gram-atoms per metric tonnes of uranium) calculated by ORIGEN-S.
The reactor core radionuclide ac tivities thus calculated and us ed as basic source terms are tabulated in Table 11.A-2. Also presented in Table 11.A-2 are the bases for the calculation of these reactor core radionuclide activities.11.A.1.2 Corrosion ProductsThe activity concentrations of activ ated corrosion products (crud) in the reactor coolant have been calculated based on the model in the ANSI/ANS standard 1 dealing with the radioactive source term for normal operation of light water reactors.
The starting point for this calculation was the reactor coolant radionuclide activit y concentrations specified in th e standard for a reference PWR with U-tube steam generators (such as those used at Millst one Unit 2). These activity concentrations were altered by adjustment factors that were prepared in accordance with the ANSI/ANS model to reflect the operating parameter differences betw een Millstone Unit 2 and the reference PWR, and were then further adjusted to the Techni cal Specificati on radionuclide concentration limits.11.A.1.3 Tritium ProductionTritium may be produced in the c oolant or enter the coolant from a number of sources. One source is from fissioning of uranium within the fuel, yi elding tritium as a tertia ry fission product. Since zircaloy fuel cladding reac ts with tritium to form zircaloy hydr ide, no tritium diffuses through the cladding 2 , 3. Therefore, the tritium released to the c oolant from the fuel originates only from defective fuel.
Tritium is also produced by the re action of neutrons with boron in the control element assemblies (CEAs). Data from ope rating plants using B 4C control rods indicates that no tritium is released from the control rods. The trit ium may combine with carbon to form hydrocarbons and/or with lithium to form lithium hydride, thereby preventing diffusi on through the inconel cladding.
Another possibility is that the low internal temperature of the B 4C control rods (relative to 1ANSI/ANS-18.1-1984, American National Standard - Radioactive Source Term for Normal Operation of Light Water Reactors, dated 12/31/1984 2James M. Smith, Jr., The Significance of Tritium in Water Reactors, GE, APED, 9/19/67 3Joseph W. Ray, et. al., Investigation of Tritium Generation and Release in PM Nuclear Power Plants, BMI-1787, 10/31/66 MPS2 UFSAR11.A-3Rev. 35 stainless clad fuel rods from which about 45%
escape can be expected) may prohibit tritium diffusion. To account for possible c ontrol rod cladding defects, it is assumed that one percent of the tritium produced in the CEAs is released to the coolant.
Another source of tritium is the activation of boron, lithium, deuterium, and nitrogen within the reactor coolant. Boro n in the form of boric acid is used in the coolant for reactivity control.
Lithium is produced in the coolant as a result of neutron-boron reaction and may also be added as a pH control agent. Deuterium is a naturally oc curring constituent of water. Nitrogen may be present due to aeration of the coolant during shutdown and due to aerated makeup water.The expected tritium releases from the combined liquid and vapor pathways was assumed to be 0.4 Ci/year per MWt, based on review of the tritium release rates at a number of PWRs evaluated in NUREG-0017, Rev. 1. The quantity of tritium released in the liquid pathway is based on the calculated volume of liqui d released, excluding se condary system wastes, with a primary coolant concentration of 1.0
µCi/ml. It was assumed that the remainder was released as a gas from building ventilation system exhaust.11.A.1.4 Fuel Experience Past operation of stainless steel-clad fuel rods in the Connecticut Yankee reactor showed fuel failure rates on the order of 0.01%.
Zircaloy-clad UO 2 fuel in the Obrigheim reactor in Germany sustained a fuel failure rate just over 0.1% in its first cycle, but this had fallen in the second cycle to e ssentially zero (0.001%). The fuel failure rate in the Dresden 1 re actor over a nine-year period had averaged < 0.1%, with the rate more recently being even lower. Fuel in the Mihama reactor in Japan and in the Point Beach reactor had exceeded the burn up at which failures in fuel of similar design were observed in Ginna, without exhibiting increases in coolan t activity (indicative of fuel defects).Fuel failure rates in the current generation of r eactors can be controlled to very low levels.
Widespread fuel defects in certain reactors have been observed, the cause having been attributed to fuel clad contamination. Appr opriate corrective actions have been devised to ensure that the occurrence of such fuel defects will be greatly minimized in th e future. Nevertheless, there is always the possibility, despite car eful testing and manufacture, th at other defects will become apparent in new fuel designs in the future th at, because of statisti cal considerations or unrecognized or uncontrollable environmental diff erences, could not be foreseen. The design refinements continuously introduced in nuclear power reactors and their fuel as a natural outcome of a dynamic industry will, on rare occasions, introduce such defects. Existing licensing regulations limit coolant activity to that associated with 1%
failed fuel, even during these transitory and infrequent periods.
The fuel failure rate attainable under more normal conditions has be en demonstrated to be nearer 0.01%. Over the lifetime of an operating reactor, the latter rate is exp ected to predominate.
MPS2 UFSAR11.A-4Rev. 3511.A.2 REACTOR COOLANT EXPECTED RADIONUCLIDE ACTIVITIES Expected reactor coolant radionuclide activitie s are based on data ge nerated from operating plants, field and laboratory test s, and plant-specific design cons iderations. These activities are built into the PWR-GALE Code (henceforth referred to as 'GALE'
), which is a computerized mathematical model for calculating the expected releases of radioactive material in liquid and gaseous effluents from pressu rized water reactors (PWRs). The expected reactor coolant radionuclide activities are tabulated in Table 11.A-1 , juxtaposed with th e design-basis reactor coolant radionuclide activities.11.A.3 CALCULATION OF LIQUID AND GASEOUS EFFLUENT RELEASES11.A.3.1 Expected Liquid and Gaseous Ra dioactive Effluent ReleasesThe Millstone Unit 2 expected li quid and gaseous radioactive ef fluent releases are calculated to determine compliance with 10 CFR Part 50 Appendix I. This calculation is performed using GALE. As previously stated, th e calculation is based on data generated from operating reactors, field and laboratory tests, and pl ant-specific design considerati ons incorporated to reduce the quantity of radioactive materi als that may be released to the environment during normal operation, including anticipated operational occurr ences. The calculation performed by GALE is based on (a) American Nuclear Society (ANS) 18.1 Working Group recommendations for adjustment factors, (b) the release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, (c) plant-sp ecific design features used to reduce the quantities of radioactive materials that are ultimately released to the environment, and (d) information received on the ope ration of nuclear power plants. The principal mechanisms that affect the concentrations of radioactive mate rials in the primary coolant are the following:*fission product leakage to the primary coolant fr om defects in the fuel cladding and fission product generation in tramp uranium,*corrosion products activated in the core (i.e., crud),*radioactivity removed in the reactor coolant treatment systems, and
- activity removed because of primary coolant leakage.
The descriptions of the Millstone Unit 2 liquid and gase ous waste processing sy stems, as well as the GALE input parameters unique to those syst ems, are presented in Appendix 11.B. The following tables provide tabu lations of the GALE results:
- Table 11.A-3: Expected Annual Airborne Effluent Releases (Curies per Year), by Radionuclide, from Each Release Point
- Table 11.A-4: Expected Annual Liquid Ac tivity Releases (Curie s/Year), by Radionuclide, from Each Waste Stream MPS2 UFSAR11.A-5Rev. 35
- Table 11.A-5: Expected Annual Liquid Effluent Conc entrations (Diluted and Undiluted), by Radionuclide, from Each Liquid Waste Stream
- Table 11.A-8: Total Annual Design Basis And Expected Releases Of Radioactive Liquid Waste To The Environment From All Sources Combined, In Curies Per Year
- Table 11.A-9: Total Annual Design Basis And Expected Releases Of Airborne Radioactive Waste To The Envi ronment From All Releas e Points Combined, In Curies Per Year11.A.3.2 Design Basis Liquid and Gaseous Radioactive Effluent Releases The Millstone Unit 2 design basis liquid and gaseous radioactive ef fluent releases are calculated to determine compliance with 10 CFR Part 20. To perform this ca lculation, the design basis primary coolant activity concentrations are used, in c onjunction with their corresponding expected primary coolant activity concentrations , to determine isotopic scaling factors. These scaling factors are then applied to the expected liquid and gaseous effluent release inventory, as calculated by GALE, in order to determine the design basis liquid and gaseous radioactive effluent releases.
The calculated design basis liquid and gaseous radioactive ef fluent releases are provided in the following tables:
- Table 11.A-6: Design Basis Radionuclide Concentrations in Liquid Ef fluent, in Fractions of 10 CFR Part 20 Concentration Limits
- Table 11.A-7: Design Basis Radionuclide Airborne C oncentrations at the Site Boundary from All Gaseous Effluent Release Po ints Combined, in Fractions of 10 CFR Part 20 Concentration Limits
- Table 11.A-8: Total Annual Design Basis And Expected Releases Of Radioactive Liquid Waste To The Environment From All Sources Combined, In Curies Per Year
- Table 11.A-9: Total Annual Design Basis And Expected Releases Of Airborne Radioactive Waste To The Envi ronment From All Releas e Points Combined, In Curies Per Year11.A.4 SOLID WASTE PROCESSING SYSTEM The information in this section is based on analyses done for the original licensing of Millstone Unit 2 and is not derived from the updated radionuclid e activities shown in the Appendices to FSAR Chapter 11.
MPS2 UFSAR11.A-6Rev. 3511.A.4.1 Spent Resins11.A.4.1.1 Spent Resins from CVCS Ion Exchanger The radioactivity buildup on the CVCS ion exchange rs is based on a conti nuous letdown flow of 40 gpm. One of the two purificati on ion exchangers and the debor ating ion exchanger will be sluiced to the solid waste processing system at the end of each core cycle. Each of the two purification ion exchangers operates for two cycles, one cycle as a Li removal exchanger and the next cycle as the continuous purification exchanger, such that one of the two exchangers will be on the Li removal cycle while the ot her is for continuous purification.
The decontamination factor for ion exchangers is assumed to be 10 for all soluble isotopes except Y, Mo, and Cs. For Y, Mo, and Cs, a removal factor of 10 with a 20% usage factor is used for the Li removal cycle for each purification ion exchanger.
The buildup of activity on the CVCS ion exchangers' resins (32 ft 3 each) is given in Table 11.A-3.11.A.4.1.2 Spent Resins from Clean Liquid Wast e Processing System Demineralizers Activity buildup on the clean li quid waste processing system de mineralizers is based on the processing of 14 system volumes pe r year of reactor coolant wast es. An average processing rate based on annual volumes of liquid wastes is assumed, with a decontamination factor of 10 3 for soluble isotopes. Operation is assumed to be divided equally between the two primary demineralizers. The loading on the secondary demine ralizer resin bed(s) will be much less than that on the primary de mineralizers and has not been tabulated.
The activity buildup on the clean liquid waste processi ng system deminerali zers' resins (42 ft 3 each) is given in Table 11.1-3.11.A.4.1.3 Spent Resins from Aerated Liquid Wa ste Processing System Demineralizer Activity buildup on the aerated li quid waste demineralizer resin is based on th e processing of the volumes of aerated liquid radw aste generated annually. An average processing rate based on annual volumes of liquid waste is assumed, with a total DF of 500 taken for soluble isotopes.
The resin activity buildup on the hard piped demineralizer resin (42 ft
- 3) is given in Table 11.1-3.The resin activity buildup on the po rtable demineralizers (15 ft 3 each) is not addr essed, since this activity is bounded by the resin activity buildup on hard-piped demineralizer.11.A.4.1.4 Spent Resins from Spent Fuel Pool Demineralizer The activity buildup on resins from the spent fu el demineralizer is ba sed on the processing of spent fuel pool water and refueli ng water with a radioactivity le vel corresponding to one-tenth of reactor coolant. The assumption is made that complete mixing of reactor coolant, spent fuel pool MPS2 UFSAR11.A-7Rev. 35water, and refueling water occurs during refueling operations, resul ting in a dilution of the reactor coolant by a factor of approximately 10. An expected DF of 1000 is used for the demineralizer.
The activity buildup on the demineralizer resin (42 ft
- 3) is given in Table 11.1-3.11.A.4.1.5 Contaminated Filter Cartridges Buildup of activity on filt er cartridges is based on processi ng of liquid radioactive waste. A decontamination factor (DF) of 10 is taken for each filter. This DF is consistent with operating experience from nuclear power stations utilizing three micr on filter cartridges.
The activity buildup on the fi lter cartridges is given in Table 11.1-3.
MPS2 UFSAR11.A-8Rev. 35TABLE 11.A-1 DESIGN-BASIS AND EXPECTED PRIMARY COOLANT ACTIVITY CONCENTRATIONS Noble Gases NUCLIDE DESIGN-BASIS PRIMARY COOLANT (µCi/gm)EXPECTED PRIMARY COOLANT (µCi/gm)Kr-85m 1.33E+00 1.40E-01Kr-85 4.50E+00 2.12E-02Kr-87 7.64E-01 1.41E-01Kr-88 2.36E+00 2.53E-01 Xe-131m 2.32E+00 1.17E-01 Xe-133m 3.39E+00 3.03E-02 Xe-133 2.17E+02 6.88E-01 Xe-135m 1.25E-01 1.25E-01Xe-35 6.13E+00 6.72E-01 Xe-137 1.38E-01 3.28E-02 Xe-138 4.81E-01 1.15E-01 Halogens NUCLIDE DESIGN-BASIS PRIMARY COOLANT (µCi/gm)EXPECTED PRIMARY COOLANT (µCi/gm)I-131 5.57E+00 4.51E-02 I-132 1.18E+00 2.04E-01 I-133 7.29E+00 1.39E-01 I-134 7.77E-01 3.30E-01 I-135 3.65E+00 2.55E-01Br-84 4.84E-02 1.55E-02 MPS2 UFSAR11.A-9Rev. 35Activated Corrosion Products (Crud)NUCLIDE DESIGN-BASIS PRIMARY COOLANT (µCi/gm)EXPECTED PRIMARY COOLANT (µCi/gm)Na-24 1.40E+00 4.63E-02Cr-51 9.50E-02 3.09E-03 Mn-54 4.90E-02 1.60E-03Fe-55 3.70E-02 1.20E-03Fe-59 9.20E-03 2.99E-04Co-58 1.40E-01 4.59E-03Co-60 1.60E-02 5.29E-04Zn-65 1.60E-02 5.09E-04W-187 7.60E-02 2.47E-03 Np-239 6.70E-02 2.19E-03 Other Particulate Radionuclides NUCLIDE DESIGN-BASIS PRIMARY COOLANT (µCi/gm)EXPECTED PRIMARY COOLANT (µCi/gm)Rb-88 5.04E-02 1.84E-01Sr-89 3.74E-03 1.40E-04Sr-90 3.83E-04 1.20E-05Sr-91 1.83E-03 9.42E-04Y-91m 1.43E-04 4.46E-04Y-91 4.83E-03 5.19E-06Y-93 1.53E-03 4.12E-03Zr-95 1.20E-02 3.89E-04Nb-95 6.39E-03 2.79E-04 Mo-998.11E-01 6.36E-03Tc-99m 1.77E-03 4.60E-03 Ru-103 5.57E-03 7.49E-03 Ru-106 1.94E-03 8.99E-02 MPS2 UFSAR11.A-10Rev. 35Ag-110m 1.28E-05 1.30E-03Te-129m 2.20E-02 1.90E-04Te-129 7.99E-03 2.33E-02Te-131m 4.75E-02 1.48E-03Te-131 8.51E-03 7.45E-03Te-132 4.46E-01 1.69E-03 Cs-134 1.92E+00 7.56E-03 Cs-136 5.10E-01 9.21E-04 Cs-137 1.58E+00 1.00E-02 Ba-140 6.49E-03 1.30E-02 La-140 5.14E-03 2.48E-02 Ce-141 6.16E-03 1.50E-04 Ce-143 4.02E-03 2.77E-03 Ce-144 4.93E-03 3.89E-03 Other Particulate Radionuclides NUCLIDE DESIGN-BASIS PRIMARY COOLANT (µCi/gm)EXPECTED PRIMARY COOLANT (µCi/gm)
MPS2 UFSAR11.A-11Rev. 35TABLE 11.A-2 CALCULATED REACTOR CORE ACTIVITIES Nobel GasesNuclideReactor Core Activities (Curies) 1Kr-85m 1.859E+07Kr-85 8.557E+05Kr-87 3.736E+07Kr-88 5.193E+07 Xe-131m 9.452E+05 Xe-133m 4.636E+06 Xe-133 1.482E+08 Xe-135m 3.021E+07 Xe-135 4.288E+07 Xe-137 1.344E+08 Xe-138 1.266E+08 Halogens NuclideReactor Core Activities (Curies) 1Br-84 1.658E+07 I-131 7.120E+07 I-132 1.036E+08 I-133 1.479E+08 I-134 1.644E+08 I-135 1.400E+08 MPS2 UFSAR11.A-12Rev. 35 Cesium and Rubidium NuclideReactor Core Activities (Curies) 1Rb-88 5.320E+07 Cs-134 1.202E+07 Cs-136 3.475E+06 Cs-137 9.888E+06 Other Nuclides NuclideReactor Core Activities (Curies) 1 Cr-51 2.413E+06 Mn-54 9.754E+04 Fe-55 4.496E+05 Fe-59 3.382E+04 Co-58 8.739E+05 Co-60 8.888E+05 Zn-65 6.496E-02 Np-239 1.347E+09 Sr-89 7.321E+07 Sr-90 7.406E+06 Sr-91 9.077E+07Y-91 9.435E+07Y-91m 5.255E+07 Y-93 7.297E+07 Zr-95 1.250E+08 Nb-95 1.255E+08 Mo-99 1.343E+08Tc-99m 1.188E+08 Ru-103 1.093E+08 Ru-106 3.752E+07 MPS2 UFSAR11.A-13Rev. 351. The reactor core radionuclide activities tabulated above are based on the following: (a) a core power level of 2,700 MWt and (b) a three-re gion equilibrium cycle core, with an end-of-cycle core average burn-u p of 36,142 MWD/MTU, the three regions having operated at a specific power of 31.74 MWt/MTU for 705, 1333 and 1550 EFPD, respectively
.Ag-110m 2.479E+05Te-129 2.136E+07Te-129m 4.324E+06Te-131 6.042E+07Te-131m 1.373E+07Te-132 1.025E+08 Ba-140 1.314E+08 La-140 1.359E+08 Ce-141 1.213E+08 Ce-143 1.127E+08 Ce-144 9.545E+07 Other Nuclides NuclideReactor Core Activities (Curies) 1 MPS2 UFSAR11.A-14Rev. 35TABLE 11.A-3 EXPECTED ANNUAL EFFLUENT RELEASES (CURIES PER YEAR), BY RADIONUCLIDE, FROM EACH RELEASE POINT Release Points: Iodines Turbine Building (Ci/yr)Unit 2 Vent (Ci/yr)Millstone Stack (Ci/yr)I-131 0.00E+00 1.77E-01 1.00E-03I-133 1.80E-04 5.51E-01 1.00E-03Totals 1.80E-04 7.28E-01 2.00E-03Release Points: Noble Gases Turbine Building (Ci/yr)Unit 2 Vent (Ci/yr)Millstone Stack (Ci/yr)Kr-85m 0.00E+00 3.00E+00 1.00E+00K-r85 0.00E+00 3.80E+01 5.60E+02Kr-87 0M.00E+00 3.00E+00 1.00E+00Kr-88 0.00E+00 5.00E+00 4.00E+00 Xe-131m 0.00E+00 2.70E+01 2.30E+01 Xe-133m 0.00E+00 1.00E+00 2.00E+00 Xe-133 0.00E+00 8.10E+01 7.10E+01 Xe-135m 0.00E+00 3.00E+00 1.00E+00 Xe-135 0.00E+00 1.90E+01 1.20E+01 Xe-137 0.00E+00 0.00E+00 0.00E+00 Xe-138 0.00E+00 2.00E+00 1.00E+00Totals 0.00E+00 1.82E+02 6.76E+02 MPS2 UFSAR11.A-15Rev. 35Release Points: Others (Totals Allocated to Unit 2 Vent Release for Conservatism)Turbine Building (Ci/yr)Unit 2 Vent (Ci/yr)Millstone Stack (Ci/yr)H-3none1.10E+02noneC-14 none7.30E+00none Ar-41none3.40E+01noneTotalsnone1.51E+02none MPS2 UFSAR11.A-16Rev. 35Release Points: Particulates Turbine Building (Ci/yr)Unit 2 Vent (Ci/yr)Millstone Stack (Ci/yr)Cr-51none9.70E-051.40E-05Mn-54none5.68E-052.10E-06 Co-57none8.20E-060.00E+00Co-58none4.79E-048.70E-06Co-60none1.13E-041.40E-05 Fe-59none2.75E-051.80E-06Sr-89none1.59E-044.40E-05Sr-90none6.29E-051.70E-05 Zr-95none1.00E-054.80E-06Nb-95none4.23E-053.70E-06Ru-103none1.66E-053.20E-06 Ru-106none7.50E-072.70E-06Sb-125none6.09E-070.00E+00Cs-134none4.74E-053.30E-05 Cs-136none3.25E-055.30E-06Cs-137none8.92E-057.70E-05Ba-140none4.00E-062.30E-05 Ce-141none1.33E-05 2.20E-06 Totalsnone1.26E-032.57E-04 MPS2 UFSAR11.A-17Rev. 35TABLE 11.A-4 EXPECTED ANNUAL LIQUID EFFLUENT ACTIVITY RELEASES (CURIES/YEAR), BY RADIONUCLIDE, FROM EACH WASTE STREAMNUCLIDE BORON RS (Ci/yr)MISC. WASTES (Ci/yr)SECONDARY (C i/yr)TURB. BLDG. (Ci/yr)DETERGENT (Ci/y r)Corrosion and Activation Products
- Na-24 2.46E-04 9.26E-03 4.62E-01 5.94E-05 0.00E+00Cr-51 3.39E-04 1.96E-03 3.81E-02 6.23E-06 4.70E-04 Mn-542.11E-04 1.04E-03 1.94E-02 3.13E-06 3.80E-04Fe-55 1.61E-04 7.84E-04 1.47E-02 2.36E-06 7.20E-04Fe-59 3.55E-05 1.92E-04 3.54E-03 5.76E-07 2.20E-04Co-58 5.71E-04 2.97E-03 5.63E-02 9.13E-06 7.90E-04Co-60 7.12E-05 3.47E-04 6.59E-03 1.06E-06 1.40E-03Zn-65 6.71E-05 3.32E-04 6.27E-03 1.01E-06 0.00E+00W-187 2.36E-05 7.16E-04 2.61E-02 3.72E-06 0.00E+00 Np-239 5.48E-05 9.81E-04 2.45E-02 3.85E-06 0.00E+00 Fission Products
- Br-84 4.64E-10 1.90E-05 3.21E-02 4.26E-09 0.00E+00Rb-88 3.25E-10 1.44E-03 2.42E-013.36E-11 0.00E+00Sr-89 1.69E-05 8.99E-05 1.68E-03 2.74E-07 8.80E-06Sr-90 1.62E-06 7.85E-06 1.47E-04 2.36E-08 1.30E-06Sr-91 2.71E-06 1.23E-04 8.95E-03 9.88E-07 0.00E+00Y-91m 1.75E-06 7.86E-05 1.37E-03 6.33E-07 0.00E+00Y-91 1.45E-06 6.86E-06 6.47E-05 1.35E-08 8.40E-06Y-93 1.29E-05 5.68E-04 3.82E-02 4.31E-06 0.00E+00Zr-95 4.80E-05 2.51E-04 4.74E-03 7.68E-07 1.10E-04Nb-95 3.95E-05 1.85E-04 3.32E-03 5.31E-07 1.90E-04 Mo-99 1.90E-04 3.02E-03 7.26E-02 1.15E-05 6.00E-06Tc-99m 1.80E-04 3.02E-03 3.73E-02 8.30E-06 0.00E+00 Ru-103 8.72E-04 4.79E-03 9.12E-02 1.49E-05 2.90E-05 Ru-106 1.19E-02 5.87E-021.11E+00 1.78E-04 8.90E-04Ag-110m 1.71E-04 8.47E-04 1.58E-02 2.55E-06 1.20E-04Te-29m 2.16E-05 1.21E-04 2.29E-03 3.74E-07 0.00E+00 MPS2 UFSAR11.A-18Rev. 35Te-29 1.40E-05 2.47E-04 9.12E-02 6.49E-07 0.00E+00Te-131m 1.85E-05 4.97E-04 1.60E-02 2.36E-06 0.00E+00Te-131 3.38E-06 9.50E-05 1.31E-02 4.32E-07 0.00E+00I-131 2.91E-03 2.64E-02 4.51E-01 1.43E-04 1.60E-04Te-132 5.87E-05 8.39E-04 1.91E-02 3.06E-06 0.00E+00I-132 8.64E-05 5.45E-03 1.07E+00 6.24E-05 0.00E+00I-133 1.04E-03 3.66E-02 1.24E+00 3.42E-04 0.00E+00I-134 5.82E-07 1.41E-03 9.69E-01 2.92E-06 0.00E+00 Cs-134 1.14E-01 2.47E-01 9.36E-02 1.54E-05 1.10E-03I-135 3.76E-04 2.22E-02 1.91E+00 3.48E-04 0.00E+00 Cs-136 9.16E-03 2.81E-021.11E-02 1.84E-06 3.70E-05 Cs-137 1.52E-01 3.28E-01 1.25E-01 2.05E-05 1.60E-03 Ba-140 1.14E-03 7.90E-03 1.51E-01 2.48E-05 9.10E-05 La-140 1.48E-03 1.29E-02 2.76E-01 4.42E-05 0.00E+00 Ce-141 1.69E-05 9.53E-05 1.79E-03 2.92E-07 2.30E-05 Ce-143 3.86E-05 9.81E-04 2.96E-02 4.42E-06 0.00E+00 Ce-144 5.15E-04 2.54E-03 4.78E-02 7.70E-06 3.90E-04 All Others Except Tritium 1.56E-01 3.73E-01 6.69E-02 2.20E-04 2.31E-04 TOTAL 4.54E-01 1.19E+00 8.90E+00 1.56E-03 8.98E-03TABLE 11.A-4 EXPECTED ANNUAL LIQUID EFFLUENT ACTIVITY RELEASES (CURIES/YEAR), BY RADIONUCLIDE, FROM EACH WASTE STREAMNUCLIDE BORON RS (Ci/yr)MISC. WASTES (Ci/yr)SECONDARY (Ci/yr)TURB. BLDG. (Ci/yr)DETERGENT (Ci/yr)
MPS2 UFSARMPS2 UFSAR19Rev. 35TABLE 11.A-5 EXPECTED ANNUAL LIQUID EFFLUENT CONCENTRATIONS (DILUTED AND UNDILUTED), BY RADIONUCLIDE, FROM EACH WASTE SYSTEMNUCLIDE BORON RS, UND ILUTED (µCi/ml)MISC. WASTES, UNDILUTED
(µCi/ml)SECONDARY, UNDILUTE D (µCi/ml)TURB. BLDG., UNDILUTED
(µCi/ml)DETERGENT, UNDILU TED (µCi/ml)BORON RS, DILUTED (µCi/ml)MISC. W ASTES, DILUTED (µCi/ml)SECONDARY , DILUTED
(µCi/ml)TURB. BLDG., DILUTED (µCi/ml)DETERGENT, DILUTED (µCi/ml)Corrosion and Activation Products
- Na-24 5.45E-086.04E-061.57E-06 5.97E-090.00E+002.64E-139.94E-124.96E-10 6.38E-140.00E+00Cr-51 7.50E-081.28E-061.30E-07 6.26E-106.30E-073.64E-132.10E-124.09E-11 6.69E-155.05E-13 Mn-54 4.67E-086.78E-076.60E-08 3.15E-105.09E-072.27E-131.12E-122.08E-11 3.36E-154.08E-13Fe-55 3.56E-085.11E-075.00E-08 2.37E-109.65E-071.73E-138.42E-131.58E-11 2.53E-157.73E-13Fe-59 7.86E-091.25E-071.20E-085.79E-112.95E-073.81E-142.06E-133.80E-12 6.18E-162.36E-13 Co-58 1.26E-071.94E-061.91E-07 9.18E-101.06E-066.13E-133.19E-126.04E-11 9.80E-158.48E-13 Co-60 1.58E-082.26E-072.24E-08 1.07E-101.88E-067.65E-143.73E-137.07E-12 1.14E-151.50E-12 Zn-65 1.49E-082.16E-072.13E-08 1.02E-100.00E+007.20E-143.56E-136.73E-12 1.08E-150.00E+00W-187 5.22E-094.67E-078.88E-08 3.74E-100.00E+002.53E-147.69E-132.80E-11 3.99E-150.00E+00Np-239 1.21E-086.40E-078.33E-08 3.87E-100.00E+005.88E-141.05E-122.63E-11 4.13E-150.00E+00Fission Products
- Br-84 1.03E-131.24E-081.09E-07 4.28E-130.00E+004.98E-192.04E-143.45E-11 4.57E-180.00E+00 Rb-88 7.19E-149.39E-078.23E-07 3.38E-150.00E+003.49E-191.55E-122.60E-10 3.61E-200.00E+00Sr-89 3.74E-095.86E-085.71E-092.75E-111.18E-081.81E-149.65E-141.80E-12 2.94E-169.45E-15Sr-90 3.59E-105.12E-095.00E-10 2.37E-121.74E-091.74E-158.43E-151.58E-13 2.53E-171.40E-15Sr-91 6.00E-108.02E-083.04E-089.93E-110.00E+002.91E-151.32E-139.61E-12 1.06E-150.00E+00Y-91m 3.87E-105.13E-084.66E-096.36E-110.00E+001.88E-158.44E-141.47E-12 6.80E-160.00E+00Y-91 3.21E-104.47E-092.20E-10 1.36E-121.13E-081.56E-157.37E-156.95E-14 1.45E-179.02E-15Y-93 2.86E-093.70E-071.30E-07 4.33E-100.00E+001.39E-146.10E-134.10E-11 4.63E-150.00E+00Zr-95 1.06E-081.64E-071.61E-087.72E-111.47E-075.15E-142.70E-135.09E-12 8.25E-161.18E-13Nb-95 8.74E-091.21E-071.13E-085.34E-112.55E-074.24E-141.99E-133.56E-12 5.70E-162.04E-13 MPS2 UFSARMPS2 UFSAR20Rev. 35 Mo-99 4.21E-081.97E-062.47E-07 1.16E-098.04E-092.04E-133.24E-127.79E-11 1.23E-146.44E-15Tc-99m 3.98E-081.97E-061.27E-07 8.34E-100.00E+001.93E-133.24E-124.00E-11 8.91E-150.00E+00 Ru-103 1.93E-073.12E-063.10E-07 1.50E-093.89E-089.36E-135.14E-129.79E-11 1.60E-143.11E-14 Ru-106 2.63E-063.83E-053.77E-06 1.79E-081.19E-061.28E-116.30E-111.19E-09 1.91E-139.56E-13Ag-110m 3.79E-085.52E-075.37E-08 2.56E-101.61E-071.84E-139.09E-131.70E-11 2.74E-151.29E-13Te-129m 4.78E-097.89E-087.79E-093.76E-110.00E+002.32E-141.30E-132.46E-12 4.02E-160.00E+00Te-129 3.10E-091.61E-073.10E-076.52E-110.00E+001.50E-142.65E-139.79E-11 6.97E-160.00E+00Te-131m 4.10E-093.24E-075.44E-08 2.37E-100.00E+001.99E-145.34E-131.72E-11 2.53E-150.00E+00Te-131 7.48E-106.20E-084.45E-084.34E-110.00E+003.63E-151.02E-131.41E-11 4.64E-160.00E+00I-131 6.44E-071.72E-051.53E-06 1.44E-082.14E-073.12E-122.83E-114.84E-10 1.54E-131.72E-13Te-132 1.30E-085.47E-076.50E-08 3.08E-100.00E+006.30E-149.01E-132.05E-11 3.29E-150.00E+00I-132 1.91E-083.55E-063.64E-06 6.27E-090.00E+009.28E-145.85E-121.15E-09 6.70E-140.00E+00I-133 2.30E-072.39E-054.22E-06 3.44E-080.00E+001.12E-123.93E-111.33E-09 3.67E-130.00E+00I-134 1.29E-109.19E-073.30E-06 2.94E-100.00E+006.25E-161.51E-121.04E-09 3.14E-150.00E+00 Cs-134 2.52E-051.61E-043.18E-07 1.55E-091.47E-061.22E-102.65E-101.00E-10 1.65E-141.18E-12I-135 8.32E-081.45E-056.50E-06 3.50E-080.00E+004.04E-132.38E-112.05E-09 3.74E-130.00E+00 Cs-136 2.03E-061.83E-053.77E-08 1.85E-104.96E-089.84E-123.02E-111.19E-11 1.98E-153.97E-14 Cs-137 3.36E-052.14E-044.25E-07 2.06E-092.14E-061.63E-103.52E-101.34E-10 2.20E-141.72E-12 Ba-140 2.52E-075.15E-065.14E-07 2.49E-091.22E-071.22E-128.48E-121.62E-10 2.66E-149.77E-14La-140 3.28E-078.41E-069.39E-07 4.44E-090.00E+001.59E-121.39E-112.96E-10 4.75E-140.00E+00 Ce-141 3.74E-096.21E-086.09E-092.94E-113.08E-081.81E-141.02E-131.92E-12 3.14E-162.47E-14 Ce-143 8.54E-096.40E-071.01E-07 4.44E-100.00E+004.14E-141.05E-123.18E-11 4.75E-150.00E+00 Ce-144 1.14E-071.66E-061.63E-07 7.74E-105.23E-075.53E-132.73E-125.13E-11 8.27E-154.19E-13TABLE 11.A-5 EXPECTED ANNUAL LIQUID EFFLUENT CONCENTRATIONS (DILUTED AND UNDILUTED), BY RADIONUCLIDE, FROM EACH WASTE SYSTEMNUCLIDE BORON RS, UNDILUTED
(µCi/ml)MISC. WASTES, UNDILUTED
(µCi/ml)SECONDARY, UNDILUTED (µCi/ml)TURB. BLDG., UNDILUTED
(µCi/ml)DETERGENT, UNDILUTED
(µCi/ml)BORON RS, DILUTED (µCi/ml)MISC. WASTES, DILUTED (µCi/ml)SECONDARY , DILUTED
(µCi/ml)TURB. BLDG., DILUTED (µCi/ml)DETERGENT, DILUTED (µCi/ml)
MPS2 UFSARMPS2 UFSAR21Rev. 35All Others Except Tritium 3.45E-05 2.43E-04 2.28E-07 2.21E-08 3.10E-07 1.68E-10 4.01E-10 7.18E-11 2.36E-13 2.48E-13 TOTAL 1.01E-047.73E-043.03E-05 1.57E-071.20E-054.88E-101.27E-099.56E-09 1.67E-129.64E-12TABLE 11.A-5 EXPECTED ANNUAL LIQUID EFFLUENT CONCENTRATIONS (DILUTED AND UNDILUTED), BY RADIONUCLIDE, FROM EACH WASTE SYSTEMNUCLIDE BORON RS, UNDILUTED
(µCi/ml)MISC. WASTES, UNDILUTED
(µCi/ml)SECONDARY, UNDILUTED (µCi/ml)TURB. BLDG., UNDILUTED
(µCi/ml)DETERGENT, UNDILUTED
(µCi/ml)BORON RS, DILUTED (µCi/ml)MISC. WASTES, DILUTED (µCi/ml)SECONDARY , DILUTED
(µCi/ml)TURB. BLDG., DILUTED (µCi/ml)DETERGENT, DILUTED (µCi/ml)
MPS2 UFSAR11.A-22Rev. 35TABLE 11.A-6 DESIGN BASIS RADIONUCLIDE CONCENTRATIONS IN LIQUID EFFLUENT, IN FRACTIONS OF 10 CFR PART 20 CONCENTRATION LIMITS NUCLIDEFRACTION OF 10 CFR PART 20 MAXIMUM PERMISSIBLE CONCENTRATIONS 1Na-24 7.79E-05Cr-51 6.77E-07 Mn-54 6.90E-06Fe-55 7.03E-07Fe-59 2.20E-06Co-58 2.03E-08Co-60 5.52E-06Zn-65 2.29E-06W-187 1.27E-05 Np-239 8.54E-06Br-84 2.77E-07Rb-88 1.84E-07Sr-89 1.73E-05Sr-90 1.83E-05Sr-91 2.73E-07Y-91m 1.72E-10Y-91 2.76E-06Y-93 5.18E-07Zr-95 2.87E-06Nb-95 9.34E-07 Mo-99 5.27E-05Tc-99m 2.82E-09 Ru-103 9.79E-07 Ru-106 2.78E-06Ag-110m 5.99E-09Te-129m 1.03E-05Te-129 4.28E-08 MPS2 UFSAR11.A-23Rev. 351. Based on 10 CFR Part 20, Appendix B, Table II, Column 2, prior to 1994.2. In the course of plant operation, individua l isotopic fractions of maximum permissible concentration (MPC) might vary from the va lues tabulated above. The purpose of this table is to demonstrate that the design of th e liquid radioactive waste processing system is adequate at design conditions such that the sum of the fractions of MPCs in the pre-1994 version of 10 CFR Part 20 will not exceed the limit of 1.0.Te-131m 9.77E-06Te-131 1.99E-07 I-131 2.17E-01Te-132 1.89E-04 I-132 8.52E-04 I-133 7.32E-02 I-134 1.25E-04 Cs-134 1.39E-02 I-135 7.69E-03 Cs-136 3.24E-04 Cs-137 5.26E-03 Ba-140 2.86E-06 La-140 3.23E-06 Ce-141 9.79E-07 Ce-143 1.21E-06 Ce-144 7.07E-06H-3 3.54E-04 Sum of MPC Fractions 2 3.19E-01TABLE 11.A-6 DESIGN BASIS RADIONUCLIDE CONCENTRATIONS IN LIQUID EFFLUENT, IN FRACTIONS OF 10 CFR PART 20 CONCENTRATION LIMITS NUCLIDEFRACTION OF 10 CFR PART 20 MAXIMUM PERMISSIBLE CONCENTRATIONS 1
MPS2 UFSAR11.A-24Rev. 35TABLE 11.A-7 DESIGN-BASIS RADIONUCLIDE AIRBORNE CONCENTRATIONS AT THE SITE BOUNDARY FROM ALL GASEOUS EFFLUENT RELEASE POINTS COMBINED, IN FRACTIONS OF 10 CFR PART 20 CONCENTRATION LIMITS IodinesNuclide Fraction of 10 CFR Part 20 Maximum Permissible Concentration (1)I-131 4.13E-04 I-133 7.65E-04Noble GasesNuclide Fraction of 10 CFR Part 20 Maximum Permissible Concentration (1)Kr-85m 7.03E-05Kr-85 7.87E-02Kr-87 2.01E-04Kr-88 7.77E-04 Xe-131m 4.60E-04 Xe-133m 2.07E-04 Xe-133 2.93E-02 Xe-135m 1.85E-05 Xe-135 5.23E-04 Xe-137 none Xe-138 1.16E-04 ParticulatesNuclide Fraction of 10 CFR Part 20 Maximum Permissible Concentration (1)Cr-51 7.83E-09 MPS2 UFSAR11.A-25Rev. 35 Sum of MPC Fractions (2) 1.12E-01.
- 1. Based on 10 CFR Part 20, Appendix B, Table II, Column 1, prior to 1994.2. In the course of plant operation, individua l isotopic fractions of maximum permissible concentration (MPC) might vary from the va lues tabulated above. The purpose of this table is to demonstrate that the design of the gaseous radio active waste processing system is adequate at design conditions such that the sum of the fractions of MPCs in the pre-1994 version of 10 CFR Part 20 will not exceed the limit of 1.0.
Mn-54 3.35E-07Co-57 2.53E-10Co-58 1.38E-06Co-60 2.43E-06Fe-59 8.26E-08Sr-89 9.92E-07Sr-90 2.37E-06Zr-95 8.57E-08Nb-95 6.50E-08 Ru-103 9.19E-10 Ru-1066.99E-11 Sb-125 1.61E-10 Cs-134 9.39E-06 Cs-136 6.49E-07 Cs-137 9.94E-06 Ba-140 2.49E-09 Ce-141 2.28E-08Ar-41 1.57E-04C-14 1.35E-05H-3 1.02E-04 Particulates Nuclid e Fraction of 10 CFR Part 20 Maximum Permissible Concentration (1)
MPS2 UFSAR11.A-26Rev. 35TABLE 11.A-8 TOTAL ANNUAL DESIGN-BASIS AND EXPECTED RELEASES OF RADIOACTIVE LIQUID WASTE TO THE ENVIRONMENT FROM ALL SOURCES COMBINED, IN CURIES PER YEAR NUCLIDE ANNUAL DESIGN BASIS LIQUID EFFLUENT RELEASE (Ci/yr)ANNUAL EXPECTED LIQUID EFFLUENT RELEASE (Ci/yr)
Activated Corrosion Products (Crud)
- Na-24 1.45E+01 4.80E-01Cr-51 1.26E+00 4.10E-02 Mn-54 6.43E-01 2.10E-02Fe-55 5.24E-01 1.70E-02Fe-59 1.23E-01 4.00E-03Co-58 1.89E+00 6.20E-02Co-60 2.57E-01 8.50E-03Zn-65 2.14E-01 6.80E-03W-187 8.31E-01 2.70E-02 Np-239 7.95E-01 2.60E-02 Fission Products
- Br-84 1.03E-01 3.30E-02Rb-88 6.85E-02 2.50E-01Sr-89 4.81E-02 1.80E-03Sr-905.11E-03 1.60E-04Sr-91 1.79E-02 9.20E-03Y-91m 4.82E-04 1.50E-03Y-91 7.73E-02 8.30E-05Y-93 1.45E-02 3.90E-02Zr-95 1.60E-01 5.20E-03Nb-95 8.70E-02 3.80E-03 Mo-99 9.81E+00 7.70E-02Tc-99m 1.58E-02 4.10E-02 Ru-103 7.29E-02 9.80E-02 Ru-106 2.59E-02 1.20E+00Ag-110m 1.67E-04 1.70E-02 MPS2 UFSAR11.A-27Rev. 35Te-129m 2.89E-01 2.50E-03Te-129 3.19E-02 9.30E-02Te-131m 5.46E-01 1.70E-02Te-131 1.49E-02 1.30E-02 I-131 6.05E+01 4.90E-01Te-132 5.28E+00 2.00E-02 I-132 6.35E+00 1.10E+00 I-133 6.82E+01 1.30E+00 I-134 2.33E+00 9.90E-01 Cs-134 1.17E+02 4.60E-01 I-135 2.86E+01 2.00E+00 Cs-136 2.71E+01 4.90E-02 Cs-137 9.79E+01 6.20E-01 Ba-140 7.98E-02 1.60E-01 La-140 6.00E-02 2.90E-01 Ce-141 8.22E-02 2.00E-03 Ce-143 4.50E-02 3.10E-02 Ce-144 6.59E-02 5.20E-02H-3 9.90E+02 9.90E+02TOTAL WITHOUT TRITIUM 4.25E+02 1.10E+01TABLE 11.A-8 TOTAL ANNUAL DESIGN-BASIS AND EXPECTED RELEASES OF RADIOACTIVE LIQUID WASTE TO THE ENVIRONMENT FROM ALL SOURCES COMBINED, IN CURIES PER YEAR NUCLIDE ANNUAL DESIGN BASIS LIQUID EFFLUENT RELEASE (Ci/yr)ANNUAL EXPECTED LIQUID EFFLUENT RELEASE (Ci/yr)
MPS2 UFSAR11.A-28Rev. 35TABLE 11.A-9 TOTAL ANNUAL DESIGN BASIS AND EXPECTED RELEASES OF AIRBORNE RADIOACTIVE WASTE TO THE ENVIRONMENT FROM ALL RELEASE POINTS COMBINED, IN CURIES PER YEAR NUCLIDEANNUAL DESIGN BASIS AIRBORNE EFFLUENT RELEASE (Ci/yr)ANNUAL EXPECTED AIRBORNE EFFLUENT RELEASE (Ci/yr)
I-131 2.23E+01 1.80E-01 I-133 2.89E+01 5.50E-01Kr-85m 3.80E+01 4.00E+00Kr-85 1.28E+05 6.00E+02Kr-87 2.17E+01 4.00E+00Kr-88 8.38E+01 9.00E+00 Xe-131m 9.94E+02 5.00E+01 Xe-133m 3.36E+02 3.00E+00 Xe-133 4.74E+04 1.50E+02 Xe-135m 4.00E+00 4.00E+00 Xe-135 2.83E+02 3.10E+01 Xe-137 0.00E+00 0.00E+00 Xe-138 1.25E+01 3.00E+00Cr-51 3.38E-03 1.10E-04 Mn-54 1.81E-03 5.90E-05Co-57 8.20E-06 8.20E-06Co-58 1.49E-02 4.90E-04Co-60 3.93E-03 1.30E-04Fe-59 8.92E-04 2.90E-05Sr-89 5.35E-03 2.00E-04Sr-90 2.56E-03 8.00E-05Zr-95 4.63E-04 1.50E-05 MPS2 UFSAR11.A-29Rev. 35Nb-95 1.05E-03 4.60E-05 Ru-103 1.49E-05 2.00E-05 Ru-106 7.55E-08 3.50E-06 Sb-125 6.10E-07 6.10E-07 Cs-134 2.03E-02 8.00E-05 Cs-136 2.10E-02 3.80E-05 Cs-137 2.69E-02 1.70E-04 Ba-140 1.35E-05 2.70E-05 Ce-141 6.16E-04 1.50E-05Ar-41 3.40E+01 3.40E+01C-14 7.30E+00 7.30E+00H-3 1.10E+02 1.10E+02TABLE 11.A-9 TOTAL ANNUAL DESIGN BASIS AND EXPECTED RELEASES OF AIRBORNE RADIOACTIVE WASTE TO THE ENVIRONMENT FROM ALL RELEASE POINTS COMBINED, IN CURIES PER YEAR NUCLIDEANNUAL DESIGN BASIS AIRBORNE EFFLUENT RELEASE (Ci/yr)ANNUAL EXPECTED AIRBORNE EFFLUENT RELEASE (Ci/yr)
MPS2 UFSAR11.A-30Rev. 35TABLE 11.A-10 BASIS FOR REACTOR COOLANT SYSTEM ACT IVITY NUREG-0017 GALE CODE INPUT EPower Level, MWt 2754 Percent Failed Fuel 1.0 CVCS Purification Ion Exchanger Decontamination Factors:
Iodine 100 Cs, Rb 2 others 100 Purification Flow Rate (CVCS Pu rification Ion Exchanger), gpm 60Effective Purification Flow Rate for Lithium and Cesium Removal, gpm 0Fission Product Escape Rate Coefficients, sec
-1 Noble Gases 6.5 x 10-8 Halogens, Cs 2.3 x 10-8 Te, Mo 1.4 x 10-9 All others 1.4 x 10-11 Feed and Bleed Liquid Waste, gal./day 2670 MPS2 UFSAR11.A-31Rev. 35FIGURE 11.A-1 ESCAPE RATE COEFFICIENTS MPS2 UFSAR11.B-1Rev. 3511.B RADIOACTIVE WASTE PROCESSING OF RELEASES TO ENVIRONMENT11.B.1 BASES Discussed in this section are th e bases used for determining the expected and design-basis annual releases of radionuclides from the radioactive waste processing sy stem to the environment. The description of radioactive wa ste processing system design and operation provided in this appendix is a representation of how the design and operation have been modeled in the 10 CFR 50 Appendix I, NUREG-0017, GALE Code analysis. GALE, the NRC computer program, was used to calculate expected releases of radioactive material in liquid, gaseous and airborne effluents. The input s to GALE are summarized in Table 11.B-1. The total expected annual releases are given in Appendix 11.A, Tables 11.A-3 and 11.A-4.Expected releases, with anti cipated operational occurrences included, are calculated (using GALE) for the purpose of ascertaining that the radioactive wa ste processing syst ems and building ventilation systems have sufficient capability to ensure that annua l releases will be within the limits specified by 10 CFR Part 50, Appendix I. Design basis releases ar e calculated (as discussed in Appendix 11.A) to demonstrate that the radioactiv e waste processing systems and building ventilation effluents are within the limits of 10 CFR Part 20, Sections 105 and 106 and Appendix B (version prior to January 1, 1994).Actual system operation (discussed in FSAR Section 11.1) may differ from the model input to GALE. However, meeting the appl icable criteria is ensured through the use of process and ef fluent radiation monitoring a nd sampling systems used in conj unction with the Radiological Effluent Monitoring & Offsite Dose Calculation Manual (REMODCM).11.B.2 LIQUID WASTE PROCESSING SYSTEM11.B.2.1 Processing of Clean Liquid Waste The expected releases during nor mal operation from the clean li quid waste processing system (comprised of primary coolant waste and shim bleed) are based on processing and dischar ging of approximately 1,200,000 gallons per year of reactor coolant wastes with quantities as shown in Figure 11.B-1.The clean liquid waste processing sy stem is desig ned to process reactor coolant waste as well as the letdown flow from the CVCS. The following three sources of liquid waste contribute to the clean liquid waste processing system:*The primary drain tank, containing clean liquid waste from containment*The equipment drain sump tank, containing cl ean liquid waste from the auxiliary building*Shim bleed from the primary coolant letdown that has been processed by a deborating or purification ion exchanger MPS2 UFSAR11.B-2Rev. 35 Primary coolant letdown is processed by either the purificat ion or deborating ion exchangers. A portion of what leaves these ion exchangers, referred to as shim bleed, is diverted to the clean liquid waste system for further processing. The remainder of th e primary coolant letdown is passed on to the volume control tank (VCT). In the volume contro l tank, primary coolant liquid is degassed and the gases are purged and sent to the gaseous wast e processing system with the remaining liquid returned to the primary coolant system.
The three liquid sources, listed abo ve, are sent to a mixed-bed ion exchanger. Following processing in the mixed-bed ion ex changers, the liquid is sent to the coolant waste receiver tanks.
From there, the liquid is then sent to a secondary mixed-bed i on exchanger for further processing, after which this liquid is passed into the coolant waste monitor ta nks. From these tanks, the liquid is released to the environment via the discharge canal.
This process stream is shown schematically in Figure 11.B-1. Shown in that figure are the tank volumes, flow rates, and decontam ination factors (DFs) associated with this liquid waste process stream that are inputs to GALE.
11.B.2.2 Processing of Aerated Liquid WasteContributions to the aerated liquid waste processing system include the following:*Waste water from on-site laundry*Drainage from hand wash sinks
- Equipment and area decontamination waste water*Spent fuel pool liner drainage*Primary coolant sampling system drainage
- Auxiliary building floor drainage*Primary coolant leakage from miscellaneous sourcesThese contributions are directed into aerated waste drain tanks. Th e aerated liquid waste collected in these drain tanks is then sent to a mixed-bed ion exchanger for processing, after which it is sent to an aerated waste monitor tank.
The liquid content of this tank is releas ed to the environment via the discharge canal.
11.B.2.3 Processing of Secondary Side Liquid Waste Secondary side liquid waste is comprise d of the following three contributions:*Steam generator blowdown MPS2 UFSAR11.B-3Rev. 35*Turbine building floor drainage*Condensate demineralizer regenerant solutions from condensate polishing facilitySteam generator blowdown is se nt to a blowdown flash tank, wh ere one-third of the liquid blowdown flashes to steam and can be released as gaseous effluent. Turbine building floor drainage flows to the tu rbine building sump and is released to the environment via the discharge canal when radioactivity is detected. With respect to the li quid waste from the condensate polishing facility, the condensate demineralizer regenerant solu tions are sent to the waste neutralization sumps. The liquid in the waste neutralization sumps is released to the environment via the discharge canal.
These three contributions to the secondary si de liquid waste (ident ified above) are shown schematically in Figure 11.B-1.11.B.3 GASEOUS WASTE PROCESSING SYSTEM Airborne releases from the ga seous radioactive waste processi ng system, main condenser air ejector, and containment vents are discharged via the Millstone stack. Containment purges, as well as airborne releases from buildings such as the auxiliary and fuel building, are discharged via the Unit 2 enclosure building roof vent. Turbine building releases are discharged via the Unit 2 turbine building roof vent. Stea m generator blowdown is discharged via a separate blowdown vent located on the Unit 2 enclosure building r oof. For demonstrating compliance with the requirements of 10 CFR Part 20, Sections 1 05 and 106, and Appendix B (version prior to January 1, 1994) and 10 CFR Part 50, Appendix I, maximum dispersion factors derived from annual average meteorological data were assumed in the 10 CFR Part 50 Appendix I analysis.
The sources and pathways of air borne releases are the following:
- Auxiliary Building/Fuel Building Ventilation:
Airborne releases from these sources are normally processed by HEPA filt er and monitored prior to re lease to the environment via the Unit 2 enclosure building roof vent.
- Steam Generator Blowdown Tank Vent:
Approximately one-third of the steam generator blowdown flashes to steam and is released to the environm ent at the Unit 2 enclosure building roof.
- Containment Building: Containment purge air is processed by a HEPA filter and monitored before being released to the envir onment via the Unit 2 en closure building roof vent. Releases of routine venting of contai nment are processed by charcoal adsorber and HEPA filter and monitored befo re being released to the environment via the Millstone stack.*Gaseous Waste Gas Processing System:
Gases from liquid waste processing system tanks and gases stripped in the VCT from the primary coolant letdown flow are stored in six gas MPS2 UFSAR11.B-4Rev. 35decay tanks and monitored before being released to the environment via the Millstone stack. No credit is taken for the HEPA filter because it does not satisfy Reg. Guide 1.140.
- Main Condenser/Air Ejector:
Non-condensable gases from steam flow to the main condenser are monitored and released to the environment via the Millstone stack.
- Turbine Building: Steam leakage into the turbine build ing is released to the environment at the Unit 2 turbine building roof vent.The sources and pathways to the gaseous radioactive waste processing sy stem (identified above) are shown schematically in Figure 11.B-2. These contributions are included in the inputs to GALE and are summarized in Table 11.B-1.
MPS2 UFSAR11.B-5Rev. 35TABLE 11.B-1 INPUTS TO PWR-GALE CODE DESCRIPTIONUNITSVALUEThermal Power LevelMWt 2,754 Mass of Primary Coolant 10 3 lbm 461.1 Primary Coolan t Letdown Rate GPM 60 Letdown Cation Demine ralizer Flow Rate GPM 0Number of Steam Generators 2Total Steam Flow 10 6 lbm/hr11.8 Mass of Liquid in Ea ch Steam Generator 10 3 lbm 132.257Total Blowdown Rate 10 3 lbm/hr 73.7Blowdown Treatment Method 0, 1, or 2 2 Condensate Demineraliz er Regeneration Time days 56 Condensate Demineraliz er Flow Fraction (equal to average flow for al l condensate demineralizers ÷ tota l steam flow)fraction 0.813CLEAN LIQUID WASTE (see notes)
Shim Bleed Rate
- Shim Bleed Flow Rate GPD 2,670 DF for Iodine 10,000 DF for Cs, Rb 200DF for Other Nuclides 5,000Collection Time days 16.66Process and Discharge Time days 0.2865Fraction Discharged 1Equipment Drains (Coolant Waste) Inputs:
Flow Rate GPD 600 PCA (Fraction of Prim ary Coolant Activity) 0.145 DF for Iodine 1,000 DF for Cs, Rb 20DF for Other Nuclides 1,000Collection Time days 16.66 MPS2 UFSAR11.B-6Rev. 35Process and Discharge Time days 0.2865Fraction Discharged 1Clean Wastes Inputs
- None (see notes)AERATED LIQUID WASTE (see note)
Dirty Wastes (Aerated Waste) Inputs:
Flow Rate GPD1,110 PCA (Fraction of Prim ary Coolant Activity) 0.0427 DF for Iodine 100 DF for Cs, Rb 2.0DF for Other Nuclides 100Collection Time days 2.61Process and Discharge Time days 0.0598Fraction Discharged 1 MISC. other liquid waste inputs (see note)
Steam Generator Blowdown:
Blowdown Fraction Processed 1 DF for Iodine 1 DF for Cs, Rb 1DF for Other Nuclides 1Collection Time days 0Process and Discharge Time days 0Fraction Discharged 1 Regenerant Inputs
- Regenerant Flow Rate GPD 850 DF for Iodine 1 DF for Cs, Rb 1DF for Other Nuclides 1Collection Time days 0Process and Discharge Time days 0Fraction Discharged 1TABLE 11.B-1 INPUTS TO PWR-GALE CODE (CONTINUED)DESCRIPTIONUNITSVALUE MPS2 UFSAR11.B-7Rev. 35 Misc. other GALE Code INPUTS
- Continuous Stripping of Full Letdown Flow 2Hold-up Time For Xenon days 90Hold-up Time for Krypton days 90Fill Time for Decay Tanks for Gas Stripping days 90AIRBORNE INPUTS
- Waste Gas System Particulat e (HEP A) Filter Removal Efficiency 0%Fuel Handling Building Charcoal Filter Removal Ef ficiency 0%Fuel Handling Building HEPA Filter Removal Efficiency 99%Auxiliary Building Charcoal Filter Removal Ef ficiency 0%Auxiliary Building HEPA Filter Removal Ef ficiency 99%Containment Volume 10 6 ft 3 1.899 Containment Atmosphere Cl ean-up Ch arcoal Removal Efficiency 0%Containment Atmosphere Clean-up HEPA Removal Efficiency 0%Containment Atmos phere Clean-up Rate 10 3 CFM 0Containment High Volume (Shutdown) Purge Charcoal Removal Efficiency 0%Containment High Volume (Shut down) Purge HEPA Filter Removal Efficiency 99%No. of Purges per Year (2 purg es internal to GALE code, see notes)0Containment Low Volume Purge (Normal) Charcoal
Ef ficiency (i.e.: containment venting, see note) 70%Containment Low Volume Purge (Normal) Filter Efficiency (i.e.: containment venting, see note) 99%Containment Low Volume Purge (Normal) Rate (i.e.: containment venting, see note)
CFM 10 Fraction of Iodine Releas ed from Blowdown Tank Vent 0.05TABLE 11.B-1 INPUTS TO PWR-GALE CODE (CONTINUED)DESCRIPTIONUNITSVALUE MPS2 UFSAR11.B-8Rev. 35NOTES: The component and system designa tions used in this table reflect the terminology used in NUREG-0017, GALE Code, and may differ fr om that used by Millstone Unit 2.
The GALE Code has intern al calculations that impact the input data. The data provided as GALE Code inputs in th is table and Figures 11.B-1 and 11.B-2 may differ from data listed in other tables in the FSAR. A specifi c example of this are the tank volumes listed on Figure 11.B-1 versus Table 11.1-1. The GALE Code ap plies a 0.8 or 0.4 factor to tank volumes to calculate collection, pr ocess, dischar ge and decay times.
Percent of Iodine Removed From (Main Condenser) Air Ejector Release 0DETERGENT WASTES
- Detergent Waste PF (Purification Factor), (DF)
-1 0.1TABLE 11.B-1 INPUTS TO PWR-GALE CODE (CONTINUED)DESCRIPTIONUNITSVALUE MPS-2 FSAR Rev. 22FIGURE 11.B-1 NUREG 0017 GALE CODE INPUT DIAGRAM - LIQUID WASTE SYSTEML
MPS2 UFSAR11.C-1Rev. 3511.C DOSES FROM RADIOACTIVE RELEASES AND COST-BENEFIT ANALYSIS11.C.1 DOSES TO HUMANSThe analysis of annual doses to the maximum i ndividual and to the populat ion residing within an 50 mile radius of Millst one Unit 2 are based on the methodology and equations presented in U. S.
NRC Regulatory Guide 1.109 Rev. 1, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for The Purpos e of Evaluating Compliance with 10 CFR Part 50, Appendix I. To automate the Millstone Unit 2 pathways-to-man dose analyses, U.S. NRC Computer Code LADTAP II was used for the liquid effluent releases, and U.S. NRC Computer Code GASPAR was used for the gaseous effluent releases.
The Regulatory Guide 1.109 equations germane to liquid and gaseous releases are presented in Sections 11.C.2 and 11.C.3, respectively. In the Millstone Unit 2 liquid and ga seous pathways-to-man dose anal yses, the NRC default values provided in Regulatory Guide 1.109 were used as input to LADTAP II and GASPAR whenever site-specific data were unavailable.
The ensuing sections present the equations
/methodologies employed in LADT AP II and GASPAR to calculate the dose contributions co rresponding to the release/uptake pathways considered in the Millstone Unit 2 pathways-to-man dose analyses.11.C.2 METHODS FOR CALCULATING DOSES FROM LIQUID RELEASESLADTAP II was used to calculate the radiation exposure to man fr om ingestion of aquatic foods, shoreline deposits, swimming, and boating. Do ses are calculated for both the maximum individual and for th e population and are summarized for each pathway by age group and organ. LADTAP II also calculates the doses to certain representa tive biota other than man in the aquatic environment, such as fish, invertebrates, alga e, muskrats, herons, and ducks, using the models presented in WASH-1258.
The equations and assump tions used in the LADTAP II anal ysis are presented in the ensuing subsections. Final dose results based on pathway doses calculated by LADTAP II are presented in Table 11.C-1.11.C.2.1 Generalized Equation for Calculating Radiat ion Doses to Humans via Liquid Pathways R aipj = (C jp) . (U ap) . (D aipj)where: R aipj = the annual dose to organ j of an individual of age group a from nuclide i via pathway p, in mrem/yr C ip = the concentration of nuclide i in the media of pathway p , in pCi/l, pCi/kg, or pCi/m 2 U ap = the exposure time or intake rate (usage) associated with pathway p for age group a , in hr/yr, yr
-1 , or kg/yr (as appropriate)
MPS2 UFSAR11.C-2Rev. 35 D aipj = the dose factor, specific age group a , radionuclide i , pathway p, and organ j, in mrem/pCi ingested, mrem/hr per pCi/m 2 from exposure to deposited activity in sediment or on the ground, or mrem/hr per pCi/liter due to exposure from boating and swimming11.C.2.2 Doses from Aquatic Foods R apj = (1,100)[(U ap)(M p)/F] . i Q i l. B ip . D aipj . where: B ip = the equilibrium bioaccumul ation factor for nuclide i in pathway p , expressed as the ratio of the concentration in biota (in pC i/kg) to the radionuclid e concentration in water (in pCi/l), in l/kg M p = the mixing ratio (reciprocal of the dilution factor) at the point of exposure or point of harvest of aquatic food, dimensionless F = the flow rate of the liquid ef fluent in ft 3/sec Q i = the release rate of nuclide i , in Ci/yr R apj = the total annual dose to organ j of individuals of age group a from all of the nuclides i in pathway p , in mrem/yr i = the radioactive decay constant of nuclide i , in hours t p = the average transit time required for nucli des to reach the point of exposure. For internal dose, t is the total time elapsed betw een release of the nuclides and ingestion of food, in hours.
1,100 = the factor to convert from (Ci/yr)/(ft 3/sec) to pCi/l All the other symbols are as previously defined.11.C.2.3 Doses from Shoreline Deposits R apj = 110,000 x [(U ap)(M p)(W)/F] x i Q i T i D aipj x x (1-)where: W = the shoreline width factor that describes the geometry of the e xposure, dimensionless T i = the radiological half-life of nuclide i , in days t b = the period of time for which sediment or soil is exposed to the contaminated water, in hours ei-t p ei-t p ei-t b MPS2 UFSAR11.C-3Rev. 35110,000 = the factor to convert from Ci/yr per ft 3/sec to pCi/l and to account for the proportionality constant used in the sediment radioactivity model All other symbols are as previously defined.11.C.2.4 Doses from Swimming and Boating The doses from swimmin g and boating were cal culated using the methodology described in WASH 1258 (Atomic Energy Commission 1973).
The equation for calculation of the external dos e to skin and the total body dose from swimming (water immersion) or boating (water surface) is:
R apj = 1,000 x [(U ap)(M p)/(F)(K p)] x i Q i D aipj x where: K p = geometry correction factor equal to 1 for swimming a nd 2 for boating, dimensionless (no credit is taken for the shielding provided by the boat).
All other symbols are as previously defined.11.C.3 METHOD FOR CALCULATING DOSES FROM GASEOUS RELEASESFASPAR implements the air release dose mode ls of the NRC Regulatory Guide 1.109 for noble gases (semi-infinite plume onl y) and for radioiodine and pa rticulate emissions. GASP AR computes both populat ion and individual doses using site data, me teorological data, and radionuclide release source terms as inputs. Location-dependent me teorological data for selected individuals are specified as input data. The site data include the popula tion distribution and the quantities of milk, meat , and vegetation produced. The meteor ological data include dispersion ÚQ, /Q decayed, /Q decayed and depleted, and deposition D/Q. Population doses, individual doses, and cost benefit results are calculated.
There are two basic types of calculations, the indi vidual dose calculation and the population dose calculation. Seven pathways by which the nuclides tr avel to man are consid ered. These are plume, ground, inhalation, vegetation, cows' milk, goats
' milk, and meat. Fo r the individual dose calculations, man is subdivided into the four age groups of infant (0 to 1 year), child (1 to 1 1 years), teenager (12 to 18 years), and adult (over 18 years). Each of these calculations takes into account eight body organs: total body , gastrointestinal (GI) tract, bone, liver, kidney, thyroid, lung, and skin.
The equations and assumptions used in the GASPAR analysis are presented in the ensuing subsections. Final dose results by GASPAR, based on activity releases tabulated in Table 11.A-3 , are presented in Table 11.C-1.ei-t p MPS2 UFSAR11.C-4Rev. 3511.C.3.1 Gamma and Beta Doses from Noble Gas Dischar ged to the Atmosphere11.C.3.1.1 Annual Air Doses from Noble Gas Releases (Non-Elevated)*Annual Gamma Air Dose Equation:*Annual Beta Air Dose:
where: D(r,) = the annual gamma air doses at the distance r in the sector at angle from the discharge point in mrad/year D(r,) = the annual beta air doses at the distance r in the sector at angle from the discharge point in mrad/year Q i = the release rate of the radionuclide i , in Ci/year/Q (r,) = the annual average gaseous dispersion factor at distance r in sector , in sec/
m 3 (corrected for radioactive decay)
DFi = the gamma air dose factors for a unifo rm semi-infinite cloud of radionuclide i , in mrad*m 3/pCi*yr DFi = the beta air dose factors for a uniform semi-infinite cloud of radionuclide i , in mrad*m 3/pCi*yr 3.17 x 10 4 = the number of pCi per Ci divided by the numbe r of seconds per year Dr(,)3.17x10 4i Q i X Qr(,)DFi=Dr(,)3.17x10 4i Q i X Qr(,)DFi
=
MPS2 UFSAR11.C-5Rev. 3511.C.3.1.2 Annual Total Body Dose from Nobl e Gas Releases (Non-Elevated) where: D T(r,) = the total body dose due to immersion in a semi-infinite cloud at distance r in sector , in mrem/year S F = the attenuation factor that accounts for dose reduction due to shielding provided by residential structures, dimensionlessi (r,) = the annual average ground-level concentration of radionuclide i at distance r in sector , in pCi/ m 3 DFB i = the total body dose factor for a semi-infinite cloud of the radionuclide i , which includes radiation attenuati on through a depth of 5 cm into the body, in mrem*m 3/pCi*yr11.C.3.1.3 Annual Skin Dose from Noble Gas Releases (Non-Elevated) where: D S (r ,) = the annual skin dose due to immersion in a semi-infinite cloud at the distance r in sector , in mrem/yr DFS i = the beta skin dose f actor for a semi-infinit e cloud of radionuclide i , which includes the attenuation by the outer "dead" layer of the skin, in mrem*m 3/pCi*yr1.11 = the average ratio of tissue to air energy absorption coefficients All other parameters are as previously defined.
D Tr(,)S Fii r(,)DFB i=D Sr(,)1.11S Fii r(,)DFiii r(,)DFS i+=
MPS2 UFSAR11.C-6Rev. 3511.C.3.1.4 Annual Gamma Air Dose and Annual Total Body Dose Due to Noble Gas Releases from Free-Standing Stacks More Than 80 Meters Tall Regulatory Guide 1.109 provides gamma air dos e and total body dose eq uations that are applicable to elevated releases of noble gases from free-st anding stacks that are more than 80 meters tall. However, GASPAR does not incl ude these equations, and any attempt to use GASPAR to calculate cloud shine dos es from elevated re leases would lead to non-conservative results within approximately 1 kilometer from the point of release. What was done in the Millstone Unit 2 analysis was to simulate an elevated release by making a separate GASPAR run using the stack release radi onuclide inventory as input, but with ground-level release /Qs used in lieu of stack release /Qs. This conservative approach wa s taken to calcula te noble gas gamma shine doses (both gamma air doses and total body doses) to maximum individuals.11.C.3.2 Doses from Radioiodines and Other Radionuclide s, Exclusive of Noble Gases, Released to the Atmosphe re11.C.3.2.1 Annual Organ Dose Due to External Irradiation from Ground Deposition of Radionuclides where: D G j (r,) = the annual dose to organ j at location (r ,), in mrem/yr S F = a shielding factor that accounts for the dose reduction due to shielding provided by residential structures dur ing occupancy (assumed to be 0.7), dimensionless C G i (r,) = the ground plane concentration of radionuclide i at distance r in sector , in pCi/m 2 DFGij = the open field ground plane dose conversion factor for organ j from radionuclide i , in mrem*m 2/pCi*hr 8,760 = the number of hours in a year11.C.3.2.2 Annual Organ Dose from Inhala tion of Radionuclides in Air D G j r(,)8760S Fi C G i r(,)DFG ij,=D A ja r(,)R aii r(,)DFAija=
MPS2 UFSAR11.C-7Rev. 35 where: D A ja (r,) = the annual dose to organ j of an individual in the age group a at location (r,) due to inhalation, in mrem/yr R a = the annual air intake for individuals in the age group a , in m 3/yr i (r,) = the annual average concentration of radionuclide i in air at location (r,), in pCiÚm 3DFAija = the inhalation dose f actor for radionuclide i, organ j , and age group a , in mremÚpCi11.C.3.2.3 Annual Organ Dose from Ingestion of Atmo spherically Released Radionuclides in Food where: C v i(r,), C m i(r,), C L i(r,), C F i(r,) = the concentrations of radionuclide i in produce (non-leafy vegetables, leafy vegeta bles, fruits, and grains), milk, and meat, at location (r ,), in pCi/kg or pCi/l D D ja(r,) = the annual dose to the organ i of an individual in age group a from ingestion of produce, milk, leafy vegetables, and meat at location (r
,), in mrem/year DFIija = the ingestion dose factor for radionuclide i, organ j , and age group a in mrem/pCi f g , f l = the respective fractions of the ingestion rates of produce and leafy vegetables that are produced in the garden of interest U v a , U m a , U F a , U La = the annual intake (usage) of produce, milk, meat, and leafy vegetables, respectively, for individuals in the age group a , in kg/yr or l/yr11.C.4 COMPARISON OF CALCULATED AN NUAL MAXIMUM INDI VIDUAL DOSES WITH APPENDIX I DESIGN OBJECTIVESA comparison of calculated annual maximum individual doses with 10 CFR Part 50 Appendix I design objectives is provided in Table 11.C-1.D D ja r(,)i DFIija U V a f g C V i r(,)U m a C m i r(,)U F a C F i r(,)U L a f 1 C L i r(,)+++=
MPS2 UFSAR11.C-8Rev. 3511.C.5 GENERAL EXPRESSION FOR POPULATION DOSES The general expression for calculating the annual population-integrated dose is:
where: D P j = the annual population-integrated dose to organ j (total body or thyr oid), in man-rem or thyroid man-rem P d = the population associated with sub-region d D jda = the annual dose to organ j (total body or thyroid) of an average individual of age group a in sub-region d , in mrem/yr f da = the fraction of the population in sub-region d that is in age group a 0.001 = the conversion factor from mrem to rem The equation above, used in conjunction with the preceding equations with parameters adjusted for each age group, is used to calculate the popul ation doses. The populati on doses due to annual releases of expected liquid and airborne radionuclide conc entrations are presented in Table 11.C-2.11.C.6 COST-BENEFIT ANALYSIS Presented in this section are the cost-benefit analyses and their results, performed in accordance with the requirements set forth in 10 CFR Part 50, Appendix I, Section II.D. In these analyses, potential augments to the liqui d and gaseous radioactive waste processing systems are examined for cost-effectiveness using the methodology and data provided in U. S. NRC Regulatory Guide 1.110 Rev. 0 (March 1976), Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors. The beneficial savings of each augment is calculated by multiplying the calculated dose reduction by $1,000 per man-rem. The cost of borrowed money is conservatively assumed to be 9%
(1). Provided in Appendix 11.B is information pertinent to the model input to GALE representing the design and configuration of the Millstone Unit 2 radioactive waste processing systems.
(1) This is predicated on Northeast Utilities' June 1998 Cost of Capital Study, in which it was concluded that the weighted cost of borrowed money for NU, CL&P, WMECO, PSNH, and combined system is 9.3%. Based on th is conclusion, it is conservative in the cost-benefit analysis to round down to 9%.
D P j 0.001d P daD jda f da=
MPS2 UFSAR11.C-9Rev. 3511.C.6.1 Procedure Used for Performing Cost-Benefit AnalysisThe procedure employed for performing the co st-benefit analyses presented in Sections 11.C.6.2 and 11.C.6.3 below is taken directly from Regulatory Guide 1.1 10 Rev. 0. This procedure is summarized in the following steps:1.Obtain the direct cost of equipment and materials from Table A-1 in Regulatory Guide 1.110.2.Multiply the direct labor cost obtaine d from Table A-1 in Regulatory Guide 1.110 by the appropriate labor cost correction factor, obtained from Table A-4 in Regulatory Guide 1.110, to obtain the corr ected labor cost for the geographical region in which the plant is located (Millstone Unit 2 being in Region I).3.Add the costs obtained from the previous tw o steps to obtain the total direct cost (TD C).4.Obtain the appropriate indirect cost factor (IC F) from Table A-5 in Regulatory Guide 1.110.5.Determine the total capital cost (TCC) using the following equation:
TCC = TDC x ICF 6.Obtain the appropriate ca pital recovery factor (CRF) from Table A-6 in Regulatory Guide 1.110.7.Determine the annual fixed cost (AFC) using the following equation:
AFC = TCC x CRF8.Obtain the annual operating cost (AOC) and the annual maintenance cost (AMC) from Regulatory Guide 1.110 Tables A-2 and A-3, respectively.9.Determine the total annual cost (TA C) using the following equation:TAC = AFC + AOC + AMC10.Determine the benefit of each augmen t by multiplying the dose reduction to be achieved by $1,000 per man-rem.11.The benefit calculated in the previous step minus TA C provides the net benefit of adding the augment to the radwaste system. A positiv e net benefit means that adding the augment would be cost-beneficial; conversely, a negative net benefit means that it would not be cost-beneficia l to add the augment to the radwaste system.
MPS2 UFSAR11.C-10Rev. 3511.C.6.2 Augments to the Liquid Radioactive Waste Processing SystemTable 11.C-2 presents the calculated base case annual total body dose (man-rem), annual maximum organ dose (man-rem), and annual thyroid dose (man-rem) associated with the operation of the Millstone Unit 2 liquid radioactive waste processing system for the population expected to live within a 50 mile ra dius of the plant. In the cost-b enefit analysis performed for the liquid radioactive waste proc essing system, the augment th at was chosen was a 50-gpm demineralizer on the blow-down waste stream. This demineralizer was selected be cause it was judged to be the least expensive augment that could provide any benefi t in the reduction of population dose. Assuming that this augment is capable of reducing the population doses to zero (an extremely conservative assumption), the maxi mum annual benefit to be realized would be approximately $4,880. However, off-sett ing this benefit is a total annua l cost associated with this augment that is estimate d to be approximately $43,800.
Conclusion
- It would not be cost-beneficial to a dd additional liquid radwaste processing equipment to the existing system.
The salient inputs to the liquid ra dioactive waste processing system cost-benefit analysis (taken from the tables in Regulatory Guide 1.1
- 10) are summarized below:*Equipment/material cost: $43,000
- Labor cost: $29,000*Labor cost correction factor: 1.6*Indirect cost factor (ICF):
1.58, based on the fact that Millst one is a three unit site, with each unit having its own radwaste system*Capital recovery factor (CRF): 0.0973
- Annual operating cost (AOC): $25,000*Annual maintenance cost (AMC): $5,000*Dose reduction goal: 4.88 man-re m (assuming that the goal is to reduce to zero the largest population dose resulting from liquid effluent released to th e environment, that population dose having been calculated to be the maximum organ dose)11.C.6.3 Augments to the Gaseous Radioactive Waste Processing SystemTable 11.C-2 presents the calculated base case annual total body dose (man-rem) and annual thyroid dose (man-rem) associated with the operation of the Un it 2 gaseous radioactive waste processing system for the population expected to live within a 50 mile radius of the plant. In the cost-benefit analysis performed for the gaseous radioactive wast e processing system, the augment that was chosen was a pair of 30,000 cfm charcoal/HEPA filtration sy stems on the Unit 2 MPS2 UFSAR11.C-11Rev. 35auxiliary building vent.
This choice was made because it was judged to be the least expensive augment that could provide any benefit in the reduction of population dose. Assuming that this augment is capable of reduci ng the population doses to zero (an extremely conservative assumption), the maximum annua l benefit to be realized would be approximately $19,600. However, off-setting this benefi t is a total annual cost associated with this augment that is estimated to be approximately $127,500.
Conclusion
- It would not be cost-beneficial to a dd the two 30,000-cfm charcoal/HEPA filtration systems to the existing gaseous radioactive wast e processing system. The salient inputs to the gaseous radioactive waste processing system cost-benefit analysis (taken from the tables in Regulatory Guide 1.110) are summarized below:*Equipment/material cost: $314,000*Labor cost: $102,000
- Labor cost correction factor: 1.6*Indirect cost factor (ICF):
1.58, based on the fact that Millst one is a three-unit site, with each unit having its own radwaste system*Capital recovery factor (CRF): 0.0973*Annual operating cost (AOC): $18,000
- Annual maintenance cost (AMC): $36,000*Dose reduction goal: 19.6 man-re m (assuming that the goal is to reduce to zero the largest population dose resulting from liquid effluent released to th e environment, that population dose having been calculated to be the thyroid dose)
MPS2 UFSAR11.C-12Rev. 351 Per reactor.TABLE 11.C-1 COMPARISON OF CALCULATED ANNUAL MAXIMUM INDIVIDUAL DOSES WITH 10 CFR PART 50 APPENDIX I DESIGN OBJECTIVES 10 CFR PART 50 APP ENDIX I DESIGN OBJECTIVE DOSE 1CALCULATED ANNUAL DOSEAirborne Effluent
- Gamma Air Dose 10 mrad 0.2 mrad*Beta Air Dose 20 mrad 0.078 mrad*Total Body Dose 5 mrem 0.15 mrem*Skin Dose 15 mrem 0.25 mrem*Maximum Organ Dose (Thyroid) 15 mrem 8.3 mremLiquid Effluent
- Total Body Dose 3 mrem 0.06 mrem*Maximum Organ Dose (GI-LLI) 10 mrem 0.91 mrem MPS2 UFSAR11.C-13Rev. 35TABLE 11.C-2 ANNUAL TOTAL BODY AND THYROID DOSES TO THE POPULATION WITHIN 50 MILES OF THE MILLSTONE SITE, IN MAN-REM, FROM EXPECTED LIQUID AND AIRBORNE EFFLUENT RELEASES 1.Annual Total Population Doses from Airborne Effluent Releases:* Total Body Dose: 0.607 man-rem
- Thyroid Dose:
19.6 man-rem2.Annual Total Population Doses fr om Liquid Ef fluent Releases:* Total Body Dose: 1.89 man-rem
- Thyroid Dose:
2.86 man-rem* Maximum Organ Dose (GI-LLI): 4.88 man-rem MPS2 UFSAR11.D-1Rev. 3511.D EXPECTED ANNUAL INHALATION DOSES AND ESTIMATED AIR CONCENTRATIONS OF RADIOACTIVE ISOTOPES FOR MP2 FACILITIES The expected annual inha lation doses to plant personnel and es timated air concentrations were computed as outlined below:
Equilibrium airborne concentrations:
Equilibrium concentration
= (Production/Losses) = fA/where: f = flow rate of activity source (Ci/sec) A = activity of source (Ci/cc)
= loss rates = d = e + R e = leakage or exhaust R = recirculation for iodines, if any Doses: The whole body dose is determined using a semi-infinite cloud model (Safety Guide 4). The thyroid inhalation dose is found using the inhala tion model and dose conve rsion factors of TID 14844 and Safety Guide 4. The other organ doses are found by the fraction of allowable MPC, which yields the limiting doses.
Whole Body Doses
Lung Dose:
where: C I = Equilibrium concentration of the I th isotope MPC I Lung = Maximum permissible concentration for the I th isotope for lung based on a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> workweek taken from ICRP II for the airborne case MPD Lung = Maximum permissible yearly dose to lung For Cs 138 and Rb 88, D I Lung is computed as follows:
D Lung = 2.81 x 10 5 x T x E x Cwhere: Y = radiological half life D I Lung C I MPC I Lung----------------------------
-MPD Lungx=
MPS2 UFSAR11.D-2Rev. 35E = effective energy absorbed per di sintegration in MeV/disintegrationC = Concentration in
µCi/cc of CS 138 or 88 Rb.The total lung dose D L is found as follows:
Thyroid Dose:
D I Thyroid = CI x t x BR x DCF Iwhere:D I Thyroid = Thyroid dose in REMs C I = Concentration of the Iodine isotope I t = Exposure time in day = 300 days BR = Breathing rate in cc/day = 2 x 10 7 cc/day DCF = Dose conversion factors fo r Iodine in REMs/Ci per TID 14844 The total thyroid dose is found as follows:
Bases and Results:
Source terms are based on reactor opera tion with 0.1 percent failed fuel.
Containment Building:1.Containment volume - 1.92 x 10 6 ft 3 2.Purge containment for a minimum of fi ve hours at 32,000 CFM pr ior to personnel entering the containment (see FSAR Section 9.9.2).Y Min. E(MeV/dis)
CS 138 32.2 1.39 Rb 88 17.8 1.68 D L nI1=D I Lung=D TI1=n D IThyroid=
MPS2 UFSAR11.D-3Rev. 353.Activity source is conservatively as sumed to be 40 GPD of reactor coolant leakage. The expected and calculated leak age of reactor coolan t to the containment atmosphere will be a small fraction of the 40 GPD rate us ed in the analysis. The 40 GPD leakage rate was taken for "T welfth AEC Air Cleaning Conference-Analysis of Power Reactor Gaseous Waste System."4.Assume 75 days of activity buildup in the containment.5.A decontamination factor of 10, used for iodines and particulates, was taken from "T welfth AEC Air Cleaning Conference-Analysis of Power Reactor Gaseous Waste System."The results indicate the whole body dose rate will be 0.85 mr/hr. After 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the dose rate decreases to 0.44 mrem/hr. Concentrations, given in Table 11.D-1 , are below the 10 CFR Part 20 allowable airborne concentrations.
Auxiliary Building (Excludes nonradioac tive areas and fu el handling area):1.Assume 20 GPD leakage of 120
°F fluid and 1 GPD leakage of 550
°F fluid with reactor coolant concen trations of activity
.2.Ventilation air flow is 40,000 CFM for radi oactive areas of auxiliary building. This corresponds to approximately five air changes per hour
.3.Iodine and particulate partit ion factors assumed as follows:
Coolant at 120
°F 4 Coolant at 550
°F 14.2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of exposure per year.
Doses based on 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, 52 week s per year exposure are as follows:
Lung Dose - 0.046 REM Whole Body Dose - 0.468 REM
Thyroid Dose - 0.61 REM Airborne concentrations are given in Table 11.D-2.Auxiliary Building (Fuel handling area):1.Sources arise due to evaporation from the spent fuel pool surface.2.Sources assumed as one-tenth of reactor coolant concentrations (0.1% failed fuel).
Only tritium is considered since separati on factors for iodines and particulates in MPS2 UFSAR11.D-4Rev. 35 spent fuel pool water will greatly redu ce airborne concentrations. Contribution from noble gases will be insignificant.3.Evaporation rate from spent fuel pool at 120
°F is 0.37 gpm.4.There will be 20 air changes per hour at average exhaust rate of 18,000 CFM (F SAR Section 9.9).5.Only whole body dose is considered as the critical organ for tritium.
6.Exposure calculated for 2000 hrs/year.The airborne equilibrium concentration for tritium is 1.48 x 10
-8µCi/cc. The whole body dose is 25.2 mr/yr
.Turbine Building1.Sources arise due to leakage from the s econdary systems into the turbine building.2.Primary coolant system ope rating at 0.1% failed fuel.3.Primary to secondary leakage of 100 gallons per day.4.5 gpm continuous blowdow n per steam generator
.5.Processed blowdown liquid is discharged.6.Feedwater cleanup factor is 0.7.Steam leakage from the secondary side is 1700 lb/hr.8.Liquid leakage from the secondary side is 15 gpm.9.Volume of the turbine building 2,882,000 ft 3.10.Exhaust flow rate is 200,000 cfm.
Doses based on 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, 50 week s per year exposure are as follows:
Lung Dose - 7.8 x 10
-3 REM Whole Body Dos - 2.4 x 10
-5 REM Thyroid Dose - 0.314 REM Airborne concentrations are given in Table 11.D-3.
MPS2 UFSAR11.D-5Rev. 35TABLE 11.D-1 CONTAINMENT BUILDING AIRBORNE CONCENTRATIONSISOTOPE CONCENTRATION
µCi/cc I-131 1.31 x 10-8 I-132 9.75 x 10-10 I-133 7.46 x 10-9 I-134 3.99 x 10-10 I-135 2.95 x 10-9 Xe-131M 6.35 x 10-8 Xe-133M--Xe-133 4.60 x 10-6 Xe-135M--Xe-135 8.61 x 10-8 Xe-137--Xe-138 1.24 x 10-9Kr-83M--Kr-85M 1.52 x 10-8Kr-85 1.35 x 10-7Kr-87 6.07 x 10-9Kr-88 2.43 x 10-8Kr-89--
MPS2 UFSAR11.D-6Rev. 35TABLE 11.D-2 AUXILIARY BUILDING AIRBORNE CONCENTRATIONS ISOTOPE CONCENTRATION µCi/ccKr-85M 8.90 x 10-9Kr-85 5.45 x 10-9Kr-87 4.52 x 10-9Kr-88 1.53 x 10-8 Xe-131M 9.10 x 10-9 Xe-1331.11 x 10-6 Xe-135 4.56 x 10-8 Xe-138 1.42 x 10-9 I-129 2.96 x 10-18 I-131 1.63 x 10-10 I-132 4.23 x 10-11 I-133 2.31 x 10-10 I-134 2.21 x 10-11 I-135 1.09 x 10-10Br-84 1.53 x 10-12Rb-88 7.24 x 10-11Rb-89 1.72 x 10-12Sr-89 2.08 x 10-13Sr-90 1.07 x 10-14Sr-91 1.44 x 10-13Te-129 9.24 x 10-13Te-132 1.35 x 10-11Te-134 9.08 x 10-13 Be-140 2.50 x 10-13 Ru-103 1.69 x 10-13 MPS2 UFSAR11.D-7Rev. 35 Ru-106 1.02 x 10-15 La-140 2.39 x 10-13 Ce-144 1.69 x 10-13Pr-143 2.39 x 10-13 Mo-99 8.31 x 10-11Y-90 4.17 x 10-14Y-91 4.51 x 10-12 Cs-134 4.10 x 10-12 Cs-136 1.05 x 10-12 Cs-137 1.31 x 10-11 Cs-138 2.27 x 10-11Cr-51 1.56 x 10-14 Mn-54 1.31 x 10-14Fe-59 8.73 x 10-15Co-58 1.91 x 10-12Co-60 2.13 x 10-13Zr-95 3.83 x 10-16TABLE 11.D-2 AUXILIARY BUILDING AIRBORNE CONCENTRATIONS ISOTOPE CONCENTRATION
µCi/cc MPS2 UFSAR11.D-8Rev. 35TABLE 11.D-3 TURBINE BUILDING AIRBORNE CONCENTRATIONS ISOTOPE CONCENTRATION Ci/ccH-3 5.14E-10Cr-51 1.59E-15 Mn-54 1.21E-15Fe-59 9.08E-16Co-58 2.01E-13Co-60 2.29E-14Br-84 1.09E-14Kr-85M 4.61E-12Kr-85 2.88E-12Kr-87 2.19E-12Kr-88 7.80E-12Rb-88 5.74E-14Rb-89 1.17E-15Sr-89 2.17E-14Sr-90 1.15E-15Sr-91 3.24E-15Y-90 2.85E-15Y-91 4.72E-13Zr-95 4.04E-17 Mo-99 5.75E-12 Ru-103 1.75E-14 Ru-106 1.09E-16Te-129 2.96E-15Te-132 9.90E-13Te-134 1.80E-15 I-129 1.56E-18 I-1317.22E-11 I-132 1.29E-12 I-1334.32E-11 MPS2 UFSAR11.D-9Rev. 35Where E-14 = 10
-14 , etc.I-134 2.60E-13 I-135 8.81E-12 Xe-131M 4.83E-12 Xe-133 5.89E-10 Xe-1352.41E-11 Xe-138 5.88E-13 Cs-134 4.40E-13 Cs-136 1.01E-13 Cs-137 1.41E-12 Cs-138 3.36E-14 Ba-140 2.40E-14 La-140 1.34E-14 Ce-144 1.81E-14Fr-143 2.24E-14TABLE 11.D-3 TURBINE BUILDING AIRBORNE CONCENTRATIONS ISOTOPE CONCENTRATION Ci/cc MPS2 UFSAR11.E-1Rev. 3511.E AIRBORNE ACTIVITY SAMPLING SYSTEM FOR CONTAINMENT, SPENT FUEL AND RADWASTE ATMOSPHERES In-plant control of personnel exposure from airborne radioactivity will be ef fected by a continuing program of sampling for airborne activity and ad ministrative controls through radiation work permits. 1.Containment atmosphere is continuously monitored by two sample lines inside containment. The sample is pulled through a fixed particulate filter , a charcoal cartridge and a gas chamber before being pumped back to the containment.
Particulate activity is measured by a beta scintillator through a derivative (rate of change) circuit. Any rapid changes in beta activity will be alarmed in the control room. Gaseous iodine is absorbed on the charcoal cartridge. The particulate filter is changed on a routine schedule to prev ent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with the particulate filter. The rem oved filter and cartridge ma y be counted in accordance with approved station procedures, if neede d, to evaluate the airborne radioactivity concentration during the co llection period. Gaseous activ ity is measured with a beta scintillator type detector. Increased ac tivity above pre-set levels will alarm in the control room.
Upon indication of a significant increase in activity in the containment building, the Shift Supervisor will notify radiation protection. Radiation protection will obtain and analyze samples of contai nment air to determine the isotopes responsible for the increased activity
. During power operation, if conditions warrant, radiation protection personnel may enter the containment and take
portable air samples in selected areas in order to locate the source of the activity.
Appropriate respiratory equi pment will be indicated on the radiation work permits.
During shutdown and maintenance conditions, air samples will be taken with the
portable sampling equipment throughout th e containment building on a scheduled basis and as required, to determine the pr esence of any unusual concentrations of airborne activity. Appr opriate respiratory protection, if required, will be indicated on the radiation work permits.
While work is being performed on the reac tor vessel head and there is a potential for release of airborne activity , conti nuous radiation protecti on monitoring of the work area will be in effect. This monitoring will consist of pa rticulate and iodine sampling. Workers will be instructed to leave the area if they suspect any unusual problem.The spent fuel pool exhaust monitor is a continuous sampling system that takes suction upstream to the HEPA filter. This sa mple passes through a particulate filter, charcoal cartridge and gaseous 4 pi beta, gamma detection chamber. The particulate filter is changed on a routine schedule to prevent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with MPS2 UFSAR11.E-2Rev. 35the particulate filter. Th e removed filter and cartridge may be counted in accordance with approved station procedures, if needed, to evaluate the airborne radioactivity concentration during the co llection period. The gaseous channel is monitored and alarmed in the control room. An alarm condi tion will require notification of radiation protection to determine the source of the increased
activity. Work area ai r sampling will be established co nsistent with spent fuel pool ventilation line up and work in progress. These samples will supplement the spent fuel pool exhaust monitor. During periods when the spent fuel pool exhaust or the containment atmosphere monitors are not available, air samples are taken with portable sampling equipment in these area s on a scheduled basis per the Millstone Effluent Control Program to determine the presence of any unusual concentrations of airborne activity.2.The area with the highest potential for creating airborne activity is the sampling room. All other radwaste area s contain equipment that is closed and therefore of a low potential for creating airborne activity. The sampling room (primary sample sink ar ea) has a separate ventilation system via the sample hood. The hood design is such that at least 100 linear feet per minute flow is maintained ove r the working area in the h ood to ensure that there is no contamination or airborne activity sp read into the sampling room. Visual indication is available to the personnel fr equenting the sampling room to indicate the condition of the exhaust system. A shutdown exhaust system would require health physics evaluation of airborne conditions prior to entry. 3.The radwaste airborne radioactivity monitoring system consists of four radiation monitors, each capable of continuously monitoring airborne radioactive particulates. In addition, each monitor c ontains an iodine sampling assembly consisting of a replaceable charcoal cart ridge mounted in se ries and downstream of the particulate monitor. The particulate filter is cha nged on a routine schedule to prevent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with the particulate filter. The removed filter and cartridge may be counted in accordance with approved station procedures, if needed, to evaluate the airborne radioa ctivity concentration during the collection period.Since the waste gas system of th e radwaste areas has the potential for the release of gaseous activity, the airborne radiation monitor sampling the ventilation exhaust duct servicing that area also contains a ga seous radioactivity monitor in series with the particulate and iodine monitor. The gase ous monitor will be used to detect and measure significant noble gas releas es from the waste gas system. Sampling for each monitor is accomplished by extracting a representative sample from the radwaste ventilation exhaus t system by using an isokinetic nozzle designed in accordance with ANSI N13.1-1969, Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities.
MPS2 UFSAR11.E-3Rev. 35 Since the airflow at each point select ed represents the exhaust of several compartments, the effect of dilution on the compartment having the least flow was analyzed to demonstrate the capability of detection and measurement of airborne contaminants in the areas being served by that section of the ventilation exhaust system. For this analysis, a concentration of 3x10
-9 microcuries per cubic centimeter was used. To determine the radioactive concentr ation at the sampling point, this value was multiplied by the ratio of the flow for the compartment having the lowest airflow to the total flow at the sample point.Point A (RM-8999 P&ID 25203-26029)The airflow at this sampling point is 15,495 scfm. The ventilation exhaust system at this point services the storage area, the of fices, the heat exchanger pump area, the coolant waste tank areas, and the evaporator rooms.
The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 800 scfm. Assuming the concentration of 3x10-9 microcuries per cubic centimeter exists in this room with the least flow, the radioactive concentration C A at this point is:Manufacturer's data for the radiation monitoring equipment states that the monitoring equipment placed in a 1 mrem/hr background of 1 Mev gamma ener gies produces a count rate of 78 cpm. Concentrations of 1x10-11µCi/cc (limiting isotope Cs-137) pr oduces a count rate 107 cpm.
Therefore, the count rate for the concentration at poi nt A is equal to 1656 cpm. Point B (RM-8998 P&ID 25203-26029)The airflow at this sampling point is 18,335 scfm. The ventilation exhaust system
at this point services the sampling room , the letdown heat exchanger room, the coolant waste receiver area, and the volume control tank area.
The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 160 scfm. Assuming the value of 3x10
-9µCi/cc exists in this area, the radioactive concentration C B at this point is:
C A310 9-xµCicc800scfm15495scfm,---------------------------------1.54810 10-xµCicc==
MPS2 UFSAR11.E-4Rev. 35 Assuming the 78 cpm for the background counting rate and the 107 cpm for the concentration 1x10-11µCi/cc (Cs-137), the count rate for point B is 279 cpm. Point C (RM-8997 P&ID 25203-26029)The airflow at this sampling point is 10,170 scfm. The ventilation exhaust system at this point services the char ging pum p areas, the degasifier areas, and the engineered safety features areas. The ai rflow for the compartme nt having the least flow exhausting into this section of th e ventilation exhaust sy stem is 235 scfm.
Assuming the value of 3x10
-9 microcuries per cubic centimeter exists in this area, the radioactive concentration C C , at this point is:
Assuming the 78 cpm for the background count rate and the 107 cpm for the concentration 1 x 10-11µCi/cc (Cs-137), the count rate for the concentration at point C is 742 cpm. Point D (RM-8434 A&B P&ID 25203-26029)
The air flow at this sampling point is 17,920 scfm. The ventilation system at this point services the equipment laydown area, the waste gas decay system areas, the aerated waste gas s ystem area, and the RBCCW heat exchanger area. The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 800 scfm. Assuming the value of 3x10
-9µCi/cc exists in this area, the radioactive concentration, C D , at this point D is:
C B310 9-xµCicc160scfm18335scfm,------------------
-2.6110 11-xµCicc==C C310 9-xµCicc235scfm10170scfm,--------------------------------
-6.9310 11-xµCicc==
MPS2 UFSAR11.E-5Rev. 35 Assuming the 78 cpm for the background count rate and 107 cpm for the concentration 1 x 10-11µCi/cc (Cs-137), the count rate for the concentration at point D is 1434 cpm.
The calculations and analysis for the f our sampling points show n above indicate a count rate ranging from 279 counts per minute to 1656 counts per minute. Since these count rates are greater than three times the background rate, as specified by the manufacturer, the concentration shown in the above calculations can be detected and alarmed.
The above analysis is conservative in th at the value used is based on an exposure to the concentrations specified for forty hours in any period of seven consecutive days. However, since the normal occupancy for these areas is significantly less than this 40-hour period, the exposure of personnel to airborne radioactivity will be considerably less than that stated in the above calculations. In conclusion, the above analysis clearly demonstrates the capability of the airborne radiation monito ring system to detect and measure appropriate concentrations in the radwas te ventilation exhaust system and thus to comply with the requirements for personnel safety stated in 10 CFR 20.103 and 10 CFR 20.203(d).
C D310 9-xµCicc800scfm17920scfm,---------------------------------1.3410 10-xµCicc==