ML17212A081

Final Safety Analysis Report, Rev. 30, Chapter 11, Radioactive Waste Management

ML17212A081
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Site:	Millstone
Issue date:	06/29/2017
From:	Dominion Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
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MPS-3 FSARMillstone Power Station Unit 3 Safety Analysis Report Chapter 11 MPS-3 FSAR 11-i Rev. 30CHAPTER 11 - RADIOACTIVE WASTE MANAGEMENT Table of ContentsSection Title Page

11.0BACKGROUND

......................................................................................11.0-1 11.1SOURCE TERMS....................................................................................11.1-1 11.1.1RADIONUCLIDE INVENTORY IN THE CORE..................................11.1-111.1.2RADIONUCLIDE INVENTORY IN FUEL ELEMENT GAP...............11.1-211.1.3PRIMARY COOLANT EQUILIBRIUM ACTIVITIES..........................11.1-311.1.3.1Fission Product Activities.........................................................................11.1-311.1.3.2Tritium Activity........................................................................................11.1-4 11.1.3.3Corrosion Products...................................................................................11.1-711.1.3.4Nitrogen-16 Activity.................................................................................11.1-711.1.4RADIOACTIVITY IN THE SECONDARY SIDE..................................11.1-7 11.

1.5REFERENCES

FOR SECTION 11.1.......................................................11.1-811.2LIQUID WASTE MANAGEMENT SYSTEMS.....................................11.2-111.2.1DESIGN BASES......................................................................................11.2-111.2.2SYSTEM DESCRIPTION........................................................................11.2-311.2.2.1Radioactive Liquid Waste System (LWS)................................................11.2-311.2.2.2Condensate Demineralizer Liquid Waste System (LWC)........................11.2-511.2.2.3Other Systems Discharging Radioactive Liquid Waste............................11.2-611.2.3RADIOACTIVE RELEASES..................................................................11.2-711.2.3.1Radioactive Liquid Waste System Leak or Failure (Atmospheric Release)..............................................................................11.2-711.2.3.2Liquid Containing Tank Failure...............................................................11.2-811.2.4REFERENCE FOR SECTION 11.2.........................................................11.2-811.3GASEOUS WASTE MANAGEMENT SYSTEMS................................11.3-111.3.1DESIGN BASES......................................................................................11.3-111.3.1.1Design Objective.......................................................................................11.3-111.3.1.2Design Criteria..........................................................................................11.3-211.3.1.3Cost Benefit Evaluation............................................................................11.3-311.3.1.4Equipment Design Criteria.......................................................................11.3-311.3.1.5Building Ventilation Systems...................................................................11.3-411.3.2SYSTEM DESCRIPTIONS.....................................................................11.3-411.3.2.1Radioactivity Inputs and Release Points...................................................11.3-411.3.2.2Degasifier Subsystem of Radioactive Gaseous Waste System.................11.3-511.3.2.3Process Gas Subsystem of Radioactive Gaseous Waste System.......................................................................................................11.3-6 MPS-3 FSARCHAPTER 11 -RADIOACTIVE WASTE MANAGEMENT Table of Contents (Continued)

Section Title Page 11-ii Rev. 3011.3.2.4Process Vent Portion of Radioactive Gaseous Waste System..................11.3-611.3.2.5Steam and Power Conversion System......................................................11.3-711.3.2.6System Instrumentation Requirements.....................................................11.3-7 11.3.2.6.1Radioactive Gaseous Waste System.........................................................11.3-711.3.2.6.2Ventilation Systems..................................................................................11.3-811.3.2.7Seismic Design Provisions of the Radioactive Gaseous Waste System............................................................................................11.3-811.3.2.8Quality Control.........................................................................................11.3-811.3.2.9Welding.....................................................................................................11.3-9 11.3.2.10Materials...................................................................................................11.3-911.3.2.11Construction of Process Systems..............................................................11.3-911.3.2.12System Integrity Testing...........................................................................11.3-9 11.3.3RADIOACTIVE RELEASES................................................................11.3-1011.3.3.1Radioactive Gaseous Waste System Failure...........................................11.3-1111.3.4REFERENCE FOR SECTION 11.3.......................................................11.3-1211.4SOLID WASTE MANAGEMENT..........................................................11.4-111.4.1DESIGN BASES......................................................................................11.4-111.4.2SYSTEM DESCRIPTION........................................................................11.4-211.4.2.1System Inputs............................................................................................11.4-3 11.4.2.1.1Spent Resins..............................................................................................11.4-311.4.2.1.2Waste Evaporator Bottoms.......................................................................11.4-311.4.2.1.3Regenerant Chemical Eva porator Bottoms (Removed From Service).....................................................................................................11.4-311.4.2.1.4Boron Evaporator Bottoms.......................................................................11.4-311.4.2.1.5Miscellaneous Radioactive Solid Wastes.................................................11.4-311.4.2.2Equipment Description.............................................................................11.4-411.4.2.2.1Boron, Waste Evaporator Bottoms...........................................................11.4-511.4.2.2.2Spent Resin Handling...............................................................................11.4-511.4.2.2.3Filter Handling..........................................................................................11.4-611.4.2.2.4Incompressible Waste Handling...............................................................11.4-611.4.2.2.5Waste Compaction Operation...................................................................11.4-611.4.2.3Expected Volumes....................................................................................11.4-611.4.2.4Packaging..................................................................................................11.4-611.4.2.5Temporary On-site Storage Facilities.......................................................11.4-611.4.2.6Shipment...................................................................................................11.4-711.4.2.7Protection Against Unc ontrolled Releases...............................................11.4-7 MPS-3 FSARCHAPTER 11 -RADIOACTIVE WASTE MANAGEMENT Table of Contents (Continued)

Section Title Page 11-iii Rev. 3011.5PROCESS, EFFLUENT, AND AIRBORNE RADIATION MONITORING SYSTEMS......................................................................11.5-111.5.1DESIGN BASES......................................................................................11.5-111.5.2SYSTEM DESCRIPTION........................................................................11.5-211.5.2.1Instrumentation.........................................................................................11.5-2 11.5.2.2Process and Effluent Monitors..................................................................11.5-411.5.2.2.1Ventilation Vent Monitors-Normal Range.............................................11.5-411.5.2.2.2Ventilation Vent Monitor-High Range.....................................................11.5-411.5.2.2.3Hydrogenated Vent Monitor.....................................................................11.5-511.5.2.2.4Containment Fuel Drop Monitors.............................................................11.5-511.5.2.2.5Supplementary Leak Collection and Release System Monitor................11.5-511.5.2.2.6Condenser Air Ejector Monitor................................................................11.5-611.5.2.2.7Control Building Inlet Ventilation Monitors............................................11.5-611.5.2.2.8Hydrogen Recomb iner Ventilation Monitors...........................................11.5-611.5.2.2.9Normal Range Particulate and Gas Monitors...........................................11.5-711.5.2.2.10Main Steam Relief Line Monitors............................................................11.5-811.5.2.2.11Turbine Driven Auxiliary Feedwater Pump Discharge Monitor..............11.5-811.5.2.2.12Main Steam Line Monitor; N-16 and Fission Product.............................11.5-811.5.2.3Liquid Process Monitors...........................................................................11.5-911.5.2.3.1Containment Reci rculation Cooler Service Water Outlet Monitors....................................................................................................11.5-911.5.2.3.2Liquid Waste Monitor...............................................................................11.5-911.5.2.3.3Steam Generator Blowdown Sample Monitor..........................................11.5-911.5.2.3.4Auxiliary Condensate Monitor...............................................................11.5-1011.5.2.3.5Turbine Building Floor Drains Monitor.................................................11.5-1011.5.2.3.6Reactor Plant Component Cooling Water System Monitor....................11.5-1011.5.2.3.7Deleted by FSARCR 05-MP3-015.........................................................11.5-1111.5.2.3.8Regenerant Evaporator Monitor (Removed from Service).....................11.5-1111.5.2.3.9Waste Neutralization Sump Monitor......................................................11.5-1111.5.2.4Inservice Inspection, Calibration, and Maintenance...............................11.5-1111.5.2.5Sampling.................................................................................................11.5-1211.

5.3REFERENCES

FOR SECTION 11.5.....................................................11.5-12 MPS-3 FSAR 11-iv Rev. 30CHAPTER 11 - RADIOACTIVE WASTE MANAGEMENT List of Tables Number Title11.1-1Iodine and Noble Gas Inventory in Re actor Core - Original License Basis (1) (historical)11.1-2Reactor Coolant Equili brium Concentrations - Original License Basis (HISTORICAL)11.1-2ADesign Reactor Coolant Equilibrium Concentrations at 3723 MWt11.1-3Parameters Used in the Calculation of Reactor Coolant, Secondary Side Liquid, and Secondary Side Steam Fission and Ac tivation Product Activity - Original License Basis11.1-3AParameters Used in the Calculation of Design Reactor Fission and Activation Product Activity11.1-4Tritium Production 11.1-5Reactor Coolant N-16 Activity (1)11.1-6Secondary Side Liquid Equilibrium C oncentrations Original License Basis (HISTORICAL)11.1-7Secondary Side Steam Equilibrium Con centrations - Original License Basis (HISTORICAL)11.2-1Liquid Waste Management System Daily Input Flows11.2-2Liquid Waste Manageme nt System Design Data11.2-3Tank Overflow Protection 11.2-4Expected Radioactive Liquid Concentrations From Each Liquid Release Stream

(µCi/ml) (1) Following Treatment (HISTORICAL)11.2-5Expected Annual Radioactive Liquid Releases Prior to Dilution in the Circulating Water Discharge System a nd Prior to Inclusion of Anticipated Operational Occurrences (1) (HISTORICAL)11.2-6Expected Annual Radioactive Liquid Releases After Dilution in the Circulating Water Disharge System and Inclusion of Anticipated Operational Occurrences (1) (HISTORICAL)11.2-7Design (1) Radioactive Liquid Concentrati ons From Each Liquid Release Stream (µCi/gm) Following Treatment (2)11.2-8Design (1) Annual Radioactive Liquid Releases Prior to Addition of Anticipated Operational Occurances and Dilution in the Circulating Water Discharge System MPS-3 FSAR List of Tables (Continued)

Number Title 11-v Rev. 3011.2-9Design (1) Annual Radioactive Liquid Rele ases Following Addition of Anticipated Operational Occurances a nd Dilution in the circulating Water Discharge System11.2-10Fraction of MPC Released - Design Case (1) (HISTORICAL)11.2-11Assumptions Used for the Radioactive Liquid Waste System Failure (Release to Atmosphere) and for the Liquid Containing Tank Failure11.2-12Boron Recovery Tank Concentrations (µCi/cc)11.2-13Activity Released to Atmosphere fr om a Radioactive Li quid Containing Tank Failure (Boron Recovery Tank)11.2-14Radioactive Concentrations in Groundwater Entering Niantic Bay Following a Rupture of Boron Recovery Tank11.3-1Total Expected Radioactive Gaseous Released to Atmosphere from Millstone 3 (HISTORICAL)11.3-2Radioactive Gaseous Source Term Parameters (HISTORICAL)11.3-3Radioactive Gaseous Waste System 11.3-4Codes and Standards 11.3-5Expected Radioactive Gaseous Releases to Atmosphere via Ventilation Vent (HISTORICAL)11.3-6Expected Radioactive Gaseous Releases to Atmosphere from Millstone 3 via Millstone Stac k (HISTORICAL)11.3-7Expected Radioactive Gaseous Releases to Atmosphere via Turbine Building (HISTORICAL)11.3-8Design Radioactive Gaseous Releases to Atmosphere via Ventilation Vent (HISTORICAL)11.3-9Design Radioactive Gaseous Releases to Atmosphere from Millstone 3 via Millstone Stac k (HISTORICAL)11.3-10Design Radioactive Gaseous Releases to Atmosphere via Turbine Building Roof (HISTORICAL)11.3-11Design Radioactive Gaseous Releases to Atmosphere from Millstone 3 (HISTORICAL)11.3-12Assumptions Used for the Process Gas Charcoal Bed Adsorber Bypass Analysis (1)

MPS-3 FSAR List of Tables (Continued)

Number Title 11-vi Rev. 3011.3-13Radioisotope Releases from the Pr ocess Gas Charcoal Bed Adsorber and Associated Piping11.4-1Volume of Spent Resin Generated Annually 11.4-2Radioactive Solid Waste System Component Data 11.4-3Omitted 11.4-4Radioactive Solid Waste Annual Shipments (1)(2)11.5-1Gaseous Monitors 11.5-2Liquid Process Monitors 11.5-3Radiological Samples Taken at Reactor Plant Sample Sink 11A.1-1Annual Doses to Maximum Individua l in the Adult Group from Gaseous Effluents11A.1-2Annual Doses to Maximum Individual in the Teen Group from Gaseous Effluents11A.1-3Annual Doses to Maximum Individua l in the Child Group from Gaseous Effluents11A.1-4Annual Doses to Maximum Individua l in the Infant Group from Gaseous Effluents11A.1-5Annual Doses to Maximum Individua l in the Adult Group from Gaseous Effluents11A.1-6Annual Doses to Maximum Individual in the Teen Group from Gaseous Effluents11A.1-7Annual Doses to Maximum Individua l in the Child Group from Gaseous Effluents11A.1-8Annual Doses to Maximum Individua l in the Infant Group from Gaseous Effluents11A.1-9Omitted 11A.1-10Omitted11A.1-11Omitted11A.1-12Omitted 11A.1-13Annual Doses to Maximum Indivi dual in the Adult Group from Liquid Effluents MPS-3 FSAR List of Tables (Continued)

Number Title 11-vii Rev. 3011A.1-14Annual Doses to Maximum Individual in the Teen Group from Liquid Effluents11A.1-15Annual Doses to Maximum Individual in the Child Group from Liquid Effluents11A.1-16Comparison of Maximum Calculated Doses from Millstone 3 Nuclear Plant with Appendix I Design Objectives11A.1-17Calculated Population Dose 11A.2-1Dilution Factors, Travel Times from the Site, and Population Served11A.2-2Parameters and Assumptions Used in Equations for Estimating Doses to Humans11A.2-3Meteorogical Data 11A.2-4Parameters and Assumptions Us ed in Estimating Doses to Biota11A.3-1Base Case Annual Population Doses Due to Liquid Effluents (1)11A.3-2Base Case Annual Population Doses Due to Gaseous Effluents (1)

MPS-3 FSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

11-viii Rev. 30CHAPTER 11 - RADIOACTIVE WASTE MANAGEMENT List of Figures Number Title11.2-1Radioactive Liquid Waste and Aerated Drains11.2-2Condensate Demineralizer Liquid Waste 11.2-3Expected Radioactive Liquid Wa ste Source and Discharge Paths11.3-1Radioactive Gaseous Waste 11.3-2Ventilation System Composite Drawing Normal Operation11.4-1Radioactive Solid Waste 11.4-2(HISTORICAL) Radioactive Solid Waste System Exp ected Quantities11.4-3(HISTORICAL) Radioactive Soli d Waste System Design Quantities MPS3 UFSAR11.0-1Rev. 30CHAPTER 11 - RADIOACTIVE WASTE MANAGEMENT

11.0BACKGROUND

An estimate of the radioactive effluents and pu blic dose is provided that documents projected public dose consequences are within 10 CFR 50 Appendix I radioactive release criteria. The estimates were based on nominal assumptions a nd generic models and based on plant operations at a core power level of 3636 MWt. These dose estimates were developed in support of the original license and updated during MPS-3 restart.

These dose estimates are historical and not subject to future update. This information is re tained to avoid loss of original design basis.The Radiological Effluent Monitoring and Offsite Dose Calculat ion Manual (REMODCM) provide guidance requirements for system operation, dose cal culations, and monitoring requirements, and to ensure compliance with efflue nt limits. Actual measur ed concentrations of radioactivity released and real time dilution or dispersion es timates are required to verify compliance with effluent limits. Therefore, it is operation within the requirements of the REMODCM that ensures compliance with effluent limits, rather than operation to the nominal assumptions in this chapter.The Annual Radioactive Effluent Release Re port and Annual Radiological Environmental Operating Report, which are required to be s ubmitted in May of each year, document that operation of the plant complies with effluent limits and appropriate regulations. Technical Specifications requires impl ementation of the REMODCM an d Radiation Environmental Monitoring Program.

MPS3 UFSAR11.1-1Rev. 3011.1SOURCE TERMSThe source of all radioactivity occurring in the process streams of the various radioactive systems is the radionuclides generated in the reactor core and neutron activ ation of nuclides in the reactor coolant system (RCS) and the air surrounding the reactor vessel.

Radioactive liquid and gaseous releases are described in Sections 11.2 and 11.3, respectively. All reduction factors and decontaminati on factors associated with these systems are discussed in the appropriate sections.11.1.1RADIONUCLIDE INVENTORY IN THE CORE The discussion on core inventor y presented below represents in formation used by the original license application to establish plant shielding and radwaste effluent assessments. It is retained for historical purposes to avoid loss of original design basis and is not subject to future update. The core inventory used for accident analyses is presented in Chapter 15.

The specific activity of fission products in the core is calculated using the computer program ACTIVITY2. This program calculates the contribution from "parent," "daughter," and "granddaughter" isotopes by solving the following differential equations:1.First order nuclides: (11.1-1)2.Second order nuclides: (11.1-2)3.Third order nuclides: (11.1-3)where: i, j, k = indicate first, second , and third order nuclide parameters

= concentration of nuclide i per fu el region at time t (atoms/region)t = time (sec) dN ci t ()dt------------------Fii hii++()N ci t ()-=dN cj t ()dt------------------Fji f ij N ci t ()()j hjj++()N cj t ()-+=dN ck t ()dt-------------------Fki f ik N ci t()j f jk N cj t ()k hkk++()N ck t ()-++=N c i t ()

MPS3 UFSAR11.1-2Rev. 30 F = fission rate (fissions/sec in fuel region)i = fission yield for isotope i (atoms/fission)i = decay constant for isotope i (sec

-1)i = escape rate coefficient (sec

-1)i = = burnup rate (sec

-1)f ij = branching fraction from i to j h = fraction of failed fuel The program has a basic library of 167 nuclides wi th a capability of 200 nuclides. Library data include decay scheme information, production in formation, purification factors for typical demineralizers, and fuel escape rate coefficients. The library also contai ns decay gamma spectra in seven energy groups. Input data include time in tervals, initial source inventory in the fuel, neutron flux, power level, fracti on of fuel defects, and density of reactor coolant. The program describes the system analyzed, as well as the operating history, the activities, and associated gamma spectral information as a function of time. Fuel assembly source terms for shielding design, given in Chapter 12, are calculated by this method.

The calculation of the core iodi ne fission product invent ory is consistent with the inventories given in TID-14844 (DiNunno et al., 1962). The core iodine and noble gas fission product inventories in Table 11.1-1 are based on conti nuous operation of the unit at 3,636 MWt (design core output plus 2.0 percent inst rumentation error). The core inventory source terms used for Chapter 15 radiological accident analyses have been revised to reflect requirements as described by Regulatory Guide 1.183 and is discussed in Section 15.0.9.1.The core inventory information listed in Table 11.1-1 is being retained for historical purposes because that was the basis for original plan t shielding design as described in Section 12.2.1.Fuel element heat loadings and stresses as well as fuel operating experience are presented in Chapter 4.11.1.2RADIONUCLIDE INVENT ORY IN FUEL ELEMENT GAP The gap activity is that fraction of the gaseous activity in the core that diffuses to the fuel gaps. The fuel element gap source terms used for Chapte r 15 radiological accident analyses have been revised to reflect requirements as describe d by Regulatory Guide 1.183 and is discussed in Section 15.0.9.2.a ithi r a ith=

MPS3 UFSAR11.1-3Rev. 3011.1.3PRIMARY COOLANT EQUILIBRIUM ACTIVITIES11.1.3.1Fission Product Activities The design basis fission product activities in the reactor coolant resulti ng from fuel defects associated with 1-percent power are also calculated with the AC TIVITY2 program. The following differential equations are used:a.First order nuclides: (11.1-4)b.Second order nuclides: (11.1-5)c.Third order nuclides: (11.1-6)where: = concentration of nuclide i in the main coolant at time t (atoms/cm 3)n = total number of fuel regions V w = volume of main coolant (cm 3)PF EQ = equivalent purification factor (fraction) for ii = = Burnup rate (sec

-1)Q 1 = equivalent flow into purification stream (cm 3/sec)T 1 = coolant residence time in core (sec)

T 2 = coolant circul ation time (sec) dN wi dt------------

hni V w----------N ci t()i PF EQ i Q 1 V w---------------------i T 1 T 2------++N wi t ()-=dN wj dt------------

hnj V w----------

N cj t()i f ij N wi t ()j PF EQ j Q 1 V w---------------------j T 1 T 2------++N wj t ()-+=dN wk dt-------------

hnk V w-----------

N ck t()i f ik N wi t()j f jk N wj t ()k PF EQ k Q 1 V w------------

---

k T 1 T 2------++N wk t ()-++=N w i t ()a ith MPS3 UFSAR11.1-4Rev. 30The reactor coolant system (RCS) design basis equilibrium activities in Table 11.1-2 are based on the parameters listed in Table 11.1-3 and repres ent coolant data used by the original license application to establish plant shielding and radwaste effluent assessments. This data is retained for historical purposes to avoid loss of original design basis and is not subject to future update.In these calculations, the defective fuel rods were assumed to be present in the initial core and uniformly distributed throughout th e core. Thus, the fission product escape rate coefficients were based upon average fuel temperature. Calculations are performed using the average temperature of the reactor coolant. The reactor coolant density correction of 1.4 was made in order to obtain the correct radionuclides concentration downstream of the letdown heat exchanger.Also included in Table 11.1-2 are the expected equilibrium concentrati ons for the RCS. These results were based on measured and calculat ed concentrations gi ven in NUREG-0017 (USNRC 1976) and the parameters listed in Table 11.1-3.

The expected reactor coolant activities were used to develop the source terms for gaseous and liquid effluents in Sections 11.2 and 11.3.Both expected and design reactor coolant activities were used to evaluate the ventilation design in Chapter 12.

The methodology to calculate the current design basi s primary coolant activity concentrations is similar to that used during original licensing basis with a minor modification. The design basis core inventory is calculated by industry computer code ORIGEN instead of ACTIVITY2. Since the source of primary coolant fission product activ ity is the leakage of core activity via the defective fuels, the primary coolant activity concentrations calcul ated by ACTIVITY2 are adjusted by the ratio of the core inventory deve loped by ORIGEN, and presented in Chapter 15, to the core inventory calculated by ACTIVITY2. The coolant con centration for a given isotope depends on the core inventory, the escape coeffici ents of the isotope and its precursor isotopes, and the depletion rate of the isotopes. For each isotope in the primary coolant, the decay chain and the escape coefficients are examined, and thos e isotopes that have si gnificant impact on the concentration of the referenced isotope in the coolant are identified. Amongst these major contributors, the ratios of the ORIGEN core activity to the ACTIVITY2 core activity are generally very close to each other, and the ma ximum scaling factor is used to adjust the ACTIVITY2 based coolant concentration to reflect the core developed by ORIGEN.

The current design basis reactor coolant equilibrium activities presented in Table 11.1-2A, are based on parameters listed in Table 11.1-3A.11.1.3.2Tritium Activity There are two principal contribut ors to tritium production within the pressurized water reactor system: the ternary fission source and the dissolv ed boron in the reactor coolant. Additional contributions are made by Li-6, Li-7, and deuterium in the reactor water. Tritium is also produced by nuclear reactions with boron contained in burnable poison rods. Tr itium production from different sources is shown in Table 11.1-4.

MPS3 UFSAR11.1-5Rev. 30 Fission SourceThis tritium is formed within the fuel material and may:1.Remain in the fuel rod uranium matrix2.Diffuse into the cladding and become fixed there, as zirconium tritide3.Diffuse through the cladding and be released into the primary coolant, or 4.Be released to the coolant through micr oscopic cracks or failures in the fuel cladding Previous WNES fuel design has conservatively assumed that the ratio of fission tritium released into the coolant to the total fission tritium fo rmed was approximately 0.30 for Zircaloy clad fuel. The operating experience at the R.E. Ginna Plant of the Rochester Gas and Electric Company, and at other operating pressurized water reactors using zircaloy clad fu el, has shown that the tritium release through the Zircaloy fuel cladding is less than the earlier estimates. Consequently, the release fraction has been re vised downward from 30 to 10 percent based on these data (WCAP-8253 1974).

Boric Acid SourceA direct contribution to the reactor coolant tritium concentration is made by neutron reaction with the boron in solution. The concentration of boric acid varies with core life and load follow so that this is a steadily decreasing s ource during core life. The principal boron reactions are B-10(n,2) H-3 and B-10(n,)Li-7(n,n)H-3 reactions. The Li-7 reaction is controlled by limit ing the overall lithium concentration to appr oximately 2 ppm during operation.

Burnable Poison Rod Source These rods are in the core only during the fi rst operating cycle and their potential tritium contribution is only during this period.

Lithium and Deuterium Lithium and deuterium reactions contribute only minor quantities to the tritium inventory (Table 11.1-4). These sources are due to the activation of the lithium and deuterium in the RCS as they pass through the reactor. Lithium-6 is essentially excluded from the system by using 99.9 percent Li-7.Design Bases The design intent is to reduce the tritium sources in the RCS to a practical minimum to permit longer retention of the reactor coolant within the plant without compromising operator exposures. Reduction of source terms is provided by using hafnium or Ag-In-Cd control rods instead of B4C MPS3 UFSAR11.1-6Rev. 30 and the determination that the quantity of tritium released from the fuel rods fabricated with zirconium alloy cladding is less than originally expected.

Design EvaluationTable 11.1-4 compares a typical de sign basis tritium production which has been used in the past to establish system and operational requirements of the plant and present expected values. There are two principal contributors to the expected tritium release to th e RCS: ternary fission source and the dissolved boron in the reactor coolant.

Because of the importance of the ternary fission source on the operation of the plant, WNES has been closely following operating plant data at the R. E. Ginna Plant. The R. E. Ginna Plant has a Zircaloy clad core with silver-indium-cadmium control rods. The operating levels of boron concentration during the startup of the plant are approximately 1,100 to 1,200 ppm of boron. In addition, burnable poison rods in the core contain boron which cont ributes some tritium to the coolant, but only during the first cycle. Data duri ng the operation of the plant have very clearly indicated that the present design sources were conservative. The tr itium released is essentially from the boron dissolved in the coolant and a ternary fission source which is less than 10 percent.

In addition to these data, other operating plants with Zircaloy clad cores have also reported low tritium concentrations in the RCS af ter considerable periods of operation.

This quantity of tritium becomes uniformly dist ributed in the RCS, the primary grade water system, and the boron recovery system. During refueling operations, the trit ium is further diluted when the refueling canal is filled from the borated refueling water storage tank.

The expected tritium concentration in the prim ary coolant is based on the value provided in NUREG-0017 (USNRC 1976). For radio active liquid waste analysis, it is assumed that 50 percent of the activity produced and entering into the coolant in 1 year is released in the radioactive liquid waste system effluent. The remaining 50 percent is released in the radioactive gaseous waste system and ventilation effluents.The design tritium concentration in the primary coolant is selected to allow limited access to the containment during normal operation. The wate r management plan controls tritium concentrations to design levels and also al lows for continuous c ontainment access during refueling with operation of the containment ventilation system.

Based on the above, the following conclusions have been reached:1.The tritium levels in plants operating with Zircaloy clad cores are lower than previous design predictions.2.The tritium source at fu ll power operation is reduced by using hafnium control rods.13-5 MPS3 UFSAR11.1-7Rev. 3011.1.3.3Corrosion Products Corrosion products in the reactor coolant become activated when they pa ss through the core. The most important corrosion products are Cr-51, Mn-54, Fe-55, Fe-59, Co-58, and Co-60. The corrosion product activity is dependent on many fa ctors, including the type of plant and the materials of construction. The mass transport process is complex and stochast ic, and calculational methods to predict corrosion product activity acc urately have not been successfully correlated with operational data (Bartlett 1969). Analytical predictions of the corr osion product activity levels are approximations. Therefore, design co rrosion product levels ar e assumed to be the values measured at operating reactors. The design corrosion product levels are based on NUREG-0017 (USNRC 1976) values modified by th e appropriate adjustment factor described therein.The corrosion product activities in the reactor coolant are given in Table 11.1-2, for the original license and Table 11.1-2A for current conditions.11.1.3.4Nitrogen-16 ActivityNitrogen-16 is a concern only during reactor operation because of its short half-l ife (7.1 seconds).

Nitrogen-16 is produced in circul ating primary coolant entering th e core region and irradiated by neutrons. Reactions with all three oxygen isot opes 0-16 (99.76 percent), 0-17 (0.037 percent), and 0-18 (0.204 percent) result in the production of N-16.Nitrogen-16 emits high energy gamm as in 75-percent of the disi ntegrations (70-percent at 6.13 MeV and 5 percent at 7.11 MeV).

The N-16 activity at various points in the RCS is given in Table 11.1-5.11.1.4RADIOACTIVITY IN THE SECONDARY SIDE The concentrations of principal radioisotopes in the secondary si de of the steam generators are listed for both the design and expected cases in Table 11.1-6 for liquid and Table 11.1-7 for steam.

These tables present secondary c oolant and steam data used by the original license application to establish plant shielding and radwaste effluent assessments. This da ta is retained for historical purposes to avoid loss of original design basis and is not subject to fu ture update.The design results for fission and activation products, based on parameters in Table 11.1-3, were calculated with the computer program IONEXCHANGER which solves the following differential equations for secondary liquid activities.1.First order nuclides: (11.1-7)2.Second order nuclides:

dN i dt---------R ii Q B V-------+N i t ()-=

MPS3 UFSAR11.1-8Rev. 30 (11.1-8)3.Third order nuclides: (11.1-9)where: N i = number of atoms of nuclide i (atoms)

R i = feed rate of nuclide i (atoms/sec)i = radioactive decay constant for nuclide i (sec

-1)f ij = branching fraction from i to j Q B = steam generator radio activity removal rate (cm 3/sec)V = volume of steam generator liquid (cm 3)t = time in seconds Secondary side steam activ ities are obtained by using the following relationship: (11.1-10) where: A i = steam equilibrium activity for isotope i (µCi/gm) = liquid equilibrium activity for isotope i (µCi/gm)The expected secondary liquid a nd steam activities are based on th e concentrations reported in NUREG-0017 and the parameters listed in Table 11.1-3.11.

1.5REFERENCES

FOR SECTION 11.1 11.1-1Barlett, J.W. 1969. Stochastics of Coolant Crud. In: ANS Transactions, Vol 12.11.1-2DiNunno, J.J.; Anderson, F.D.; Baker, R.E.; and Waterfield, R.L. 1962. Calculation of Distance Factors for Power and Test Reactor Sites. TID 14844, U.S. Atomic Energy Commission, U.S. National Technical Information Service, Springfield, Va.11.1-3U.S. Nuclear Regulatory Commission 1976. Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE code), NUREG-0017. U.S. National Technical Information Service, Springfield, Va.

dN i R ji f ij N j t()j Q B V-------+N j t ()-+=dN k dt----------R ki f ik N i t()j f jk N j t()k Q B V-------+N k t ()-++=A i P i A oi=A oi MPS3 UFSAR11.1-9Rev. 3011.1-4WCAP-8253 1974, "Source Term Data for Westinghouse Pressurized Water Reactors," Westinghouse Electric Corporation, Pittsburgh, Penn.

MPS-3 FSARPage 1 of 1Rev. 30

.NOTES:(1) Based on 650 days of operation at 3,636 MWt.

Historical, not subjec t to future updating. This table has been retained to preserve original license basis.TABLE 11.1-1IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE - ORIGINAL LICENSE BASIS (1) (HISTORICAL) IsotopeCore (Ci)I-1319.1E+07I-1321.3E+08I-1332.0E+08 I-1342.4E+08I-1351.9E+08Kr-83m1.6E+07 Kr-85m4.0E+07Kr-858.8E+05Kr-877.7E+07 Kr-881.1E+08Kr-891.4E+08Xe-131m8.0E+04 Xe-133m4.9E+06Xe-332.0E+08Xe-135m5.5E+07 Xe-1355.4E+07Xe-1371.8E+08Xe-1381.8E+08 MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.1-2REACTOR COOLANT EQ UILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (µCi/g)Expected (µCi/g)Noble GasesKr83m 4.5E-01 (1)2.2E-02Kr85m1.7E+009.2E-02Kr853.4E-022.1E-03Kr871.2E+006.6E-02 Kr883.4E+001.9E-01Kr891.1E-016.2E-03Xe131m1.1E-025.5E-03 Xe133m6.1E-014.1E-02Xe1332.6E+011.7E+00Xe135m1.2E+001.6E-02 Xe1355.1E+002.2E-01Xe1371.7E-011.1E-02Xe1386.0E-015.4E-02 Subtotal:4.1E+012.4E+00 HalogensBr839.0E-025.8E-03Br844.1E-023.2E-03Br855.7E-033.8E-04I1307.2E-032.4E-03 I1312.6E+002.9E-01I1329.3E-011.2E-01I1334.2E+004.3E-01 I1345.8E-015.8E-02I1352.2E+002.2E-01Subtotal:1.1E+011.1E+00 MPS-3 FSARPage 2 of 3Rev. 30 Corrosion ProductsCr512.0E-032.0E-03Mn543.3E-043.3E-04 Fe551.7E-031.7E-03Fe591.1E-031.1E-03Co581.7E-021.7E-02 Co602.1E-032.1E-03Subtotal:2.4E-022.4E-02Other NuclidesRb862.7E-049.1E-05Rb883.4E+002.5E-01Sr894.3E-033.7E-04 Sr901.7E-041.0E-05Sr912.0E-037.5E-04Y902.1E-041.3E-06 Y91m1.1E-034.5E-04Y916.9E-046.7E-05Y933.4E-043.9E-05 Zr957.1E-046.3E-05Nb957.4E-045.3E-05Mo993.4E+009.0E-02 Tc99m1.9E+005.6E-02Ru1033.4E-044.7E-05Ru1063.3E-051.0E-05 Rh103m3.4E-045.6E-05Rh1063.3E-051.3E-05Te125m9.0E-053.0E-05Te127m2.1E-03 2.9E-04Te1271.1E-039.8E-04TABLE 11.1-2REACTOR COOLANT EQ UILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (µCi/g)Expected (µCi/g)

MPS-3 FSARPage 3 of 3Rev. 30NOTE:(1) 4.5E-01 = 4.5 x 10

-1Historical, not subject to future updating. This table has been retained to preserve original license basis.Te129m3.9E-021.5E-03Te1292.2E-022.0E-03Te131m2.3E-022.7E-03 Te1311.1E-021.4E-03Te1322.7E-012.9E-02Cs1343.3E-012.7E-02 Cs1361.7E-011.4E-02Cs1371.7E+001.9E-02Ba137m1.5E+002.0E-02 Ba1404.4E-032.3E-04La1401.5E-031.6E-04Ce1417.0E-047.4E-05 Ce1435.2E-044.4E-05Ce1445.0E-043.5E-05Pr1436.8E-045.3E-05 Pr1445.0E-044.1E-05Np2393.9E-031.3E-03Subtotal:1.4E+015.2E-01 H-33.5E+001.0E+00Total (Excluding H-3)5.8E+014.0E+00Total (Including H-3)6.2E+015.0E+00TABLE 11.1-2REACTOR COOLANT EQ UILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (µCi/g)Expected (µCi/g)

MPS-3 FSARPage 1 of 5Rev. 30TABLE 11.1-2ADESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MW tNuclidePrimary Coolant Activity Concentration (µCi/g)Kr-83m3.53E-01Kr-85m1.09E+00Kr-851.33E-01 Kr-878.49E-01Kr-882.22E+00Kr-896.96E-02 Xe-131m1.93E-01Xe-133m7.68E-01Xe-1332.54E+01 Xe-135m9.45E-01Xe-1355.50E+00Xe-1371.94E-01 Xe-1386.54E-01Br-837.05E-02Br-843.50E-02 Br-853.68E-03Br-871.89E-03I-1291.79E-07 I-1304.39E-02I-1312.67E+00I-1321.09E+00 I-1334.06E+00I-1346.19E-01I-1352.39E+00 I-1366.73E-03Se-815.84E-07Se-837.94E-07Se-844.60E-07 MPS-3 FSARPage 2 of 5Rev. 30Rb-861.46E-01Rb-882.32E+00Rb-891.45E-01 Rb-901.12E-02Rb-915.70E-03Rb-923.88E-04 Sr-893.02E-03Sr-901.96E-04Sr-911.30E-03 Sr-929.45E-04Sr-934.41E-05Sr-947.49E-06 Y-903.84E-04Y-91m7.77E-04Y-911.43E-02 Y-921.12E-03Y-936.13E-04Y-942.72E-05 Y-951.14E-05Zr-955.72E-04Zr-973.67E-04 Nb-95m6.58E-06Nb-955.78E-04Nb-97m3.48E-04 Nb-973.91E-04Mo-995.27E+00Mo-1012.10E-02Mo-1021.52E-02Mo-1057.22E-04TABLE 11.1-2ADESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MW tNuclidePrimary Coolant Activity Concentration (µCi/g)

MPS-3 FSARPage 3 of 5Rev. 30Tc-99m2.73E+00Tc-1012.05E-02Tc-1021.53E-02 Tc-1057.66E-04Ru-1035.49E-04Ru-1051.34E-04 Ru-1062.02E-04Ru-1071.90E-06Rh-103m5.50E-04 Rh-105m3.82E-05Rh-1053.44E-04Rh-1062.25E-04 Rh-1071.14E-05Sn-1272.41E-06Sn-1285.04E-06 Sn-1308.31E-07Sb-1272.77E-05Sb-1282.66E-06 Sb-1293.73E-05Sb-1302.67E-06Sb-1311.16E-05 Sb-1328.85E-07Sb-1331.04E-06Te-125m3.98E-04 Te-127m3.11E-03Te-1271.10E-02Te-129m1.32E-02Te-1291.35E-02Te-131m3.34E-02TABLE 11.1-2ADESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MW tNuclidePrimary Coolant Activity Concentration (µCi/g)

MPS-3 FSARPage 4 of 5Rev. 30Te-1311.26E-02Te-1322.77E-01Te-133m1.91E-02 Te-1338.60E-03Te-1342.91E-02Cs-134m4.72E-02 Cs-1342.31E+01Cs-1363.52E+00Cs-1371.62E+01 Cs-1381.00E+00Cs-1398.99E-02Cs-1409.09E-03 Cs-1421.08E-04Ba-137m1.52E+01Ba-1397.79E-02 Ba-1403.75E-03Ba-1411.21E-04Ba-1421.66E-04 La-1401.26E-03La-1412.61E-04La-1422.44E-04 La-1431.38E-05Ce-1415.65E-04Ce-1434.19E-04 Ce-1444.46E-04Ce-1452.09E-06Ce-1467.74E-06Pr-1435.20E-04Pr-1444.50E-04TABLE 11.1-2ADESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MW tNuclidePrimary Coolant Activity Concentration (µCi/g)

MPS-3 FSARPage 5 of 5Rev. 30Pr-1451.49E-04Pr-1462.04E-05Nd-1472.22E-04 Nd-1492.28E-05Nd-1511.66E-06Pm-1471.13E-04 Pm-1491.97E-04Pm-1515.48E-05Sm-1517.50E-07 Sm-1531.55E-04Cr-512.00E-03Mn-543.30E-04 Fe-551.70E-03Fe-591.10E-03Co-581.70E-02 Co-602.10E-03Np-2391.98E-02H-33.5E+00TABLE 11.1-2ADESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MW tNuclidePrimary Coolant Activity Concentration (µCi/g)

MPS-3 FSARPage 1 of 5Rev. 30TABLE 11.1-3PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT, SECONDARY SIDE LIQUID, AND SECONDARY SIDE STEAM FISSION AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS ParameterValue (1) Core thermal power (includes 2

% instrumentation error) 3,636 MWtPercent fuel defects (design case)1.0 Percent fuel defects (expected case)0.12Fission product escape rate coefficients Noble gas nuclides 6.5 x 10-8 (sec-1)Br, Rb, I, and Cs nuclides 1.3 x 10-8 (sec-1)Te nuclides 1.0 x 10-9 (sec-1)Mo nuclides 2.0 x 10-12 (sec-1)Sr and Ba nuclides 1.0 x 10-12 (sec-1)Y, La, Ce, Pr nuclides 1.6 x 10-12 (sec-1)Reactor coolant liquid mass (without pressurizer) 4.70 x 10 5 lb Reactor coolant liquid volume (without pressurizer) 10,920 ft 3 Reactor coolant liquid volume (with pressurizer) 12,000 ft 3Reactor coolant full power average temperature 590°FPurification flow rate, normal75 gpm Mixed bed demineralizer decontamination factorsNoble Gases, N-16, H-31.0 Cations Design: Cs, Mo, Y1.0 Expected: Cs, Rb2.0 All other nuclides including activation products 10.0 MPS-3 FSARPage 2 of 5Rev. 30 Cation bed demineralizer decontamination factorsNoble Gases, N-16, Halogens, H-31.0 Cations Design: Cs, Y, Mo10.0 Expected: Cs, Rb10.0 All other nuclides including activation products Design1.0 Expected10.0 Expected halogens10.0 Ratio of cation bed demineralizer flow to purification bed demineralizer flow 0.1Reactor coolant letdown discharged via boron recovery systemDesign500 lb/hr Expected500 lb/hrSteam flow rate 1.589 x 10 7 lb/hr Primary to secondary leak rateDesign1,370 lb/dayExpected100 lb/daySteam generator partition factor (recirculating U-tube)Noble Gases: N-16, H-31.0 Halogens0.01 Cations Design: Cs, Rb0.0025 Expected: Cs, Rb0.001Others Design0.0025TABLE 11.1-3PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT, SECONDARY SIDE LIQUID, AND SECONDARY SIDE STEAM FISSION AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS ParameterValue (1)

MPS-3 FSARPage 3 of 5Rev. 30 Expected0.001 Condensate polishing demineralizer decontamination factors Cations Design: Cs, Y, Mo2.0 Expected: Cs, Rb2.0 All other nuclides in cluding activations productsDesign10.0Expected10.0 Condensate polishing flow rate 9.89 x 10 6 lb/hr Thermal neutron flux 2.0 x 10 13 n/cm 2-secOperating time (650 EFPD)15,600 hrCoolant cycle time9.9 sec Coolant in core time 0.7 secDegasification factor1.0 Secondary side equilibrium time 10 4 hrNumber of condensate demineralizers7 plus 1 spareVolume of each conde nsate demineralizer 197 ft 3Type of condensate demineralizersDeep bed Regeneration time for each demineralizerDesign3.5 days (estimate)Expected7.0 days Number of regenerations/yr for condensate polishing demineralizerDesign100 (estimate)Expected56Volume contro l tank volumesTABLE 11.1-3PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT, SECONDARY SIDE LIQUID, AND SECONDARY SIDE STEAM FISSION AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS ParameterValue (1)

MPS-3 FSARPage 4 of 5Rev. 30Vapor 240 ft 3 Liquid 160 ft 3Total secondary fluid per steam generatorExpected DesignLiquid99,253 lb103,000 lbSteam8,437 lb8,000 lb Total107,690 lb111,000 lbHypothetical Design flowrates for the steam generator blowdown system (see Section 11.2.2.3 for time periods for each release)Hot Standby:150,520 lb/hr 1% MSR from each steam generator (37,630 lb/hr per steam generator)

Intermittent Blowdown:263,410 lb/hr 1% MSR from three steam generators (37,630 lb/hr per steam generator) 4% MSR from one steam generator (150,520 lb/hr)

Hypothetical Expected flowrates for the steam generator blowdown system 1% MSR from each steam generator (37,000 lb/hr per steam generator) 148,000 lb/hr Fraction removed from steam generator blowdown (purificati on factors for design and expected cases)Noble Gases0.0Halogens0.8505 Cs, Rb0.4975Others0.8955Tritium0.0Ratio of condensate demineralizer flow rate to the total steam flow rate 0.6224TABLE 11.1-3PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT, SECONDARY SIDE LIQUID, AND SECONDARY SIDE STEAM FISSION AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS ParameterValue (1)

MPS-3 FSAR TABLE 11.1-3 (CONTINUED)

Page 5 of 5Rev. 30NOTES:(1) Values represent assumptions used to estima te liquid radiological ef fluents prior to initial plant licensing and are retained for historical purposes only. The Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) provides requirements for system operation, dose calculations and monitoring to ensure compliance with 10 CFR 20 Appendix B, Table II, Column 2 effluent limits.

MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.1-3APARAMETERS USED IN THE CALCULATION OF DESIGN REACTOR FISSION AND ACTIVATION PRODUCT ACTIVITYParameterValue Core thermal power (includes 2% instrumentation error)3,723 MWtPercent fuel defects1.0Fission product escape rate coefficients Noble gas nuclides 6.5 x 10-8 (sec-1)Br, Rb, I, and Cs nuclides 1.3 x 10-8 (sec-1)Te nuclides 1.0 x 10-9 (sec-1)Mo nuclides 2.0 x 10-12 (sec-1)Sr and Ba nuclides 1.0 x 10-11 (sec-1)Y, La, Ce, Pr nuclides 1.6 x 10-12 (sec-1)Reactor coolant average density 44.39 lbs/ft 3 Reactor coolant liquid volume (without pressurizer and surge line) 10,100 ft 3 Reactor coolant full power average temperature594.5

°FPurification flow rate, normal82 gpm

Mixed bed demineralizer decontamination factorsNoble Gases, N 16, H 31.0 CationsCs, Mo, Y1.0Halogens10.0 All other nuclides including activation products10.0 Cation bed demineralizer decontamination factorsNoble Gases, N 16, H 31.0 CationsCs, Mo, Y10.0Halogens1.0 All other nuclides including activation products1.0 MPS-3 FSARPage 2 of 2Rev. 30 Ratio of cation bed deminerali zer flow to purification bed demineralizer flow 0.01Reactor coolant letdown discharged via boron recovery system500 lb/hr Thermal neutron flux - core / coolant 3.83 x 10 13 n/cm 2 secReactor operating time (assumed 2 cycles)28,800 hr Coolant cycle time9.74 secCoolant in core time0.721 secDegasification factor1.0

Design corrosion product concentrations in RCSCr 512.0E-3

µCi/gmMn 543.3E-4

µCi/gmFe 551.7E-3

µCi/gmFe 591.1E-3

µCi/gmCo 581.7E-2

µCi/gmCo -602.1E-3

µCi/gmNp 2392.2E-3

µCi/gmDesign Tritium concentration3.5

µCi/gmTABLE 11.1-3APARAMETERS USED IN THE CALCULATION OF DESIGN REACTOR FISSION AND ACTIVATION PRODUCT ACTIVITYParameterValue MPS-3 FSARPage 1 of 1Rev. 30NOTES:Values presented are for a typical 3,565 MWt PWR. However, calculated tritium releases are based on 102 percent of this value (3,636 MWt) in order to comply with Regulatory Guide 1.49 (Section 1.8.1.49). Calculation of tritium releases are based on methodology presented in NUREG-0017.

Release fraction from fuel, 10 percent

Release fraction from burna ble poison rods, 10 percentWeight of boron-10 in bur nable poison rods, 6,160 gm Initial cycle boron, 900 ppm

Equilibrium cycle boron, 1,100 ppm

Lithium concentration (99.9 at om percent lithium-7), 2.2 ppm Initial cycle operating time, 9,240 effective full-power hours Equilibrium cycle operating time, 7,200 effective full-power hours Production in control rods is based on continuous daily load follow (12, 3, 6, 3 cycle). During base load full power operation, th e production would be negligible.TABLE 11.1-4TRITIUM PRODUCTION Tritium SourceTotal Produced (Ci/yr)Release Expected to Reactor Coolant (Ci/yr)Ternary fissionsInitial cycle14,0001,400Equilibrium cycle10,9001,090 Burnable poison rodsInitial cycle1,950195 Coolant (soluble boron)Initial cycle 388 388Equilibrium cycle 285 285 Coolant lithium, deuteriumInitial cycle141141Equilibrium cycle109109Total initial cycle16,4702,124 Total equilibrium cycle11,2941,484 MPS-3 FSARPage 1 of 1Rev. 30 Note:(1) These values are based on a typical 3,565Mwt PWR.TABLE 11.1-5REACTOR COOLANT N-16 ACTIVITY (1)Position in LoopLoop Transit Time (sec) N-16 Activity (µCi/g)Leaving core0.0189Leaving reactor vessel1.1170Entering steam generator1.4164 Leaving steam generator5.4112Entering reactor coolant pump6.0106Entering reactor vessel6.8 98 Entering core9.0 86Leaving core9.7189 MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.1-6SECONDARY SI DE LIQUID EQUILIBR IUM CONCENTRATIONS ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (1) (µCi/g)Expected (2) (µCi/g)Br 83 1.5E-05 (3)7.3E-08Br 843.1E-061.9E-08Br 855.2E-082.7E-10I 1301.7E-064.1E-08 I 1316.9E-045.4E-06I 1322.0E-042.0E-06I 1331.0E-037.7E-06 I 1346.3E-054.5E-07I 1354.9E-043.6E-06Cr 517.3E-075.6E-08 Mn 541.2E-071.2E-08Fe 556.2E-075.0E-08Fe 594.0E-073.8E-08 Co 586.2E-065.0E-07Co 607.6E-075.6E-08Rb 861.8E-074.6E-09 Rb 881.8E-041.0E-06Sr 891.6E-061.3E-08Sr 906.2E-082.5E-10 Sr 916.1E-061.6E-08Y 907.5E-085.3E-11Y 91m3.7E-071.2E-08 Y 912.5E-071.9E-09Y 93 1.1E-077.8E-10Zr 952.6E-072.5E-09Nb 952.7E-072.5E-09Mo 991.2E-032.6E-06 Tc 99m7.9E-042.5E-06 MPS-3 FSARPage 2 of 3Rev. 30Ru 1031.3E-071.3E-09Ru 1061.2E-082.5E-10Ru 103m1.2E-072.3E-09Rh 1061.2E-080.0Te 125m3.3E-086.3E-10 Te 127m7.7E-076.3E-09Te 1274.5E-072.4E-08Te 129m1.4E-053.8E-08 Te 1298.8E-066.7E-08Te 131m8.1E-066.9E-08Te 1312.2E-062.5E-08 Te 1329.6E-056.5E-07Cs 1342.2E-041.4E-06Cs 1361.1E-047.3E-07 Cs 1371.1E-031.0E-06Ba 137m5.7E-041.2E-06Ba 1401.6E-066.3E-09v La 1405.9E-074.7E-09Ce 1412.5E-072.5E-09Ce 1431.8E-076.8E-10 Ce 1441.8E-071.2E-09Pr 1432.5E-071.3E-09Pr 1441.8E-072.6E-09 Np 2391.4E-064.0E-08H 31.3E-031.0E-03Total (excluding H-3) 6.8E-033.1E-05Total (including H-3)8.0E-031.0E-03TABLE 11.1-6SECONDARY SI DE LIQUID EQUILIBR IUM CONCENTRATIONS ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (1) (µCi/g)Expected (2) (µCi/g)

MPS-3 FSARTABLE 11.1-6 (CONTINUED)

Page 3 of 3Rev. 30NOTES:(1) Based on 1,370 lb/day primary to secondary leak rate.(2) Based on 100 lb/day primary to secondary leak rate.(3) 1.5E-05 = 1.5 x 10

-5.Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.1-7SECONDARY SIDE STEA M EQUILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (1) (µCi/g)Expected (2) (µCi/g)Kr-83m 1.6E-06 (3)5.7E-09Kr-85m6.2E-062.4E-08Kr-851.2E-075.6E-10Kr-874.4E-061.7E-08 Kr-881.2E-054.9E-08Kr-893.8E-071.6E-09Xe-131m4.0E-081.5E-09 Xe-133m2.2E-061.1E-08Xe-1339.4E-054.4E-07Xe-135m4.1E-064.1E-09 Xe-1351.8E-055.6E-08Xe-1376.0E-072.9E-09Xe-1382.1E-061.4E-08 Br-831.5E-077.3E-10Br-843.1E-081.9E-10Br-855.2E-102.7E-12 I-1301.7E-084.1E-10I-1316.9E-065.4E-08I-1322.0E-062.0E-08 I-1331.0E-057.7E-08I-1346.3E-074.5E-09I-1354.9E-063.6E-08 Cr-511.8E-095.6E-11Mn-543.0E-101.2E-11Fe-551.6E-095.0E-11Fe-599.9E-103.8E-11Co-581.5E-085.0E-10 Co-601.9E-095.6E-11 MPS-3 FSARPage 2 of 3Rev. 30Rb-864.4E-104.8E-12Rb-884.6E-071.0E-09Sr-893.9E-091.3E-12Sr-902.5E-102.5E-13Sr-911.5E-091.6E-11 Y-901.9E-105.6E-14Y-91m9.3E-101.2E-11Y-916.3E-101.9E-12 Y-932.7E-107.8E-13Zr-956.5E-102.5E-12Nb-956.7E-102.5E-12 Mo-993.0E-062.6E-09Tc-99m2.0E-062.5E-09Ru-1031.1E-101.3E-12 Ru-1063.0E-112.5E-13Rh-103m2.9E-102.3E-12Rh-1063.0E-110.0 Te-125m8.2E-116.3E-13Te-127m1.9E-096.3E-12Te-1271.1E-092.4E-11 Te-129m3.6E-083.8E-11Te-1292.2E-086.7E-11Te-131m2.0E-086.9E-11 Te-1315.5E-092.5E-11Te-122.4E-076.5E-10Cs-1345.4E-071.4E-09Cs-1362.8E-077.3E-10Cs-1372.7E-061.0E-09 Ba-137m1.4E-061.2E-09TABLE 11.1-7SECONDARY SIDE STEA M EQUILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (1) (µCi/g)Expected (2) (µCi/g)

MPS-3 FSARPage 3 of 3Rev. 30NOTES:(1) Based on 1,370lb/day primary to secondary leak rate.(2) Based on 100lb/day primar y to secondary leak rate.(3) 1.6E-06 = 1.6 x 10

-6 Historical, not subject to future updating. This tabl e has been retained to preserve original design basis.Ba-1404.0E-096.3E-12La-1404.5E-094.7E-12Ce-1416.4E-102.5E-12Ce-1434.5E-106.8E-13Ce-1444.5E-101.2E-12 P-1436.2E-101.3E-12P-1444.5E-102.6E-12Np-2393.4E-094.0E-11 H-31.3E-031.0E-03Total (Excluding H 3): 1.8E-048.3E-07Total (Including H 3):1.5E-031.0E-03TABLE 11.1-7SECONDARY SIDE STEA M EQUILIBRIUM CONCENTRATIONS - ORIGINAL LICENSE BASIS (HISTORICAL)NuclideDesign (1) (µCi/g)Expected (2) (µCi/g)

MPS3 UFSAR11.2-1Rev. 3011.2LIQUID WASTE MANAGEMENT SYSTEMSIn accordance with General Design Criterion60, liquid waste management systems are provided to control, collect, process, store, recycle, and dispose of liquid radioactive waste generated as the result of normal plant operation, including anticipated operational occurrences. The liquid waste management systems include the radioactive liquid waste system (LWS) and condensate demineralizer liquid waste system (LWC) (which has been removed from service). Figures11.2-1 and 11.2-2 are the piping and instrumentation diagrams of the radioactive liquid waste system and condensate demineralizer liquid waste systems, respectively.The boron recovery system (Section9.3.5) also processes radioactiv e fluid for ultimate discharge from the plant. The radioactive waste handling aspects of this system are described in this section.

On occasion, the Unit will generate liquid radio active waste that cannot pr acticably 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 acco rdance with the Radiological Effluent Control Program outlined in the Administrative Section of the Technical Specifications (e.g., Unit 1 evaporator or shipped of f site for processing).The radioactivity values provided in this section are the design basis values used for the design of the liquid waste system. As such, they are considered historical and not subject to future updating.

The information is retained to avoid loss of th e original design bases.

Actual liquid radioactive release quantities can be found in the annual radioactive effluent release re ports as submitted to the NRC.11.2.1DESIGN BASES1.The design objectives of the li quid waste management systems are:a.To control the releases of radioactive materials within the limits set forth in 10CFR20 and to meet the numerical design objectives of AppendixI to 10CFR50. b.To meet the anticipated processing requirements of the plant. Adequate storage capacity is provided to hold liquid wastes during periods when major processing equipment may be down for maintenance or during periods of excessive waste generation.2.Table11.2-1 gives the daily input, in terms of average and peak flows, to the waste management systems. These values ar e based on the values in NUREG-0017, April 1976.3.Listings of expected and design case concentrations and annual quantities of radionuclides released to the plant discharge are given in Tables11.2-6 and 11.2-9.

MPS3 UFSAR11.2-2Rev. 304.Design data for components in the ra dioactive liquid waste system and the condensate demineralizer liquid waste system (removed from service) are given in Table11.2-2.5.The radioactive liquid waste (LWS) an d condensate demineralizer liquid waste systems (LWC) (removed from service) are designated no nnuclear safety (NNS).

Equipment is designed and fabricated in conformance with codes and standards identified in Regulatory Guide1.143 (Section1.8).6.The foundation and walls of the radi oactive liquid waste building and the condensate polishing facility building are seismically designed in conformance with Regulatory Guide1.143.7.A cost-benefit analysis for reducing cumulative dose to the population by using available technology has been perfor med in accordance with Regulatory Guide1.110 and is included in Appendix11A.8.General Design Criterion6 1 applies with regard to provisions for suitable shielding for radiation protection of personnel under nor mal and postulated accident conditions. Radiati on protection criteria for th e radioactive liquid waste system are given in Section12.2.9.Releases to the environment are monitored prior to discharge. Process and effluent radiological monitoring systems are described in Section11.5.10.The following design features are incor porated to reduce maintenance, equipment downtime, liquid leakage, and gaseous re leases to the building atmosphere, or otherwise improve ra dwaste operations:a.Dished, sloped, and conical bottoms ar e used in vessels and tanks with a potential for high activity and suspe nded solids to minimize buildup of radioactive sludge and facilitate cleaning.b.Pressure-retaining components of the system utilize welded construction to the maximum practicable extent. Flanged joints or suitable quick-disconnect fittings are used only where maintenance or operational requirements clearly indicate that such construction is pr eferable. Screwed connections in which threads provide the only seals are not used except for instrumentation connections where we lded connections are not suitable. Process lines are not less than 3/4-inch. Screwed connections backed up by seal welding, socket welding, or mechanical joints are used on lines greater than 3/4-inch but less than 2.5 inch nom inal size. For lines of 2.5 inches and above, piping is butt-welded (except as noted by FSAR Table 1.8-1).

Backing rings are not used in lines carrying resins or other particulate material. All welding constituting the pressure boundary of pressure-retaining components is performe d in accordance with ANSI B31.1.

MPS3 UFSAR11.2-3Rev. 30c.The piping systems are hydrostatically tested. Testing of piping systems is performed in accordance with ANSI B31.1.d.Pumps handling radioactive liquids are fitted with mechanical seals and outboard restriction bushings to minimize leakage. In the event of a seal failure, the leakage is directed to a radioactive sump through a drain connection.e.Piping is designed and valves are selected to minimize crud pockets where activity could accumulate.f.Tanks that are expected to contain li quids of high radioactivity are vented to the aerated vent system (Section 9.3.3) to minimize the potential for gaseous releases into working areas.g.A centralized control panel is pr ovided to allow system operation and monitoring from one location.h.Components handling highly radioactiv e liquids are separated by shield walls to minimize exposure to ope rators and maintenance personnel.i.Plastic pipes are not used for radioactive service.11.The design provisions to control radioact ive releases due to overflows from all liquid tanks containing potentially radioactive materials are shown in Table11.2-3.11.2.2SYSTEM DESCRIPTION 11.2.2.1Radioactive Liquid Waste System (LWS)

The radioactive liquid waste syst em consists of two separate, but interconnected, portions: The high-level waste portion and the low-level waste portion.High-Level Waste PortionTwo 26,000-gallon high-level waste drain tanks a ccept and store high-level radioactive liquid waste from the sources identified in Table11.2-1. Tank capacity allows time for recirculating and sampling of one tank while the other tank is being filled from any of the above sources. Two waste evaporator feed pumps serv ice either tank. The tanks are cr oss-connected to the low-level waste drain tanks (described below) at the discharge of the pumps.Two filters, located downstream of each waste evaporator feed pump, are av ailable to pre-filter high-level waste drain tank conten ts. These filters may be operated in parallel, in series, or bypassed when recirculating or processing tank contents of a high-level waste drain tank.

Subsequent to this flowpath, the effluent from the high-level wast e drain tank is processed in one MPS3 UFSAR11.2-4Rev. 30 of two paths: (1) through the high-level radioactive waste filter and demineralizer, or (2) in the waste evaporator subsystem.The waste evaporator, which was designed as an alternate path for proce ssing high level waste has been demonstrated by analysis to not be required to operate in order to meet 10CFR50 AppendixI Release Criteria.The waste evaporator is designed with an external reboiler, a large liquid disengaging space, a vapor-liquid separator, and a tray section. Thes e features combine to form a system with extremely high separation factors for nonvolatile nu clides. A decontamination factor of greater than 10 4 for nonvolatile nuclides is expected.Waste evaporator bottoms are allowed to concentrate until either approximately 15percent total dissolved and undissolved solids by weight have concentrated or an activity level to be determined by the characteristics of the container used to ship the evaporator bottoms offsite is accumulated. Evaporator bottoms are pumped to the radioactive solid waste system (Section11.4) via the waste bottoms holding tank.Effluent from the high level radioactive wast e demineralizer or distillate from the waste evaporator is collected in the waste test tanks. Samples of the liquid are analyzed for radioactivity and chemistry parameters. Depending on analysis results, the liquid is discharged to the circulating water discharge tunnel (Section10.4.5), or is capable of being recycled to the primary grade water system.If samples indicate that the distillate is unacceptable for reuse or discharge, it is either passed through a second waste demineralizer and resampled, or sent back to the high-level waste drain tanks for reprocessing. The second waste demineralizer is a mixed bed of ion exchange resins in the H and OH form. The resin is replaced when anal ysis of influent and ef fluent samples indicate that the decontamination factor becomes unaccep table or when the radiation level exceeds a predetermined limit.

It is expected that liquid from the waste test tanks is totally discharged. For the purpose of evaluating the radiological impact on the environm ent, one hundred percent of the input flow is assumed to be discharged. Assurance that wast e above activity limits is not inadvertently discharged to the environment is provided through sampling of the waste test tank effluent and by the radiation monitor in the discharge line. This monitor provides audible and visual alarms if activity levels in the effluent exceed limits. An air-operated valve in the discharge line is actuated to terminate the release.

Each batch is isotopically analyzed prior to release and the total activity discharged is recorded. Composite samples are retained in accordance with the procedures outlined in Regulatory Guide1.21. Detailed administrative records of al l radioactive liquid rel eases are maintained.

Minimum expected decontamination factors are shown on Figure11.2-3.

MPS3 UFSAR11.2-5Rev. 30Low-Level Waste Portion Two 4,000-gallon low-level waste drain tanks accept and store low level radioactive liquid waste from the reactor plant aerated drains system (Section9.3.3), as identified in Table11.2-1. These tanks and their respective pumps are arranged in the same manner as the high-level waste drain tanks.Turbine building floor and equipment drains are normally discharged directly to th e environment. When there is high radioactivity in the discharg e line, flow is diverted to the low-level waste system for further processing.

System design provides for conveying the contents of the low-level waste drain tanks to the high-level waste drain tanks if the acti vity level of the liquid in the lo w-level tanks is greater than a predetermined level. Normally, the low-level waste is sampled, analyzed and discharged to the circulating water discharg e tunnel. If the particulate concentration is above permit discharge levels, the contents of the lo w-level waste drain tanks will be recirculated through filter assemblies using the low-level waste drain pump(s) until the particulate c oncentration levels are acceptable. These filters are cartr idge-type filters provide d to remove particulate matter from the effluent. Filter elements are ch anged when the radioactivity leve l at the filter surface or the pressure drop across the filter exceeds a predetermined value.

Assurance that high-level radioact ive waste is not inadvertently discharged to the environment is provided through the analysis of low-level wast e drain tank liquid samples and the radiation monitor in the discharge line.

Minimum expected decontamination factors are shown on Figure11.2-3.11.2.2.2Condensate Demineralizer Liquid Waste System (LWC)The LWC has been removed from service and is no longer used.

This system has been isolated from the plant via locked closed valves, where possible. The system piping remains intact, such that if any leakage occu rs across the locked, closed valves, it will be contained within the existing system boundary. The system is interconnected to the Unit2 LWC at the regenerant evaporator feed tanks, regenerant evaporator feed pump suction and discharge lines, and at the system discharge downstream of the regenerant demi neralizer (originally designed to provide added system availability, capacity, and re dundancy). These interconnections are isolated also, with locked closed valves, where possible. Electrical power to the system components is administratively controlled by plant procedures.The LWC was originally designed and installed as an evaporator system to receive and process potentially radioactive liquid waste from the condensate demineralizer-mixed bed system (Section10.4.6). Evaporator bottoms were designed to be pumped to Unit2 for processing.

MPS3 UFSAR11.2-6Rev. 30 An evaluation has been performed which has determined that the LWC is not needed. This evaluation is documented by change to the Radiological Effluent Monitoring Offsite Dose Calculation Manual (ref.REMODCMCR#95-7).Analysis has shown that LWC is not required to be operated to meet 10CFR50 AppendixI release criteria.

For the purpose of evaluating the radiological impa ct on the environment, it is assumed that 100 percent of the regenerant chemical waste is discharged to the environment.11.2.2.3Other Systems Discharging Radioactive Liquid Waste The boron recovery system (Section9.3.5) receives degasified reactor coolant letdown and reactor plant gaseous drains flow, as identified on Figure11.2-3. For the purpose of evaluating the radiological impact on the environment, one hundred percent of the input flow is assumed to be discharged. Provisions made to protect against inadvertent discharg e are similar to those for high-level radioactive liquid waste, as the boron recovery system distillate is discharged to the liquid waste system upstream of the radiation monitor in the liquid waste system. Minimum expected decontamination factors are shown on Figure11.2-3.

During open-cycle blowdown, the steam generator blowdown (Section10.4.8) intermittently discharges directly to the circulating water discharge tunnel as part of the secondary water chemistry control program. For the purpose of evaluating the radiological impact on the environment, the following hypotheti cal cases have been developed.

For design base case:

• 100 percent discharge to the circulating water discharge tunnel.

• Four steam generators blow down at 1percent maximum steaming rate (MSR) for 14days per year during hot standby.

• One steam generator blows down intermittently at 4percent MSR, and the remaining three steam generators blow down at 1percent MSR for 5minutes per week per steam generator.

For expected case:

• 10 percent discharge to the circulating water discharge tunnel.

• Four steam generators each blowdown at 37,000 lb/hr for a total of 148,000 lb/hr.The actual radioactive releases will be reported annually.

MPS3 UFSAR11.2-7Rev. 3011.2.3RADIOACTIVE RELEASES Models and assumptions contained in NURE G-0017 are used to calc ulate the expected radioactivity concentrations in the liquid discharge. The concentration from each of the parent liquid waste streams following treatment are presented in Tables11.2-4 and 11.2-7 for the expected and design conditions, respectively. Liquid releases to the environment are listed in Tables11.2-5 and 11.2-6 for the expected nuclide c oncentrations prior and subsequent to dilution with the circulating water discharge system. A similar evaluation shown in Tables11.2-8 and 11.2-9 are for design nuclide concentrations prior a nd subsequent to diluti on with the circulating water discharge system. The diluted release for th e expected nuclide concentrations are further analyzed for the environmental impact as described in AppendixI of 10CFR50 and in Appendix11A of this FSAR. Table11.2-10 presents the design nucli de concentration releases to the unrestricted area in terms of fraction of ma ximum permissible concen tration (MPC) limits described in 10CFR20, AppendixB, TableII, Column2. The results indicate that the sum of the fractions of MPC values does not exceed the limits in 10CFR20. 11.2.3.1Radioactive Liquid Waste System L eak or Failure (Atmospheric Release)

The accident is defined as an unexpected and uncontrolled atmospheric release from the postulated rupture of a boron recovery tank. This tank is the highest potential atmospheric release source term because of its large volume, the re latively high potential for activity in streams feeding the tank, and its location in the yard area.The atmospheric release due to the postulated rupture of this tank is minimized by prior degasification and demineralization of letdown, which removes essentially all xenon and krypton and most of the iodine from the tank feed. The liquid released from the tank is held in a dike until drained or pumped, under administra tive control, to the liquid radioactive waste system for further processing.

The radiological consequences of a postulated radioactive liquid waste system failure resulting atmospheric release is reported in Table 15.0-8 based on design rele ase assumptions in Table 11.2-11, boron recovery tank concentrations in Table 11.2-12 and the X/Q values in Table 15.0-11. The resulting releases are listed in Table 11.2-13. The dose methodology is discussed in Appendix15A.In order to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric uncertainty), each of the iodine and noble gas isotopes released in Table 11.2-13 were scaled based on factors determined from the ratio of the primary activity concentrations (RCS concentration from Table 15.0-10 and Design concentration from Table 11.1-2) to determine the updated doses for the SPU operating condition.

The radiological consequences ar e consistent with the guideli nes of the pre-1991 version of 10CFR20, i.e., the whole body dose does not exceed 500 mRem to an individual at the nearest exclusion area boundary, and is substantially below the guidelines of 10CFR100.

MPS3 UFSAR11.2-8Rev. 3011.2.3.2Liquid Containing Tank FailureThis accident is defined as an unexpected a nd uncontrolled postulated rupture of the boron recovery tank. The boron recovery tanks are located in the yard area northeast of the containment structure (Figure 1.2-2). This area is provided with dikes to retain any liquid released from a tank rupture. This analysis assumes a combined rupt ure of a boron recovery tank and leakage into the groundwater.

Description of the analysis of this event is provided in Section 2.4.13.3.

The concentration of radionuclides in Niantic Ba y resulting from a liqui d containing tank failure are given in Table 11.2-14 based upon assumptions in Table 11.2-11.In order to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric uncertainty), each of the isotopes in Table 11.2-14 were scaled based on factors determined from the ratio of the primary activity concentrations (RCS concentration from Table 15.0-10 and Design concentration from Table 11.1-2) to determine the updated concentrations for the SPU operating condition.Concentrations in Niantic Bay are within the concentration of 10CFR20, Appendix B, Table II as they existed prior to the 1991 revision to 10CFR20 (see Section 2.4.13).11.2.4REFERENCE FOR SECTION 11.2 MPS-3 FSAR Page 1 of 1 Rev. 30TABLE 11.2-1LIQUID WASTE MANAGEM ENT SYSTEM DAILY INPUT FLOWS NUREG-0017 Liquid Waste Stream CategorySource Peak Flow Rate (gal/day) Average Flow Rate ( gal/day) Misc. Liquid Waste System Clean WastesHigh Level Waste Drain TanksContainment Building Sump30040 Auxiliary Building Sump1,500200Sample Fluids20035 Laboratory Waste500400 Misc. High Level Waste800660Dirty WastesLow Level Waste Drain TanksMisc. Low Level Waste65040Turbine Plant Floor Drains Turbine Plant Leakage7,200 Boron Recovery System Shim BleedBoron Recovery Tank1,440 Equipment DrainsBoron Recovery Tank300Secondary Waste System Steam Generator BlowdownSteam Generators426,411 Regenerant ChemicalsRegenerant Line3,400Detergent Wastes Laundry Facility450 MPS-3 FSARPage 1 of 9Rev. 30TABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATARadioactive Liquid Waste System Waste Evaporator Feed PumpsNumber2Capacity (gpm)35Design pressure (psig)350

Design temperature (°F)250Material of construction316 SSWaste Evaporator Reboiler PumpNumber1Capacity (gpm)4,000Design pressure (psig)600

Design temperature (°F)350Material of constructionAlloy 20Waste Evaporator Bottoms PumpNumber1Capacity (gpm)15Design pressure (psig)350

Design temperature (°F)250Material of constructionIncoloyWaste Distillate PumpNumber1Capacity (gpm)50Design pressure (psig)350

Design temperature (°F)250Material of construction316 SSWaste Test Tank PumpsNumber2Capacity (gpm)150 Design pressure (psig)350 Design temperature (°F)250 MPS-3 FSARPage 2 of 9Rev. 30Material of construction316 SSLow-Level Waste Drain PumpsNumber2Capacity (gpm)50Design pressure (psig)350

Design temperature (°F)250Material of construction316 SSWaste Bottoms Coolant PumpNumber1Capacity (gpm)120Design pressure (psig)350

Design temperature (°F)250Material of construction316 SSWaste Test Heating PumpsNumber2Capacity (gpm)60Design pressure (psig)350

Design temperature (°F)250Material of construction316 SSWaste Bottom Holding Tank PumpNumber1Capacity (gpm)50Design pressure (psig)250

Design temperature (°F)170Material of constructionAlloy 20Waste DemineralizerNumber1 Capacity (ft 3)35Design pressure (psig)150 Design temperature (°F)140Material of construction304 SSTABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 3 of 9Rev. 30High-Level Waste DemineralizerNumber1 Capacity (ft 3)35Design pressure (psig)140 Design temperature (°F)200Material of construction304 SSWaste Evaporator ReboilerNumber1Total duty (Btu/hr)26,200,000Shell Side Tube Side Total fluid entering29,9001,930,000Design pressure (psig)180100Operating pressure inlet (psig)10025 Temperature in/out (°F)338 / 338251.2 / 265.1 Material of constructionCarbon steelIncoloy 825Waste Evaporator Bottoms Cooler (3LWS-E2)Total duty (Btu/hr)637,500Shell Side Tube Side Total fluid entering (lb/hr)63,7507,500 Temperature in/out (°F)140 / 150255 / 170Design pressure (psig)210170Operating pressure inlet (psig)12550 Material of constructionCarbon steelCarpenter 20Waste Bottoms Coolant Preheater (3LWS-E3)Total duty (kW)90 Flow rate (gpm)150Design pressure (psig)210Temperature in/out (°F)140 / 185Material of constructionCarbon steelWaste Evaporator CondenserTotal duty (Btu/hr)22,717,947TABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 4 of 9Rev. 30Shell Side Tube Side Total liquid entering (lb/hr)24,0401,135,898Temperature in/out (°F)250 / 25095 / 115Design pressure (psig)100150Material of construction304 SS304 SSWaste Distillate CoolerTotal duty (Btu/hr)2,285,500Shell Side Tube Side Total liquid entering (lb/hr)17,500114,275Temperature in/out (°F250 / 12095 / 115Design pressure (psig)150200 Material of construction304 SS304 SSWaste Test Tank HeatersNumber2 Total duty (Btu/hr)10,236Operating flow (gpm)60Design pressure (psig)50 Material of construction304 SSWaste Bottoms Holding Tank HeatersNumber2 Design temperature (°F)200Design pressure (psig)10Total load (kW)18 Material of constructionIncoloy 825Waste EvaporatorCapacity (gpm)35 Design pressure (psig)100 and full vacuum Design temperature (°F)350Material of constructionTop, 316 SS Bottom, Incoloy 825Waste Demineralizer FilterTABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 5 of 9Rev. 30Design flow rate (gpm)160Design pressure (psig)165 Design temperature (°F)200Material of constructionInternals, 304 SSEffluent FiltersNumber2Design Flow rate (gpm)N/ADesign pressure (psig)150

Design temperature (°F)N/ADesign pressure drop, clean (psi)1Material of constructionInternals, 304 SSHigh-Level Waste Recirc FiltersNumber4Design Flow rate (gpm)75 Design pressure (psig)150 Design temperature (°F)200Material of construction304 SSHigh Waste Demineralizer FilterFlow rate (gpm)75Design pressure (psig)110

Design temperature (°F)200Material of constructionInternal, 304 SSLow-Level Waste Recirc FiltersNumber2Design flow rate (gpm)60Design pressure (psig)150 Design Temperature (°F)150Design Pressure drop, clean (psi)1Material of constructionHousing 304 LHigh-Level Waste Drain TanksNumber2TABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 6 of 9Rev. 30Capacity (gal)26,000Design pressure (psig)25 Design temperature (°F)200 Material of constructionShell 304 SSWaste Distillate TankCapacity (gal)500Design pressure (psig)100 Design temperature (°F)340 Material of constructionShell, 304 SS Waste Test TanksNumber2 Capacity (gal)24,000Design pressure (psig)atm and full liquid Design temperature (°F)200 Material of constructionShell, 304 SSLow-Level Waste Drain TanksNumber2 Capacity (gal)4,000Design pressure (psig)25 Design temperature (°F)212 Material of constructionShell, 304 SSWaste Bottoms Holding TankNumber1 Capacity (gal)3,000Design pressure (psig)atm Design temperature (°F)212Material of constructionIncoloy 825 Condensate Demineralizer Liquid Waste System (removed from service)

Regenerant Evaporator Feed Tanks (removed from service)Number2Capacity (gal)13,000TABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 7 of 9Rev. 30Design pressure (psig)atm Design temperature (°F)212Material of constructionFiberglassRegenerant Distillate Ta nk (removed from service)Number1 Capacity (gal)500Design pressure (psig)100 Design temperature (°F)340Material of construction304 SS Regenerant Evaporator Feed Pump (removed from service)Capacity (gal)50 Design pressure (psig)150 Design temperature (°F)150Material of construction316 SS Regenerant Evaporator Bottom s Pump (removed from service)Capacity (gal)15Design pressure (psig)150

Design temperature (°F)274Material of constructionA-20 SS Regenerant Bottoms Coolant Pump (removed from service)Capacity (gal)120Design pressure (psig)150 Design temperature (°F)185Material of construction316 SS Regenerant Distillate Pu mp (removed from service)Capacity (gal)50 Design pressure (psig)150 Design temperature (°F)274Material of construction316 SS Regenerant Evaporator Reboile r Pump (removed from service)Capacity (gal)4,000TABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 8 of 9Rev. 30Design pressure (psig)600 Design temperature (°F)350Material of constructionAlloy 20 Regenerant Distil late Cooler (removed from service)Shell Side Tube Side Total fluid entering (lb/hr)17,500114,275Design pressure (psig)150150 Design temperature (°F)274274Total duty (Btu/hr)2,285,5002,285,500Material of construction304SS304SS Regenerant Evaporator Reboi ler (removed from service)Shell Side Tube Side Total fluid entering (lb/hr)23,7001,930,000Design pressure (psig)180100

Design temperature (°F)380350Total duty (Btu/hr)21,600,00021,600,000 Material of constructionCarbon steelIncoloy 825 Regenerant Evaporator Bottoms Cooler (removed from service)Shell Side Tube SideTotal fluid entering (lb/hr)63,7507,500Design pressure (psig)150150 Design temperature (°F)274274Total duty (Btu/hr)637,500637,500 Material of constructionCarbon steelCarpenter 20 Regenerant Evaporator Conde nser (removed from service)Shell Side Tube Side Total fluid entering (lb/hr)24,0401,135,898Design pressure (psig)100150 Design temperature (°F)350350Total duty (Btu/hr)22,717,94722,717,947Material of construction304 SS304 SSTABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 9 of 9Rev. 30 Regenerant Bottoms Coolant Re heater (removed from service)Design flow (gpm)150Duty (kW)90Design pressure (psig)175 Design temperature (°F)185Material of constructionSA 106 Regenerant Demineralizer (removed from service)Design flow rate (gpm)50 Design pressure (psig)150 Design temperature (°F)250Material of construction304 SS Regenerant Evaporator (removed from service)Capacity (gpm)50Design pressure (psig)100 and full vacuum

Design temperature (°F)300Material of constructionTop, 316 SS Bottom, Incoloy 825 Regenerant Demineralizer Fi lter (removed from service)Flow rate (gpm)50Design pressure (psig)150

Design temperature (°F)120Material of construction304 SSTABLE 11.2-2LIQUID WASTE MANA GEMENT SYSTEM DESIGN DATA MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.2-3TANK OVERFLOW PROTECTIONTanksMark No.

Level Monitoring and AlarmsMonitoring or Alarm Location (1)Overflow ProvisionsProcessing of OverflowContainment Drains Transfer Tank3DGS-TK1IndicatorMCBOverflows to containment sumpIn RLWSPrimary Drain Transfer Tank3DGS-TK2Indicator HighMCBOverflows to auxiliary building sumpIn RLWSBoron Recovery Tanks3BRS-TK1A/1BIndicator HighBRPOverflows to second tank then to diked areaIn RLWS if it overflows to dikeBoron Test Tanks3BRS-TK2A/2BIndicator HighBRPOverflows to the waste disposal building sumpIn RLWS if it overflows to dikeBoron Distillate Tank3BRS-TK3Indicator HighBRPClosed tankNoneCation Regeneration Tank3CND-TKIndicatorLocalLocated in diked area. Overflows to waste neutralization sumpContents of sumps are discharged to the recirculating water discharge tunnel or, if radioactive, processed

in the condensate demineralizer liquid waste systemAnion Regeneration Tank3CND-TK2IndicatorLocalLocated in diked area. Overflows to waste neutralization sumpContents of sumps are discharged to the recirculating water discharge tunnel or, if radioactive, processed in the condensate demineralizer liquid waste systemResin Mix and Storage Tank3CND-TK3IndicatorLocalLocated in diked area. Overflows to waste neutralization sumpContents of sumps are discharged to the recirculating water discharge tunnel or, if radioactive, processed in the condensate demineralizer liquid waste system MPS-3 FSARPage 2 of 3Rev. 30Recovered Caustic Tank3CND-TK8Indicator HighCDLocated in diked area. Overflows to waste neutralization sumpContents of sumps are discharged to the recirculating water discharge tunnel or, if radioactive, processed

in the condensate demineralizer liquid waste systemRecovered Water Tank3CND-TK9Indicator HighCDLocated in diked area. Overflows to waste neutralization sumpContents of sumps are discharged to the recirculating water discharge tunnel or, if radioactive, processed in the condensate demineralizer liquid waste systemRegenerant Evaporator Feed Tanks (removed from service)3LWC-TK1A/BIndicator HighLWCLocated in diked area. Overflows to contaminated floor drainsHigh-Level Waste Drain Tanks3LWS-TK1A/1BIndicator HighLWPOverflows to the waste disposal building sumpIn RLWSLow-Level Waste Drain Tanks3LWS-TK4A/4BIndicator HighLWPOverflows to the waste disposal building sumpIn RLWSWaste Distillate Tank3LWS-TK2Indicator HighLWPClosed tankNoneRegenerant Distillate Tank (removed from service)3LWC-TK3Indicator HighLWCClosed tankNoneWaste Test Tanks3LWS-TK3A/3BIndicator HighLWPOverflows to the waste disposal building sumpIn RLWSSpent Resin Dewatering Tank3WSS-TK1Indicator HighSWPTo solid waste building sumpIn RLWSTABLE 11.2-3TANK OVERFLOW PROTECTIONTanksMark No.

Level Monitoring and AlarmsMonitoring or Alarm Location (1)Overflow ProvisionsProcessing of Overflow MPS-3 FSARPage 3 of 3Rev. 30NOTE (1) Location Symbols BRP = Boron Recovery Panel MCB = Main Control BoardLWP = Radioactive Waste PanelSpent Resin Hold Tank3WSS-TK3IndicatorSWPOverflows to spent resin dewatering tankIn RLWSWaste Bottoms Holding Tank3LWS-TK5Indicator HighSWPOverflows to the waste disposal building sumpIn RLWSCondensate Surge Tank3CNS-T2Indicator HighMCBOverflows to turbine building floor drains sump Contents of sump are routed to the radioactive liquid waste systemTABLE 11.2-3TANK OVERFLOW PROTECTIONTanksMark No.

Level Monitoring and AlarmsMonitoring or Alarm Location (1)Overflow ProvisionsProcessing of Overflow MPS-3 FSAR Page 1 of 4 Rev. 30TABLE 11.2-4EXPECTED RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM

(µCI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)ISOTOPE BORON RECOVERY (µCi/ml) MISC. WASTES (µCi/ml)SECONDARY (2) SYSTEM WASTES (µCi/ml) TURB BLDG DRAINS (µCi/ml) TOTAL LWS (µCi/ml)ADJUSTED (3) TOTAL LWS

(µCi/ml) DETERGENT (4) WASTES (µCi/ml) CORROSION AND ACTIVATION PRODUCTSCr-51 1.66E-08 (5)7.05E-076.68E-090.00E+008.77E-099.17E-090.00E+00Mn-544.16E-091.21E-071.75E-090.00E+002.10E-092.19E-091.61E-06Fe-552.91E-086.15E-077.08E-090.00E+008.97E-099.36E-090.00E+00 F-591.25E-083.73E-074.71E-090.00E+005.82E-096.08E-090.00E+00Co-582.16E-076.06E-066.54E-080.00E+008.37E-088.74E-086.43E-06Co-603.74E-087.68E-078.01E-090.00E+001.04E-081.08E-081.40E-05 Zr-950.00E+000.00E+000.00E+000.00E+000.00E+000.00E+002.25E-06Nb-950.00E+000.00E+000.00E+000.00E+000.00E+000.00E+003.21E-06Np-2390.00E+003.68E-073.68E-090.00E+004.75E-094.96E-090.00E+00 FISSION PRODUCTSBr-830.00E+003.89E-077.00E-090.00E+008.05E-098.41E-090.00E+00Br-840.00E+001.58E-081.83E-090.00E+001.84E-091.94E-090.00E+00Br-850.00E+000.00E+003.36E-110.00E+003.29E-113.29E-110.00E+00Rb-864.16E-092.79E-072.69E-100.00E+001.17E-091.22E-090.00E+00 Rb-880.00E+002.47E-079.58E-080.00E+009.43E-089.85E-080.00E+00Sr-894.16E-091.31E-071.60E-090.00E+001.99E-092.07E-090.00E+00 MPS-3 FSAR Page 2 of 4 Rev. 30S-900.00E+005.26E-093.36E-110.00E+004.93E-114.93E-110.00E+00S-910.00E+001.42E-071.43E-090.00E+001.84E-091.94E-090.00E+00 Y-900.00E+000.00E+001.68E-110.00E+001.64E-111.64E-110.00E+00Y-91m0.00E+009.47E-081.16E-090.00E+001.43E-091.50E-090.00E+00Y-910.00E+002.63E-082.52E-100.00E+003.29E-103.29E-100.00E+00 Y-930.00E+005.26E-096.73E-110.00E+009.86E-119.86E-110.00E+00Zr-950.00E+002.10E-083.20E-100.00E+003.94E-104.11E-100.00E+00Nb-950.00E+002.10E-083.53E-100.00E+004.11E-104.27E-100.00E+00 Mo-994.16E-082.62E-052.45E-072.01E-093.21E-073.35E-070.00E+00Tc-99m4.16E-082.06E-052.44E-072.01E-093.02E-073.16E-070.00E+00Ru-1030.00E+001.58E-081.51E-100.00E+002.14E-102.14E-102.25E-07Ru-1060.00E+005.26E-093.36E-110.00E+004.93E-114.93E-113.86E-06Rh-103m0.00E+001.58E-082.69E-100.00E+003.12E-103.29E-100.00E+00 Rh-1060.00E+005.26E-096.73E-110.00E+008.21E-118.21E-110.00E+00Ag-110m0.00E+000.00E+000.00E+000.00E+000.00E+000.00E+007.07E-07Te-125m0.00E+001.05E-088.41E-110.00E+001.15E-101.15E-100.00E+00 Te-127m4.16E-091.05E-078.41E-100.00E+001.17E-091.22E-090.00E+00Te-1274.16E-092.37E-072.44E-090.00E+003.14E-093.29E-090.00E+00TABLE 11.2-4EXPECTED RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM

(µCI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)ISOTOPE BORON RECOVERY (µCi/ml) MISC. WASTES (µCi/ml)SECONDARY (2) SYSTEM WASTES (µCi/ml) TURB BLDG DRAINS (µCi/ml) TOTAL LWS (µCi/ml)ADJUSTED (3) TOTAL LWS

(µCi/ml) DETERGENT (4) WASTES (µCi/ml)

MPS-3 FSAR Page 3 of 4 Rev. 30Te-129m1.25E-085.21E-074.56E-090.00E+006.14E-096.41E-090.00E+00Te-1298.31E-093.58E-077.34E-090.00E+008.31E-098.67E-090.00E+00 Te-131m0.00E+007.10E-076.31E-090.00E+008.38E-098.76E-090.00E+00Te-1310.00E+001.31E-072.52E-090.00E+002.89E-093.02E-090.00E+00Te-1321.66E-088.55E-066.12E-081.00E-098.66E-089.04E-080.00E+00 I-1300.00E+005.00E-074.07E-090.00E+005.54E-095.78E-090.00E+00I-1316.54E-069.30E-051.11E-064.92E-081.40E-061.46E-069.64E-08I-1321.66E-081.46E-051.97E-074.02E-092.38E-072.49E-070.00E+00 I-1339.97E-081.02E-047.93E-075.83E-081.10E-061.14E-060.00E+00I-1340.00E+008.78E-074.40E-080.00E+004.57E-084.77E-080.00E+00I 1350.00E+003.53E-053.46E-071.81E-084.48E-074.69E-070.00E+00Cs-1344.82E-061.05E-049.40E-081.00E-094.40E-074.60E-072.09E-05Cs-1365.82E-073.83E-054.10E-080.00E+001.62E-071.69E-070.00E+00 Cs-1373.59E-067.63E-056.84E-081.00E-093.19E-073.34E-073.86E-05Ba-137m3.35E-067.14E-051.41E-070.00E+003.74E-073.90E-070.00E+00Ba-1400.00E+007.89E-086.73E-100.00E+009.04E-109.36E-100.00E+00 La-1400.00E+006.31E-085.38E-100.00E+007.23E-107.56E-100.00E+00Ce-1410.00E+002.63E-083.03E-100.00E+003.78E-103.94E-100.00E+00TABLE 11.2-4EXPECTED RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM

(µCI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)ISOTOPE BORON RECOVERY (µCi/ml) MISC. WASTES (µCi/ml)SECONDARY (2) SYSTEM WASTES (µCi/ml) TURB BLDG DRAINS (µCi/ml) TOTAL LWS (µCi/ml)ADJUSTED (3) TOTAL LWS

(µCi/ml) DETERGENT (4) WASTES (µCi/ml)

MPS-3 FSAR Page 4 of 4 Rev. 30NOTES:(1) Refer to Figure11.2-3 for respective steam processing.(2) Includes steam generator blowdown during open cy lce blowdown and regenerate chemical wastes.(3) Adjusted for releases from anticipated operational occurrences of 0.15 Ci/yr (NUREG-0017).(4) Detergent waste is conservatively in cluded for flexibility in the event that a laundry is installed in the future.(5) 1.66E-08 = 1.66 x 10

-08; values less than 1.0E-15 are reported as zero.

Historical, not subject to future updating. This table ha s been retained to preser ve original license basis.Ce-1430.00E+001.05E-086.73E-110.00E+009.86E-119.86E-110.00E+00C-1440.00E+001.05E-081.68E-080.00E+002.14E-102.14E-108.36E-06 Pr-1430.00E+001.58E-081.35E-100.00E+001.97E-101.97E-100.00E+00Pr-1440.00E+001.05E-083.20E-100.00E+003.61E-103.78E-100.00E+00TABLE 11.2-4EXPECTED RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM

(µCI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)ISOTOPE BORON RECOVERY (µCi/ml) MISC. WASTES (µCi/ml)SECONDARY (2) SYSTEM WASTES (µCi/ml) TURB BLDG DRAINS (µCi/ml) TOTAL LWS (µCi/ml)ADJUSTED (3) TOTAL LWS

(µCi/ml) DETERGENT (4) WASTES (µCi/ml)

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.2-5EXPECTED ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM AND PRIOR TO INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Concentration

(µCi/gm)Radioactivity Released (Ci/yr)CORROSION AND ACTIVATION PRODUCTSCr-51 6.85E-08 (2)5.34E-03Mn-541.64E-081.28E-03Fe-557.01E-085.46E-03Fe-594.54E-083.54E-03 Co-586.53E-075.09E-02Co-608.10E-086.31E-03Zr-950.00E+000.00E+00 Nb-950.00E+000.00E+00Np-239 3.71E-082.89E-03FISSION PRODUCTS Br-836.29E-084.90E-03B-841.44E-081.12E-03Br-852.57E-102.00E-05 Rb-869.11E-097.10E-04Rb-887.36E-075.74E-02Sr-891.55E-081.21E-03 Sr-903.85E-103.00E-05Sr-911.44E-081.12E-03Y-901.28E-101.00E-05 Y-91m1.12E-088.70E-04Y-912.57E-092.00E-04Y-937.70E-106.00E-05Zr-953.08E-092.40E-04Nb-953.21E-092.50E-04 Mo-992.51E-061.95E-01 MPS-3 FSARPage 2 of 3Rev. 30Tc-99m2.36E-061.84E-01Ru-1031.67E-091.30E-04Ru-1063.85E-103.00E-05 Rh-103m2.44E-091.90E-04Rh-1066.42E-105.00E-05Ag-110m0.00E+000.00E+00 Te-125m8.98E-107.00E-05Te-127m9.11E-097.10E-04Te-1272.45E-081.91E-03 Te-129m4.80E-083.74E-03Te-1296.49E-085.06E-03Te-131m6.54E-085.10E-03 Te-1312.26E-081.76E-03Te-1326.76E-075.27E-02I-1304.32E-083.37E-03 I-1311.09E-058.50E-01I-1321.86E-061.45E-01I-1338.56E-066.67E-01 I-1343.57E-072.78E-02I-1353.50E-062.73E-01Cs-1343.44E-062.68E-01 Cs-1361.26E-069.86E-02Cs-1372.49E-061.94E-01Ba-137m2.92E-062.27E-01Ba-1407.06E-095.50E-04La-1405.65E-094.40E-04 Ce-1412.95E-092.30E-04Ce-1437.70E-106.00E-05TABLE 11.2-5EXPECTED ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM AND PRIOR TO INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Concentration

(µCi/gm)Radioactivity Released (Ci/yr)

MPS-3 FSARPage 3 of 3Rev. 30NOTES:(1) Detergent wastes not included in this table(2) 6.85E-08 = 6.85 x 10

-83. Total annual liquid waste release volume (excluding detergent wastes) is 7.79E+10 mlHistorical, not subject to future updating. This table has been retained to preserve original license basis.Ce-1441.67E-091.30E-04Pr-1431.54E-091.20E-04Pr-144 2.82E-092.20E-04 Total (excluding H-3)4.30E-053.35E+00H-39.24E-037.20E+02Total (including H-3)9.28E-037.23E+02TABLE 11.2-5EXPECTED ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM AND PRIOR TO INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Concentration

(µCi/gm)Radioactivity Released (Ci/yr)

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.2-6EXPECTED A NNUAL RADIOACTIVE LI QUID RELEASES AFTER DILUTION IN THE CIRCULATING WATER DISHARGE SYSTEM AND INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Released

(µCi/gm)Radioactivity Released (Ci/yr)CORROSION AND ACTIVATION PRODUCTSCr-51 3.41E-12 (2)5.60E-03Mn-541.40E-122.30E-03Fe-553.47E-125.70E-03Fe-592.25E-123.70E-03 Co-583.47E-115.70E-02Co-609.12E-121.50E-02Zr-958.51E-131.40E-03 Nb-951.22E-122.00E-03Np-2391.82E-123.00E-03FISSION PRODUCTS Br-833.10E-125.10E-03Br-847.30E-131.20E-03Br-851.22E-142.00E-05 Rb-864.50E-137.40E-04Rb-883.65E-116.00E-02Sr-89 7.91E-131.30E-03 Sr-901.82E-143.00E-05Sr-917.30E-131.20E-03Y-906.08E-151.00E-05 Y-91m5.53E-139.10E-04Y-911.22E-132.00E-04Y-933.65E-146.00E-05Zr-951.52E-132.50E-04Nb-951.58E-132.60E-04 Mo-991.22E-102.00E-01 MPS-3 FSARPage 2 of 3Rev. 30Tc-99m1.16E-101.90E-01Ru-1031.64E-132.70E-04Ru-1061.46E-122.40E-03 Rh-103m1.22E-132.00E-04Rh-1063.04E-145.00E-05Ag-110m2.68E-134.40E-04 Te-125m4.26E-147.00E-05Te-127m4.50E-137.40E-04Te-1271.22E-122.00E-03 Te-129m2.37E-123.90E-03Te-1293.22E-125.30E-03Te-131m3.22E-125.30E-03 Te-1311.09E-121.80E-03Te-1323.34E-115.50E-02I-1302.13E-123.50E-03 I-1315.41E-108.90E-01I-1329.12E-111.50E-01I-1334.26E-107.00E-01 I-1341.76E-112.90E-02I-1351.76E-102.90E-01Cs-1341.76E-102.90E-01 Cs-1366.08E-111.00E-01Cs-1371.40E-102.30E-01Ba-137m1.46E-102.40E-01Ba-1403.47E-135.70E-04La-1402.80E-134.60E-04 Ce-1411.46E-132.40E-04Ce-1433.65E-146.00E-05TABLE 11.2-6EXPECTED A NNUAL RADIOACTIVE LI QUID RELEASES AFTER DILUTION IN THE CIRCULATING WATER DISHARGE SYSTEM AND INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Released

(µCi/gm)Radioactivity Released (Ci/yr)

MPS-3 FSARPage 3 of 3Rev. 30(1) 1. Detergent wastes are included in this table(2) 3.41E-12 = 3.41 x 10 123. Includes source term adjustment for anti cipated operational occurrences 1.50E-01 Ci/yr4. Total annual liquid waste re lease volume is 7.86E+10 ml/yr5. MP3 average dilution flow of 1840ft 3/sec or 1.64E+15 ml/yr Historical, not subject to future updating. This table has been retained to preserve original license basis.Ce-1443.22E-125.30E-03Pr-1437.30E-141.20E-04 Pr-1441.40E-132.30E-04 Total (excluding H-3)2.19E-093.60E+00H 34.38E-077.20E+02Total (including H-3)4.40E-077.24E+02TABLE 11.2-6EXPECTED A NNUAL RADIOACTIVE LI QUID RELEASES AFTER DILUTION IN THE CIRCULATING WATER DISHARGE SYSTEM AND INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)

Isotope Radioactivity Released

(µCi/gm)Radioactivity Released (Ci/yr)

MPS-3 FSARPage 1 of 4Rev. 30TABLE 11.2-7DESIGN (1) RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM (µCI/GM) FOLLOWING TREATMENT (2)Isotope Cont.Bldg.

Sump Aux. Bldg.

SumpLaboratory Waste Reactor Plant SamplesMisc. Low-Level Waste Misc. High-Level Waste Reactor Coolant Bleed Reactor Plant Gaseous Drains Regenerant ChemicalsTurbine Bldg.

DrainsSteam Generator Blowdown (3)Cr-51 1.7E-07 41.7E-083.3E-101.7E-078.7E-091.7E-098.8E-098.8E-086.9E-111.8E-097.2E-07Mn-543.2E-083.2E-096.5E-113.2E-083.0E-063.2E-103.0E-093.0E-081.2E-113.0E-101.2E-07Fe-551.7E-071.7E-083.4E-101.7E-071.7E-051.7E-091.7E-081.7E-076.2E-111.5E-096.2E-07Fe-599.8E-089.8E-092.0E-109.8E-086.3E-069.8E-106.4E-096.4E-083.9E-111.0E-094.0E-07Co-581.6E-061.6E-073.2E-091.6E-061.2E-041.6E-081.2E-071.2E-066.1E-101.5E-086.2E-06 Co-602.1E-072.1E-084.2E-102.1E-072.1E-052.1E-092.1E-082.1E-077.6E-111.9E-097.6E-07Sr-893.9E-073.9E-087.8E-103.9E-072.6E-053.9E-092.7E-082.7E-071.5E-103.9E-091.6E-06Sr-901.7E-081.7E-093.4E-111.7E-081.7E-061.7E-101.7E-091.7E-086.2E-121.6E-106.2E-08 Sr-913.8E-093.8E-107.6E-123.8E-091.4E-073.8E-114.8E-134.8E-127.9E-121.5E-096.1E-07Y-901.8E-081.8E-093.6E-111.8E-081.7E-061.8E-101.7E-091.7E-087.1E-121.9E-107.5E-08Y-91M2.6E-092.6E-105.2E-122.6E-099.5E-082.6E-113.2E-133.2E-125.4E-129.3E-103.7E-07 Y-916.4E-086.4E-091.3E-106.4E-084.5E-066.4E-104.6E-094.6E-082.5E-116.3E-102.5E-07Y-937.4E-107.4E-111.5E-127.4E-102.6E-087.4E-121.3E-131.3E-121.5E-122.6E-101.1E-07Zr-956.6E-086.6E-091.3E-106.6E-084.8E-066.6E-104.9E-094.9E-082.5E-116.5E-102.6E-07 Nb-957.3E-087.3E-091.5E-107.3E-086.3E-067.3E-106.4E-096.4E-082.7E-116.7E-102.7E-07Mo-998.0E-058.0E-061.6E-078.0E-051.7E-038.0E-078.5E-078.5E-068.0E-083.0E-061.2E-03Tc-99M7.6E-057.6E-061.5E-077.6E-051.6E-037.6E-078.2E-078.2E-067.5E-082.0E-067.8E-04Ru-1033.0E-083.0E-096.0E-113.0E-081.8E-063.0E-101.9E-091.9E-081.2E-113.1E-101.2E-07 MPS-3 FSARPage 2 of 4Rev. 30Ru-1063.2E-093.2E-106.4E-123.2E-093.0E-073.2E-113.0E-103.0E-091.2E-123.0E-111.2E-08Rh-103M3.0E-083.0E-096.0E-113.0E-081.8E-063.0E-101.9E-091.9E-081.2E-112.9E-101.2E-07Rh-1063.2E-093.2E-106.4E-123.2E-093.0E-073.2E-113.0E-103.0E-091.2E-122.7E-111.1E-08Te-125M8.2E-098.2E-101.6E-118.2E-095.8E-078.3E-115.9E-105.9E-093.2E-128.2E-113.3E-08Te-127M2.0E-072.0E-084.0E-102.0E-071.7E-052.0E-091.7E-081.7E-077.6E-111.9E-097.7E-07Te-1271.9E-071.9E-083.9E-101.9E-071.6E-051.9E-091.6E-081.6E-077.0E-111.1E-094.5E-07Te-129M3.3E-063.3E-076.7E-093.3E-061.5E-043.3E-081.9E-071.9E-061.4E-093.5E-081.4E-05Te-1292.1E-062.1E-074.3E-092.1E-061.2E-042.1E-081.2E-071.2E-068.7E-102.2E-088.7E-06Te-131M2.3E-072.3E-084.5E-102.3E-075.3E-062.3E-099.7E-109.7E-093.5E-102.0E-088.0E-06Te-1314.1E-084.1E-098.3E-114.1E-089.9E-074.1E-101.8E-101.8E-096.3E-115.5E-092.2E-06Te-1327.5E-067.5E-071.5E-087.5E-061.6E-047.5E-089.0E-089.0E-076.8E-092.4E-079.6E-05Ba-137M1.6E-041.6E-053.1E-071.6E-041.6E-021.6E-063.9E-047.8E-041.0E-071.4E-065.7E-04Ba-1403.0E-073.0E-085.9E-103.0E-071.0E-053.0E-099.6E-099.6E-081.4E-104.0E-091.6E-06La-1402.9E-072.9E-085.8E-102.9E-071.0E-052.9E-091.1E-081.1E-071.0E-101.5E-095.9E-07Ce-1415.9E-085.9E-091.2E-105.9E-083.4E-065.9E-103.4E-093.4E-082.4E-116.3E-102.5E-07Ce-1435.6E-095.6E-101.1E-115.6E-091.3E-075.6E-112.8E-112.8E-108.3E-124.5E-101.8E-07Ce-1444.9E-084.9E-099.8E-114.9E-084.5E-064.9E-104.5E-094.5E-081.8E-114.5E-101.8E-07Pr-1435.0E-085.0E-091.0E-105.0E-081.8E-065.0E-101.7E-091.7E-082.3E-116.1E-102.5E-07Pr-1444.9E-084.9E-099.7E-114.9E-084.5E-064.9E-104.5E-094.5E-081.8E-114.5E-101.8E-07TABLE 11.2-7DESIGN (1) RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM (µCI/GM) FOLLOWING TREATMENT (2)Isotope Cont.Bldg.

Sump Aux. Bldg.

SumpLaboratory Waste Reactor Plant SamplesMisc. Low-Level Waste Misc. High-Level Waste Reactor Coolant Bleed Reactor Plant Gaseous Drains Regenerant ChemicalsTurbine Bldg.

DrainsSteam Generator Blowdown (3)

MPS-3 FSARPage 3 of 4Rev. 30Np-2397.8E-087.8E-091.6E-107.8E-081.7E-067.8E-107.2E-107.2E-098.5E-113.4E-091.4E-06Br-835.5E-085.5E-091.1E-105.5E-081.6E-065.5E-100.00.01.7E-101.5E-071.5E-05Br-843.1E-133.1E-140.03.1E-131.6E-073.1E-150.00.04.7E-143.1E-083.1E-06 Br-850.00.00.00.02.0E-090.00.00.00.05.1E-105.2E-08I-1302.1E-072.1E-084.2E-102.1E-076.7E-072.1E-099.0E-119.0E-103.1E-101.7E-081.7E-06I-1311.4E-031.4E-042.9E-061.4E-033.8E-031.4E-053.3E-053.3E-046.0E-076.9E-066.9E-04 I-1321.2E-051.2E-062.4E-081.2E-051.7E-041.2E-079.3E-089.3E-072.3E-082.0E-062.0E-04I-1332.5E-042.5E-055.0E-072.5E-046.5E-042.5E-065.2E-075.2E-063.3E-071.0E-051.0E-03I-1341.1E-091.1E-102.2E-121.1E-093.8E-061.1E-110.00.02.2E-116.3E-076.3E-05 I-1352.2E-052.2E-064.3E-082.2E-051.1E-042.2E-072.9E-102.9E-093.8E-084.9E-064.9E-04Rb-862.0E-082.0E-094.1E-112.0E-088.6E-072.0E-102.1E-084.3E-081.6E-114.4E-101.7E-07Rb-880.00.00.00.07.5E-060.00.00.00.04.6E-071.8E-04 Cs-1343.3E-053.3E-066.6E-083.3E-053.2E-033.3E-078.0E-051.6E-042.2E-085.4E-072.2E-04Cs-1361.1E-051.1E-062.3E-081.1E-053.9E-041.1E-079.4E-061.9E-051.0E-082.7E-071.1E-04Cs-1371.7E-041.7E-053.3E-071.7E-041.7E-021.7E-064.1E-048.3E-041.1E-072.7E-061.1E-03TABLE 11.2-7DESIGN (1) RADIOACTIVE LIQUID CONCENTRATIONS FROM EACH LIQUID RELEASE STREAM (µCI/GM) FOLLOWING TREATMENT (2)Isotope Cont.Bldg.

Sump Aux. Bldg.

SumpLaboratory Waste Reactor Plant SamplesMisc. Low-Level Waste Misc. High-Level Waste Reactor Coolant Bleed Reactor Plant Gaseous Drains Regenerant ChemicalsTurbine Bldg.

DrainsSteam Generator Blowdown (3)

MPS-3 FSARTable 11.2-11 ContinuedPage 4 of 4Rev. 30NOTES:(1) Design values represent assumptions used to estimate liquid radiological effluents prior to initial plant licensing and are retained for historical purposes only. The Radiological Effluent Monitoring and Offsit e Dose Calculation Manual (REMODCM) provide requirements for system operation, dose calculations and monitoring to ensure compliance with 10CFR20 AppendixB, TableII, Column2 effluent limits.(2) Refer to Figure 11.2-3 for respective stream processing.Values less than 1.0E-15 are reported as zero.(3) During open cycle blowdown.(4) 1.7E-07 = 1.7 x 10

-7 MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.2-8DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEMIsotopeActivity Released (µCi/gm)Activity Released (Ci/yr)Cr-51 5.5E-07 (2)1.9E-02Mn-549.3E-083.2E-03Fe-554.8E-071.6E-02 Fe-593.0E-071.0E-02Co-584.7E-061.6E-01Co-605.9E-072.0E-02 Sr-891.2E-064.1E-02Sr-904.8E-081.6E-03Sr-914.5E-071.5E-02 Y-905.8E-082.0E-03Y-91M2.8E-079.4E-03Y-911.9E-076.5E-03 Y-937.9E-082.7E-03Zr-952.0E-076.7E-03Nb-952.1E-077.0E-03 Mo-999.0E-043.0E+01Tc-99M5.9E-042.0E+01Ru-1039.5E-083.2E-03 Ru-1069.2E-093.1E-04Rh-103M8.9E-083.0E-03Rh-1068.7E-093.0E-04 Te-125M2.5E-088.5E-04Te-127M5.9E-072.0E-02Te-1273.6E-071.2E-02Te-129M1.1E-053.7E-01Te-1296.7E-062.3E-01 Te-131M6.0E-062.0E-01 MPS-3 FSARPage 2 of 3Rev. 30Te-1311.6E-065.5E-02Te-1327.2E-052.4E+00Ba-137M4.4E-041.5E+01Ba-1401.2E-064.0E-02 La-1404.6E-071.5E-02Ce-1411.9E-076.5E-03Ce-1431.3E-074.5E-03 Ce-1441.4E-074.7E-03Pr-1431.9E-076.3E-03Pr-1441.4E-074.7E-03 Np-2391.0E-063.4E-02Br-831.1E-053.8E-01Br-842.3E-067.9E-02 Br-853.9E-081.3E-03I-1301.3E-064.3E-02I-1315.2E-041.8E+01 I-1321.5E-045.0E+00I-1337.7E-042.6E+01I-1344.7E-051.6E+00 I-1353.6E-041.2E+01Rb-861.3E-074.4E-03Rb-881.4E-044.6E+00 Cs-1341.7E-045.6E+00Cs-1368.2E-052.8E+00Cs-1378.3E-042.8E+01H-37.5E-022.5E+03TABLE 11.2-8DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEMIsotopeActivity Released (µCi/gm)Activity Released (Ci/yr)

MPS-3 FSAR Table 11.2-8 ContinuedPage 3 of 3Rev. 30NOTES:(1) Design values represent assumptions used to estimate liquid radiological effluents prior to initial plant licensing and are retained for historical purposes only. The Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) provide requirements for system operation, dose calculations and monitoring to ensure compliance with 10CFR20 AppendixB, TableII, Column2 effluent limits.(2) 5.5E-07 = 5.5 x 10

-7.Total liquid waste release is 3.4E+10 gm/yr.Total release (excluding tritium) is 1.7+02 Ci/yr.

Total release concentration (excl uding tritium) is 5.1E-03 mCi/gm.

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.2-9DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES FOLLOWING ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEMIsotopeActivity Released (µCi/gm)Activity Released (Ci/yr)Cr-51 6.3E-12 (2)1.9E-02Mn-541.1E-123.2E-03Fe-555.5E-121.6E-02 Fe-593.5E-121.0E-02Co-585.4E-111.6E-01Co-606.8E-122.0E-02 Sr-891.4E-114.1E-02Sr-905.5E-131.6E-03Sr-915.2E-121.5E-02 Y-906.6E-132.0E-03Y-91M3.2E-129.4E-03Y-912.2E-126.5E-03 Y-939.0E-132.7E-03Zr-952.3E-126.7E-03Nb-952.4E-127.0E-03 Mo-991.0E-083.0E+01Tc-99M6.7E-092.0E+01Ru-1031.1E-123.2E-03 Ru-1061.1E-133.1E-04Rh-103M1.0E-123.0E-03Rh-1061.0E-133.0E-04 Te-125M2.9E-138.5E-04Te-127M6.7E-122.0E-02Te-1274.1E-121.2E-02Te-129M1.2E-103.7E-01Te-1297.6E-112.3E-01 Te-131M6.8E-112.0E-01 MPS-3 FSARPage 2 of 3Rev. 30Te-1311.9E-115.5E-02Te-1328.2E-102.4E+00Ba-137M5.0E-091.5E+01Ba-1401.4E-114.0E-02 La-1405.2E-121.5E-02Ce-1412.2E-126.5E-03Ce-1431.5E-124.5E-03 Ce-1441.6E-124.8E-03Pr-1432.1E-126.3E-03Pr-1441.6E-124.7E-03 Np-2391.2E-113.5E-02Br-831.3E-103.8E-01Br-842.6E-117.9E-02 Br-854.4E-131.3E-03I-1301.5E-114.3E-02I-1315.9E-091.8E+01 I-1321.7E-095.0E+00I-1338.7E-092.6E+01I-1345.3E-101.6E+00 I-1354.1E-091.2E+01Rb-861.5E-124.4E-03Rb-881.6E-094.6E+00 Cs-1341.9E-095.6E+00Cs-1369.4E-102.8E+00Cs-1379.4E-092.8E+01H-38.6E-072.5E+03TABLE 11.2-9DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES FOLLOWING ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEMIsotopeActivity Released (µCi/gm)Activity Released (Ci/yr)

MPS-3 FSARPage 3 of 3Rev. 30NOTES:(1) Design values represent assumptions used to estimate liquid radiological effluents prior to initial plant licensing and are retained for historical purposes only. The Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) provide requirements for system operation, dose calculations and monitoring to ensure compliance with 10CFR20 Appendix B, Table II, Column 2 effluent limits.(2) 6.3E-12 = 6.3 x 10

-12 Anticipated operational occurrences = 1.5E-01 Ci/yr.Dilution release rate is 2.97E+15 gm/yr.

Total release (excluding tritium) is 1.7E+02 Ci/yr.Total release concentration (e xcluding tritium) is 5.8E-08

µCi/gm.

MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.2-10FRACTION OF MPC RELEASED - DESIGN CASE (1) (HISTORICAL)IsotopeConc (µCi/cc)MPC (µCi/cc)Fraction MPC ReleasedCr-51 6.3E-12 (1)2.0E-033.1E-09Mn-541.1E-121.0E-041.1E-08Fe-555.5E-128.0E-046.9E-09Fe-593.5E-125.0E-057.0E-08 Co-585.4E-119.0E-056.0E-07Co-606.8E-123.0E-052.3E-07Sr-891.4E-113.0E-064.7E-06 Sr-905.5E-133.0E-071.8E-06Sr-915.2E-125.0E-051.0E-07Y-906.6E-132.0E-053.3E-08 Y-91M3.2E-123.0E-031.1E-09Y-912.2E-123.0E-057.3E-08Y-939.0E-133.0E-053.0E-08 Zr-952.3E-126.0E-053.8E-08Nb-952.4E-121.0E-042.4E-08Mo-991.0E-084.0E-052.5E-04 Tc-99M6.7E-093.0E-032.2E-06Ru-1031.1E-128.0E-051.4E-08Ru-1061.1E-131.0E-051.1E-08 Rh-103M1.0E-121.0E-021.0E-10Te-125M2.9E-131.0E-042.9E-09Te-127M6.7E-125.0E-051.3E-07 Te-1274.1E-122.0E-042.0E-08Te-129M1.2E-102.0E-056.0E-06Te-1297.6E-118.0E-049.5E-08Te-131M6.8E-114.0E-051.7E-06Te-1328.2E-102.0E-054.1E-05 Ba-1401.4E-112.0E-057.0E-07 MPS-3 FSARPage 2 of 2Rev. 30NOTE (1) 1. 6.3E-12 = 6.3 x 10

-12Total concentration (excluding tritium) is 5.1E-08

µCi/gm.Total fraction of MPC released (excluding tritium) is 3.07E-02.Historical, not subject to future updating. This table has been retained to preserve original license basis.La-1405.2E-122.0E-052.6E-07Ce-1412.2E-129.0E-052.4E-08Ce-1431.5E-124.0E-053.7E-08Ce-1441.6E-121.0E-051.6E-07Pr-1432.1E-125.0E-054.2E-08 Np-2391.2E-111.0E-041.2E-07Br-831.3E-103.0E-064.3E-05I-1301.5E-113.0E-065.0E-06 I-1315.9E-093.0E-072.0E-02I-1321.7E-098.0E-062.1E-04I-1338.7E-091.0E-068.7E-03 I-1345.3E-102.0E-052.6E-05I-1354.1E-094.0E-061.0E-03Rb-861.5E-122.0E-057.5E-08 Cs-1341.9E-099.0E-062.1E-04Cs-1369.4E-106.0E-051.6E-05Cs-1379.4E-092.0E-054.7E-04 H-38.6E-073.0E-032.9E-04Totals9.1E-072.4E-023.1E-02TABLE 11.2-10FRACTION OF MPC RELEASED - DESIGN CASE (1) (HISTORICAL)IsotopeConc (µCi/cc)MPC (µCi/cc)Fraction MPC Released MPS-3 FSARPage 1 of 1Rev. 30NOTE:1.Transit time to Niantic Bay is the travel time of the ground water. Credit for sorption in soil was not assumed to this analysis.TABLE 11.2-11ASSUMPTIONS USED FOR THE RADIOACTIVE LIQUID WASTE SYSTEM FAILURE (RELEASE TO ATMOSPHERE) AND FOR THE LIQUID CONTAINING TANK FAILUREDesignTank with Assumed Highest Radionuclide InventoryBoron Recovery TankTank Volume (gal)150,000 Fill Fraction of Tank0.8Reactor Coolant Feed Rate to Tank (gpm)75Degasification Fraction Upstream to Tank1.0 Source Stream Feeding TankReactor Coolant (Table 11.1-2)Fraction of Volatile Nuclides Released to AtmosphereNoble Gases1.0 Halogens0.1Dilution Factor of Ground Water to Niantic Bay73Transition Time to Niantic Bay (years)

(1)6.64Dilution Factor at Point of Entry to Niantic Bay13,052Dilution Factor at 1,000 Feet from Point of Entry to Bay32,151 MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.2-12BORON RECOVERY TANK CONCENTRATIONS (µCI/CC)IsotopeDesign (µCi/cc)IsotopeDesign (µCi/cc)IsotopeDesign (µCi/cc)Kr-83m 0.13E-03 (1)Y-940.39E-08Te-1270.19E-04Xe-131m0.90E-03Y-950.12E-08Te-129m0.43E-03Xe-133m0.13E-03Zr-950.79E-05Te-1290.42E-03 Xe-1330.24E-02Zr-970.28E-05Te-131m0.20E-03Xe-135m0.25E-02Nb-95m0.16E-06Te-1310.41E-04Xe-1350.61E-02Nb-950.83E-05Te-1320.27E-02 Br-830.13E-03Nb-97m0.27E-05Te-133m0.14E-04Br-840.13E-04Nb-970.31E-05Te-1330.15E-04Br-850.17E-06Mo-990.33E-01Te-1340.14E-04 Br-870.32E-07Mo-1010.24E-05Cs-1340.20E+00I-1290.53E-09Mo-1020.12E-05Cs-1360.10E+00I-1310.28E-01Mo-1050.95E-09Cs-1370.10E+01 I-1320.37E-02Tc-99m0.27E-01Cs-1380.16E-01I-1330.31E-01Tc-1010.46E-05Cs-1390.45E-03I-1340.32E-03Tc-1020.16E-05Cs-1400.54E-05 I-1350.85E-02Tc-1050.12E-07Cs-1420.22E-08I-1360.92E-07Ru-1030.38E-05Ba-137m0.93E+00Se-810.11E-09Ru-1050.81E-07Ba-1390.52E-03 Se-830.24E-09Ru-1060.37E-06Ba-1400.48E-04Se-840.17E-10Ru-1070.61E-11Ba-1410.23E-07Rb-880.33E-01Rh-103m0.38E-05Ba-1420.22E-07 Rb-890.85E-03Rh-105m0.81E-07La-1400.23E-04Rb-900.40E-03Rh-1050.76E-06La-1410.67E-06Rb-910.64E-05Rh-1060.37E-06La-1420.79E-09 Rb-920.28E-07Rh-1070.21E-09La-1430.21E-08Sr-890.61E-04Sn-1270.21E-08Ce-1410.77E-05Sr-900.19E-05Sn-1280.21E-08Ce-1430.45E-05Sr-910.10E-04Sn-1300.14E-10Ce-1440.56E-05Sr-920.13E-05Sb-1270.15E-06Ce-1450.59E-10 MPS-3 FSARPage 2 of 2Rev. 30NOTE: (1) 0.13E-03 = 0.13 x 10

-3Sr-930.46E-08Sb-1280.97E-11Ce-1460.10E-08Sr-940.87E-10Sb-1290.87E-07Pr-1430.75E-05Y-900.23E-05Sb-1300.29E-09Pr-1440.56E-05Y-91m0.65E-05Sb-1310.26E-08Pr-1450.56E-06 Y-910.78E-05Sb-1320.27E-10Pr-1460.58E-08Y-920.28E-05Sb-1330.45E-10Nd-1470.26E-05Y-930.18E-05Te-127m0.24E-04Nd-1490.18E-07 Nd-1510.12E-09Sm-1510.44E-09Mn-560.47E-04Pm-1470.93E-06Sm-1530.13E-06Fe-550.57E-04Pm-1490.98E-06Cr-510.11E-04Fe-590.12E-04 Pm-1510.29E-06Mn-540.88E-05Co-580.30E-03Co-600.86E-05TABLE 11.2-12BORON RECOVERY TANK CONCENTRATIONS (µCI/CC)IsotopeDesign (µCi/cc)IsotopeDesign (µCi/cc)IsotopeDesign (µCi/cc)

MPS-3 FSARPage 1 of 1Rev. 30NOTE: (1) 5.90E-02 = 5.90 x 10

-2TABLE 11.2-13ACTIVITY RELEASED TO ATMOSPHERE FROM A RADIOACTIVE LIQUID CONTAINING TANK FAILURE (BORON RECOVERY TANK)IsotopeRadioactivity Released (Ci)Kr-83m 5.90E-02 (1)Xe-131m4.08E-01Xe-133m5.81E-02Xe-1331.08E+00Xe-135m1.15E+00 Xe-1352.77E+00I-1311.28E+00I-1321.67E-01 I-1331.40E+00I-1341.45E-02I-1353.84E-01 MPS-3 FSARPage 1 of 1Rev. 30NOTES:* Nuclide concentration less than 1.0E-15, µCi/cc** 1.72E-12 = 1.72 x 10

-12TABLE 11.2-14RADIOACTIVE CONCENTRATIONS IN GROUNDWATER ENTERING NIANTIC BAY FOLLOWING A RUPTURE OF BORON RECOVERY TANKIsotope Concentration at Entry Point

(µCi/cc)Concentration 1000 Feet from Entry Point (µCi/cc)I-129**

Sr-901.72E-12***Y-901.72E-12*Ru-1063.99E-15*

Rh-1063.99E-15*Te-127m**Te-127**

Cs-1342.27E-087.07E-13Cs-1379.22E-072.86E-11Ba-137m8.48E-072.64E-11 Ce-1441.60E-14*Pr-1441.60E-14*Pm-1471.75E-13*

Sm-151**Mn-544.34E-14*Fe-551.09E-11*

Co-58**Co-603.77E-12*H-32.54E-067.90E-11 MPS-3 FSAR Rev. 30FIGURE 11.2-1 (SHEETS 1-3) P&ID RAD IOACTIVE LIQUID WASTE AND AERATED DRAINS The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS-3 FSAR Rev. 30FIGURE 11.2-2 CONDENSATE DEMINERALIZER LIQUID WASTE The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS3 UFSAR11.3-1Rev. 3011.3GASEOUS WASTE MANAGEMENT SYSTEMS This section describes the capabilities of Millstone3 to control, collect, process, store, and dispose of gaseous radioactive waste generated from normal opera tion and anticipa ted operational occurrences. Section 11.5 describes the process and effluent radiation monitoring systems. The reactor plant gaseous and aerated vents systems are described in Section 9.3.3. The gaseous waste management systems include the radioactive ga seous waste system a nd ventilation systems.

The radioactive gaseous waste system (Figure 11.3-1, Sheets1 and 2) consists of three subsystems:the degasification subsystem, the process gas (hydrogenated) subsystem, and low activity process vent (aerated) subsystem.

In the degasification subsystem, the fluid from the reactor cool ant letdown stream (CHS), or alternately, from the reactor plant gaseous drains system (DGS), is sent to a degasifier, where noncondensable fission product gase s are removed. The remaining li quid may be transferred to the volume control tank (CHS) or to the boron rec overy system (BRS). The normal flowpath is from the reactor coolant letdow n to the volume control tank. The gases are forwarded to the process gas portion of GWS.

In the process gas (hydrogenated) subsystem, the noncondensable fission product gas stream is first dehydrated. Then, radioactive iodine is removed and the activity of the radioactive xenon and krypton is reduced. Finally, the gas is rele ased into the process vent portion of GWS.

In the low activity process vent (aerated) subsystem, aerated a nd hydrogenated gas streams from various plant inputs (including the process ga s portion of GWS) are collected, dehydrated, and discharged to the reactor plant ventilation system (HVR) for release to the environment via the Millstone stack. The gas streams are monitored for radioactivity prior to release.In a separate flow path, relief effluents from the degasifier, the waste evaporator, and the boron evaporator are collected and discharged to the reactor plant ventilation system (HVR) for release to the environment via the turbine building stack.The radioactivity values provided in this section are the design basis values used for the design of the Gaseous Waste System. As such, they are cons idered historical and not subject to future updating. The information is retained to avoid loss of the orig inal design bases. Actual airborne radioactivity release quantities ca n be found in the annual radioactive effluent release reports as submitted to the NRC.11.3.1DESIGN BASES 11.3.1.1Design Objective

The objective of the gaseous waste management syst em is to process and control the release of radioactive gaseous effluents to the site environs so as to maintain the exposure to radioactive gaseous effluents of persons in unrestricted areas to as low a le vel as is reasonably achievable (AppendixI to 10CFR50, May5, 1975). This is to be accomplished while also maintaining MPS3 UFSAR11.3-2Rev. 30 occupational exposure as low as reasonably achi evable and without limiting plant operation or availability.11.3.1.2Design Criteria

The design of the radioactive gaseous waste sy stem and the ventilation systems meet the following criteria.1.Section11.1 discusses the design basis source terms. Expected radioactive gaseous effluents from all sources (Table11.3-1) have been calculated using the data shown in Table11.3-2. These values ar e consistent with NUREG-0017, April 1976.2.The systems have the capability to meet the requirements of 10CFR20 and the dose design objectives specified in AppendixI to 10CFR50, including provisions to treat gaseous radioactive wastes such that:a.The calculated annual total quantity of all radioactive ma terial released from Millstone3 to the atmosphere doe s not result in an estimated annual external dose from gaseous effluents to any individual in unrestricted areas in excess of 5millirems to the total body or 15millirems to the skin.b.The calculated annual total quantit y of all radioact ive iodine and radioactive material in particulate form released from Millstone3 to the atmosphere does not result in an estimated annual dose or dose commitment from such radioactive i odine and radioactive material in particulate form for any individual in an unrestricted area from all pathways of exposure in excess of 15mRem to any organ.c.The concentrations of radioactive materials in gaseous effluents released to an unrestricted area do not exceed the limits in 10CFR20, AppendixB, Table2, Column1. 3.The radioactive gaseous waste system is designed to meet the anticipated processing requirements of the plant. Ade quate capacity is provided to process gaseous wastes during periods when major processing equipment may be down for maintenance (single failures) and during periods of excessive waste generation.4.The system design contains provisions to control leakage and to facilitate operation and maintenance in accordanc e with the guidelines of Regulatory Guide1.143 (formerly Branch Technical Position ETSB11-1, Rev.1) (Section1.8).5.The radioactive gaseous waste system m eets General Design Criteria60 and 64 of AppendixA to 10CFR50 as discussed in Sections3.1.2.60 and 3.1.2.64.

MPS3 UFSAR11.3-3Rev. 3011.3.1.3Cost Benefit EvaluationAppendix11A shows that the systems contain al l items of reasonably demonstrated technology that affect a reduction in dose to the population reasonably expected to be within a 50-mile radius of the plant with a favorable cost benefit.11.3.1.4Equipment Design Criteria Table11.3-3 lists the radioactive gaseous waste system major equipment items. This list includes materials, rates process conditions, and numbe r of units supplied. Equipment and piping are designed and constructed in accordance with the requirements of the applicable codes (Table11.3-4).Table3.2-1 shows the safety classes of the various systems. Seismic category, safety class, quality assurance requirements, and principal construction codes information is contained in Section3.2.

The system is designed to Safety Classification NNS.

The design of the system precludes an explosiv e mixture from accumulating. Since the system operates above atmospheric pressure, in-leakage cannot occur. In strumentation with automatic alarm functions mon itors the concentrations of hydrogen and oxygen in portions of the system having the potential for cont aining explosive mixtures.The radioactive gaseous waste system processes letdown from the chemi cal and volume control system (CHS) or reactor plant gaseous drain through the degasifi er at a maximum rate of 150 gpm. Maximum letdown from the CHS is 120 gpm a nd the maximum flow rate from the gaseous drains system is no more than 150 gpm.The following design features are incorporated to minimize maintenan ce, equipment downtime, leakage, and radioactive gaseous releases. These f eatures facilitate radwas te operation, and assist in maintaining occupational exposures as low as is reasonably achievable.1.Redundant degasifier recirc ulation pumps prevent degasifier outage due to pump failure.2.Components requiring servicing are placed in individual shielded cubicles to minimize personnel exposure during maintenance.3.Leakage from pumps is piped to sumps.4.The radioactive gaseous waste system can be operated locally from the radioactive gaseous waste and process gas treatment control panels.

Conservative analyses of the radioactive ga seous waste system, presented in Section15.7, demonstrate that equipment failure results in doses well within the guidelines of 10CFR100.

MPS3 UFSAR11.3-4Rev. 3011.3.1.5Building Ventilation SystemsFigure11.3-2 gives a composite diagram of ventil ation systems which may release radioactivity during normal operations. These systems are:1.Fuel building ventilation (Section9.4.2) 2.Auxiliary building ventilation (Section9.4.3)3.Turbine building ventilation (Section9.4.4)4.Containment structure ventilation (Section9.4.7)5.Engineered safety features building ventilation (Section9.4.5)6.Waste disposal building ventilation (Section9.4.9)7.Service building ventilation (Section9.4.12)11.3.2SYSTEM DESCRIPTIONSFigures11.3-1, Sheets1& 2 show the piping and instrumentation drawings (P&IDs) of the radioactive gaseous waste system and Figure 9.3-5 shows reactor pl ant gaseous drains.

The appropriate subsections of Section9.4 provide specific co mponent data wi th P&IDs for ventilation systems subject to radioactive release.11.3.2.1Radioactivity Inpu ts and Release Points Radioactivity in process streams is proces sed by the radioactive gaseous waste system. Ventilation releases for buildi ng housing systems which could poten tially be radioactive during normal operations are:1.Ventilation venta.Containmentb.Auxiliary building

• c.Fuel building

• • For calculation purposes, releases from the fu el building, waste disposal building, service building, and engineered safety fe atures building are combined wi th auxiliary building releases, in accordance with NUREG-0017, April 1976.

MPS3 UFSAR11.3-5Rev. 30d.Waste disposal building

• e.Service building

• 2.Engineered safety features building

• 3.Millstone stacka.Radioactive gaseous waste system b.Main condenser air ejector4.Turbine Buildinga.Roof exhausters b.Steam generator blowdown flash tank vent5.Condensate polishing buildinga.Turbine gland sealing system exhaust Radioactivity releases are provided in Tables11.3-5 through 11.3-10.

Building volumes and expected flow rates are provided in Section9.4.11.3.2.2Degasifier Subsystem of Radioactive Gaseous Waste System Reactor coolant letdown, containing dissolved hydrogen and fission gases, is normally directed to the degasifier from the letdown line upstream of the volume control tank in the chemical and volume control system. Alternately, liquid collected by the reactor plant gaseous drains system (Section 9.3.3) may be also direct ed to either the degasifier or the boron recovery system.

Dissolved gases are separated from the liquid in the degasifier.

The degasifier processes reactor coolant letdown continuously except as needed to process gaseous (hydrogenated) drains. However, reactor coolant letdown may bypass the degasifier, if desired. The degasifier design flow of 150gpm exceeds the maximum expected throughput for the liquid portion of the process gas subsystem.

Separation of dissolve d gases at all reactor coolant letdown flow rates is thus ensured. The degasifier operates at approximately 2 psig. If the degasifier is not operating, it may be placed in either the standby mode or in a shutdown condition.

MPS3 UFSAR11.3-6Rev. 3011.3.2.3Process Gas Subsystem of Radioactive Gaseous Waste System The process gas (hydrogenated) portion of the radi oactive gaseous waste system is designed to treat gases stripped from in the reactor c oolant letdown and the reactor plant gaseous (hydrogenated) drains (Figure 11.3-1).Effluent gases from the degasifier contain primarily hydrogen and water vapor. A small amount of nitrogen and traces of xenon, krypton, argon, carbon, and iodine are also present in the effluent gases. These gases and any hydroge nated gas stream from the reac tor plant gaseous vent header are dehumidified (dew point 35

°F) in one of the two redundant process gas refrigerant dryers. Condensation effluent from the drye rs is returned to the suction of the degasifier recirculation pump. The dry stream is passed through and filtered by the am bient temperature process gas charcoal bed adsorbers and one of two redundant HEPA filters. The heat due to ra dioactive decay is small and does not affect the adsorption of noble gases on the charcoal. The charcoal bed adsorbers are designed to provide holdup of most krypton and xenon isotopes long enough in comparison with their half-lives so that passage through the beds will result in the effective removal of these isotopes. In add ition, decontamination of iodine to negligible levels is obtained during passage through the charcoal beds. The char coal is divided evenly between two vertical tanks in series. The tanks are piped so that either one may be bypassed, if necessary, with a corresponding decrease in decay time; however, bypa ss is not anticipated.

The only radioisotope present in any quantity in the predominantly hydrogen stream after the decay period is krypton-85. This processed hydroge nated stream is monitored by the supplementary leak collection monitor and released to the envir onment through the Millstone stack in accordance with technical specifications. Th e normal flow path for the proces sed stream is to the Millstone stack. When a test of the gas at the vent dampers at the discharge of the process vent fans is required, then the process gas flow is monitored and released through the reactor pl ant ventilation vent.Liquid seals are provided in the drain lines from the process ga s precoolers and the process gas water traps to prevent backflow of vapor from the degasifier. The process gas precooler drains continuously at approximately 3 pounds per hour of water at 120

°F. The process gas water trap drains at approximately 0.3 pound per hour at 35

°F. In the event of a sudden vacuum condition at the suction of the degasifier recirculation pump, the seals are temporarily lost. However, the process gas precooler drain seal refills in approximately 10minutes and the process gas water trap seal refills in approximately 100minutes.

A process gas monitor is provide d for the gaseous releases from the process gas portion to monitor radioactivity release to the environment and to automatically isolate the flow from the process gas receiver when a predetermined level is exceeded (Section11.5).11.3.2.4Process Vent Portion of Radioactive Gaseous Waste SystemThe process vent (aerated) portion of the radioactive gaseous wa ste system is designed to collect the low activity aerated gas stre am from the following sources:1.Reactor plant aerated vents system MPS3 UFSAR11.3-7Rev. 302.Reactor plant gaseous vent system3.Condenser air removal system4.Containment vacuum system5.Low activity effluent from the pro cess gas (hydrogenated) portion of the radioactive gaseous waste system6.Radioactive gaseous waste system component purges7.Boron recovery system relief valve discharge8.Liquid waste system relief valve discharges9.Degasifier relief valve discharge The process vents, with the exception of the relief valve vent header, are monitored by the supplementary leak collection monitor and discharged to the Millstone stack. The process vent portion is operated continuously, unless required to be shutdown for maintenance. The relief valve vent header is discharged to the ventilation ve nt where it is monitored and released without filtration.11.3.2.5Steam and Power Conversion System

The main condenser is evacuated by steam jet air ejectors (Section10.4.2). Ai r ejector exhaust is monitored and discharged to the Millstone stack.

The turbine gland seal steam condenser (Section10.4.3) exhausts directly to the atmosphere. These releases are given in Tables11.3-1 and 11.3-11 for expected and design cases, respectively.During the intermittent and hot standby blowdown (open cycle), the steam generator flash tank is vented directly to the atmosphere through a vent on the turbine building roof. Releases are given in Tables11.3-7 and 11.3-10.The quantity of steam released during steam dumps to the atmosphere is provided in Section10.3.

Actual unit trips are expected to be less than those assumed in NUREG-0017 (2turbine trips per year).11.3.2.6System Instrumentation Requirements11.3.2.6.1Radioactive Gaseous Waste System

The process gas radiation moni tor is located downstream of the process gas charcoal bed adsorbers prior to discharge into the process ve nt portion. This monitor automatically isolates flow from the process gas receiver. Radioactive releases from the process vent portion of the MPS3 UFSAR11.3-8Rev. 30 radioactive gaseous waste system are monitored by the supplementary leak collection monitor.

The radiation monitoring system is discussed in Section11.5.

Hydrogen and oxygen analyzers ar e provided to measure the process gas composition and to alarm before a dangerous concentration exists.Temperature and moisture dete ctors and alarms are provided downstream of the process gas refrigerant dryer to assure proper moisture cont rol in the process gas charcoal bed adsorbers.11.3.2.6.2Ventilation SystemsRadiation monitoring of ventilation system effluents is discussed in Section11.5. Instrumentation requirements are discussed in Section9.4.11.3.2.7Seismic Design Provisions of the Radioactive Gaseous Waste System The radioactive gaseous waste system has not been provided with special seismic design. However, the entire portion of the system that provides for treatment of gases stripped from the reactor coolant is located in the auxiliary building, a Seismic CategoryI building (Section3.8).

The ventilation in the auxiliary building system (Section9.4.3) ha s the capability of detecting radioactive gas leakage and filteri ng prior to release. This is an alternate method of meeting the design guidance given in Regulatory Guide1.143.11.3.2.8Quality Control A program is established to ensure that the design, c onstruction, and testing requirements are met.

The following areas are included in the program.1.Design and Procurement Document Control- Procedures are established to ensure that requirements are specified and included in design and procurement documents and that deviations th erefrom are controlled.2.Inspection- A program for inspection of activities affecting quality is established and executed by or for the organization performing the activity to verify conformance with the documented instru ctions, procedures, and drawings for accomplishing the activity.3.Handling, Storage, and Shipping- Proce dures are established to control the handling, storage, shipping, cleaning, and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.4.Inspection, Test, and Operating Status- Pr ocedures are established to provide for the identifications of items which have satisfactorily passed required inspections and tests.

MPS3 UFSAR11.3-9Rev. 305.Corrective Action- Procedures are established to assure that conditions adverse to quality, such as failures, malfunctions, defi ciencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected.11.3.2.9Welding All welding constituting the pressure boundary of pressure retaining components is performed by qualified welders empl oying qualified welding procedures according to Table11.3-4.

Nonconsumable weld inserts are prohibited in process lines unles s they are ground out after the weld is completed.11.3.2.10Materials

Materials for pressure retaini ng components of process systems are selected from those covered by the material specifications listed in SectionII, PartA of the ASME Boiler and Pressure Vessel Code, except that malleable, wrought, cast iron materials, or plastic pipe is not used. The components meet all of the mandato ry requirements of the material specifications with regard to manufacture, examination, repair, test ing, identification, and certification.A description of the ma jor process equipment, including the design temperature and pressure and the materials of construction, is given in Table11.3-3.11.3.2.11Construction of Process Systems Pressure retaining components of process syst ems use welded construction to the maximum practicable extent. Process piping systems include the first root va lve on sample and instrument lines. Process lines are not less than 3/4-inch nom inal pipe size. Sample and instrument lines are not considered as portions of th e process systems. Flanged joints or suitable rapid disconnect fittings are not used except where maintenance requirements clearly indicate that such construction is preferable. Scre wed connections in which thread s provide the only seal are not used. Screwed connections backed up by seal welding or mechanical joints are used only on lines of 3/4-inch nominal pipe size.In lines 3/4-inch or greater, but less than 21/2-inch nominal pipe size, socket type welds are used. In lines 21/2-inch nominal pipe size and larger, pipe welds are of the butt joint type.11.3.2.12System Integrity Testing Completed process systems are pressure tested to the maximum practicable extent. Piping systems are hydrostatically tested in their entirety, using availa ble valves for pressure test boundaries (lines vented to atmosphere do not requi re a pressure test).

Hydrostatic testing of piping systems is performed at a pressure 1.2 times the design pressure and held for a minimum of 30minutes with no leakage indicated. Pneumatic testing may be substituted for hydrostatic testing in accordance with the applicable codes.

MPS3 UFSAR11.3-10Rev. 3011.3.3RADIOACTIVE RELEASESTables11.3-5 thru 11.3-10 give the calculated sour ces of radioactive nuclide inventory released via gaseous effluents. Table11.3-1 gives the expe cted radioactive gaseous isotope releases from each release point assumed in terms of curies per year per nuclide. Table11.3-11 provides the design releases for these release points.Table11.3-2 lists the paramete rs in these calculations.A summary of the estimated expected annual radioactivity doses is presented in Appendix11A. A summary of design release c oncentrations at the site boundary, maximum permissible concentration (MPC) and Fraction of MPCs is presented in Tables 11.3-8, 11.3-9 and 11.3-10.The "design" releases are within the limits of 10CFR20. The doses from "expected" releases are within the numerical design objectives of Appendix I of 10CFR50. Atmospheric diffusion and ground deposition factors used in the dose calculations are discussed in Section2.3.5.Gaseous effluents are discharged to the e nvironment through the following release points:1.The reactor plant ventilation vent2.The turbine building roof exhausters3.The Millstone stack 4.The engineered safety features building exhaust5.The steam generator blowdown flash tank vent6.The turbine gland seal steam condens er exhaust on the condensate polishing building roof Release points are identified on Figure11.3-2.

Exhausts from the auxiliary building, the fu el building, the waste disposal building, the containment purge, the service building, and the gaseous waste process vent are released from the reactor plant ventilation vent. This vent is located on the turbine building 133feet above grade and 157feet-0inches above sea level. The base elevation is 75feet above grade. The square vent cross-sectional dimensions are 10feet by 10feet and the discharge velocity is 3,000feet per minute. The maximum discharge temperature is 104

°F. Containment purge is considered an intermittent release. Others are assumed as continuous.

The turbine building ventilation system exhausts through the turbine building roof exhausters, located on the turbine building r oof. The exhausters are approximately 114feet above grade, 138feet above sea level, and have a base elevation of 107feet ab ove grade. The system exhausts through roof-mounted, mushroom-type hoods with dimensions of 14feet-6inches by MPS3 UFSAR11.3-11Rev. 3014feet-6inches. The discharge velocity is approximately 900feet per minute and the maximum temperature is 104

°F. These releases are assume d to be continuous releases.

During the open cycle blowdown, the steam generato r blowdown flash tank is vented directly to the atmosphere through a vent on the turbine build ing roof. The vent is a 10-inch diameter pipe attached to a roof-mounted silencer. The discharge elevation is approximately 113feet above sea level. Exhaust flow velocity is 8,450feet per minute.

The turbine gland sealing system exhaust is ve nted to the atmosphere at a point above the condensate polishing building. The vent is a 10-inc h diameter pipe at an elevation of 72feet above sea level.

The reactor plant aerated vents, the reactor plant gaseous vents, condenser air removal effluent, radioactive gaseous waste system discharges, steam generator blowdown tank condenser vent, and containment vacuum pump discharge are released through the Millstone stack. The Millstone stack is 375feet above grade, 389feet above sea level, and has a circular orifice with a 7foot inside diameter. The stack discharges at a maximum velocity of 332feet per minute for accident condition and 40feet per minute for normal condi tion (MP2 & MP3) at a maximum temperature of 104°F. These discharges are assume d as continuous releases. The portion of the discharge vent between the ESF building and the stack is underground.

The engineered safety features building ventil ation system exhausts through the ESF building vent located on the east wall. The vent is 12feet-9inches above grade, 36feet-9inches above sea level. The vent cross sectional dimensions are 3feet-0inches by 10feet-0inches and the discharge velocity is 350feet per minute. The maximum discharg e temperature is 104

°F. The release from the ESF building is considered cont inuous and is included as part of the continuous release from the auxiliary building in Table11.3-5.11.3.3.1Radioactive Gaseous Waste System FailureA gaseous waste system failure is postulated to produce a unique unplanned release by a pathway not normally used for planned releases and requiring a reasonable time to detect and take remedial action to terminate the release.

An inadvertent bypass of the proc ess gas charcoal bed adsorbers with continued normal operation of the gaseous waste system for one hour is assumed. This source is then assumed to be continually released to the auxiliary building and then directly to the environment without holdup.

The gaseous waste system design, described in Section 11.3, precludes the single failure of an active component from causing this event, but it is postulated as the design basis accident as prescribed by Branch Technical Position ETSB 11

-5. For this event to occur the two manual bypass valves must fail or be improperly ali gned, and the piping system downstream of the adsorbers must experience a passive failure to re sult in a release directly into the auxiliary building.

MPS3 UFSAR11.3-12Rev. 30 The process gas charcoal bed adso rber tanks are designed in accordance with ASME III, Class 3, and have a 335 psig design pressure. Because the process gas charcoal bed adsorber normal internal pressure is approximately 1 psig, a gross rupture of the tanks is not considered credible.

Radiation monitors are provided fo r the ventilation system in orde r to detect the release of noble gases to the environs from the building ventilation system. However, no credit is taken for normal operating plant systems, instrument ation or controls, nor operation of any engineered safety feature systems to mitigate the consequences of this event. Normally, all noncondensable gases are removed from the reactor coolant letdown stream in the degasifier. For this analysis it is assumed that the activity released following the bypass of process gas charcoal bed adsorbers consists of noble gas activities discharged from the degasifier for 60 minutes and released without benefit of delay in charcoal beds. It is also assumed that a fraction of the noble gases adsorbed in the charcoal beds is released. The fractions re leased were calculated using the model developed for the Fast Flux Test Facility (Underhill). Table 11.3-13 lists the following:

1. Maximum radioisotope inventory in the process gas charcoal bed adsorbers and associated piping.2. Fraction of the charcoal bed ra dioisotope inventory released.3. Sixty-minute discharge from the degasifier.4. Total radioisotope release.Table 11.3-12 gives the assumptions used for anal yzing the postulated bypass of the process gas charcoal bed adsorbers.

The radiological consequences of bypassing the ch arcoal bed adsorbers are listed in Table 15.0-8 based on design release assumptions in Table 11.3-12, the releases in Table 11.3-13 and the X/Q values in Table 15.0-11. The dose met hodology of Appendix 15A is used here.In order to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric uncertainty), each of the noble gas isotopes released in Table 11.3-13 were scaled based on factors determined from the ratio of the primary activit y concentrations (RCS concentration from Table 15.0-10 and Design concentration from Table 11.1-2) to determine the updated doses for the SPU operating condition.

This event will not cause a Condition IV even t as defined in Section 15.0.1. The radiological consequences are well within the guidelines of 10CFR100.11.3.4REFERENCE FOR SECTION 11.3 MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.3-1TOTAL EXPECTED RADI OACTIVE GASEOUS RELEASED TO ATMOSPHERE FROM MILLSTONE 3 (HISTORICAL) NuclideMillstone Stack (Ci/yr) Ventilation Vent (Ci/yr) Turbine Building (Ci/yr) Condensate Polishing Building (1) (Ci/yr) Total (Ci/yr)Kr-83m 2.9E-01 (2) 4.7E-013.1E-051.5E-047.6E-01K-85m1.2E+002.0E+001.3E-046.5E-043.2E+00Kr-852.4E+021.5E+003.0E-061.5E-052.4E+02Kr-878.7E-011.4E+009.2E-054.6E-042.3E+00 Kr-882.5E+004.0E+002.7E-041.3E-036.5E+00Kr-898.2E-021.3E-018.7E-064.3E-052.1E-01Xe-131m2.4E-019.6E-018.1E-064.0E-051.2E+00 Xe-133m5.4E-011.9E+005.9E-053.0E-042.4E+00Xe-1332.3E+011.5E+022.4E-031.2E-021.7E+02Xe-135m2.1E-013.4E-012.2E-051.1E-045.5E-01 Xe-1352.9E+005.0E+003.0E-041.5E-037.9+00Xe-1371.5E-012.3E-011.6E-057.8E-053.8-01Xe-1387.2E-011.1E+007.6E-053.8E-041.8E+00 I-1311.9E-024.6E-027.3E-032.2E-047.3E-02I-1332.8E-026.8E-029.9E-033.1E-041.1E-01Co-580.06.4E-020.00.06.4E-02 Co-600.02.9E-020.00.02.9E-02Mn-540.01.9E-020.00.01.9E-02Fe-590.06.4E-030.00.06.4E-03 Sr-890.01.4E-030.00.01.4E-03Sr-900.02.1E-040.00.02.1E-04Cs-1340.01.9E-020.00.01.9E-02 Cs-1370.03.2E-020.00.03.2E-02C-147.0E+001.0E+000.00.08.0E+00Ar-410.02.5E+010.00.02.5E+01H-30.07.3E+020.00.07.3E+02 MPS-3 FSAR TABLE 11.3-1 Continued Page 2 of 2Rev. 30NOTES:(1) Releases from turbine gland sealing system exhaust.(2) 2.9E-01 = 2.9 x 10

-1.Historical, not subject to future updating. Has been retained to pr eserve original license basis.

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.3-2RADIOACTIVE GASEOUS SOURCE TERM PARAMETERS (HISTORICAL)Plant Capacity Factor0.8Fuel Defects (%)0.12 (Expected) 1 (Design)

Containment Building:1.Noble gas release to containment building (fraction/day of primary coolant activity) 0.012.Iodine release to containment build ing (fraction/day of primary coolant activity)10-53.Purge exhaust ventilation rate (cfm)35,0004.Purge exhaust ventilation time per purge (hr)85.Containment air filtration subsyste m recirculation rate during purge (cfm)12,000 (1)6.Charcoal iodine adsorber depth (in)47.Iodine exhaust filter efficiency (%)95 8.Particulate exhaust filter efficiency (%)959.Number of cold purges/year410.Continuous ventilation exhaust rate (cfm)011.Free containment volume (cu ft) 2.32 x 10 6 Containment Internal Cleanup System:1.Operates prior to purging cold shutdown (hr)162.Operates prior to purging hot shutdown (hr)163.Mixing efficiency (%)70 4.Containment air filtration subsystem recirculation rate prior to purge (cfm)12,000 (1)5.Charcoal iodine adsorber depth (in)46.Iodine filter efficiency (%)907.Particulate filter efficiency (%)90 Auxiliary Building:1.Iodine exhaust filter efficiency (%)

(2)02.Particulate exhaust filter efficiency (%)

(2)0 MPS-3 FSARPage 2 of 3Rev. 303.Primary coolant leakage rate into building (lb/day)1604.Iodine partition factor0.0075Turbine Building:1.No special design to collect valve leakage)2.Steam leakage (lb/hr)1,700

Main Condenser

/Air Ejector:1.Volatile iodine/total iodine in primary system0.052.Volatile iodine is treated as noble gas in steam generator3.Primary to secondary leak rate (lb/day)1370 (Design) 100 (Expected) 4.MC/AE volatile iodine partition factor0.15 (Design)5.Charcoal iodine adsorber depth (in)0.06.Iodine exhaust filter efficiency (%)0.07.Particulate exhaust filter efficiency (%)0.0 8.Volatile iodine condenser bypass fraction0.35Steam Generator Blowdown:1.Flash tank vented to the atmo sphere during open cycle blowdown2.Flash tank iodine partition factor0.053.Hypothetical flow assumptions (see Section 11.2.2.3 for time periods for each release)Hot Standby:150,520 lb/hr1% MSR from four steam generators (37,630 lb/hr per steam generator) Intermittent Blowdown:263,410 lb/hr

1% MSR from three steam gene rators (37,630 lb/hr per steam generator) 4% MSR from one steam generator (150,520 lb/hr)Turbine Gland Sealing System Exhaust:1.Total steam flow rate8,460 lb/hr2.Iodine partition factor0.15Radioactive Gaseous Waste System (Process Gas System):1.Letdown flow to degasifier (lb/hr)35,900TABLE 11.3-2RADIOACTIVE GASEOUS SOURCE TERM PARAMETERS (HISTORICAL)

MPS-3 FSARPage 3 of 3Rev. 30NOTES:(1) This is the flow rate for one fan unit.

The containment air filtration subsystem includes two fan units, which operate simultaneously.(2) Filters installed but not us ed during normal operation. Capabil ity exists to filter exhaust upon high activity in building.

Data and Assumptions from NUREG-0017, April 1976.Historical, not subject to future updating. This table has been retained to preserve original license basis.2.Holdup time prior to charcoal beds (minutes)7.413.Krypton dynamic adsorption coefficient (cm 3/gm)6.34.Xenon dynamic adsorption coefficient (cm 3/gm)1465.System flow rate (scfm) expected/maximum0.3/36.Total mass of charcoal in beds (lb x 1,000)27 (No iodine or particulates are released from system)7.Krypton holdup time in delay bed (hr)1478.Xenon holdup time in delay bed (hr)3,410 (Complete degasification is handled by the same equipment as normal operation)TABLE 11.3-2RADIOACTIVE GASEOUS SOURCE TERM PARAMETERS (HISTORICAL)

MPS-3 FSARPage 1 of 8Rev. 30TABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEMProcess Gas Charcoal Bed Absorbers Parameters Number2 Vessel pressureOperating (psig)1Design (psig)335Vessel temperatureOperating (psig)104Design (psig)150Total weight - one unitEmpty16,650Operating30,150 Construction materialStainless steelDegasifier Number1 Capacity (gpm)150PressureOperating (psig)2 Design (psig)150 + full vac.Temperature Operating (°F)219 Design (°F)366 and 100 Construction materialStainless steelProcess Gas Receiver Number1Pressure MPS-3 FSARPage 2 of 8Rev. 30 Operating (°F)Approximately Atmospheric Design (°F)1,800Temperature Operating (°F)120 Design (°F)450Total weightEmpty (lb)1,755Full (lb)1,762 Construction materialStainless steel Degasifier Recovery Exchangers Shell Side Tube Side Number2Total duty (Btu/hr) 6,300,000Total liquid entering (lb/hr)75,00075,000 PressureOperating - inlet (psig)100150Design (psig)300300Temperature In (°F219115Out (°F)135199Construction materialStainless steelStainless steelTABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 3 of 8Rev. 30Degasifier Feed Preheater Shell Side Tube Side Number1 Total duty (Btu/hr) 4,600,000Total fluid entering (lb/hr)5,40075,000PressureOperating - inlet (psig)14550 Design (psig)300 + full vac.300 + full vac.Temperature In (°F363199Out (°F)363260Construction materialCarbon steelStainless steel Degasifier Condenser Shell Side Tube Side Number1Total duty (Btu/hr)3,219,000 PressureOperating - inlet (psig)1252Design (psig)18550 + full vac.Temperature In (°F95219Out (°F)116190Construction materialCarbon steelStainless steelProcess Gas Compressor (Abandoned in Place, Currently Bypassed)Number2Capacity (scfm)3TABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 4 of 8Rev. 30Pressure - suctionMinimum (psig)14.7 Maximum (psig)240PressureDischarge (psig)75 Design (psia)1,800Discharge temperature Leaving aftercooler, max

(°F)110Process Gas Compressor Aftercooler Number2 Gas flow (cfm)2.4Gas temperatureInlet (°F408Outlet (°F)105Cooling water flow (gpm)2 Cooling water temperatureInlet (°F95Outlet (°F)100PressureOperating (psig)75Design (psia)1,800 Process Gas Prefilter Max Design Refueling Purge Number 2PressureOperating (psig)1 Design (psia)335TemperatureOperating (psig)104TABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 5 of 8Rev. 30Design (psia)150Filtration efficiency (%) (based on DOP test)99.9799.9799.9799.97 Process Gas Prefilter Max Design Refueling Purge Flow rate (scfm) 8 9 3 20 Pressure drop (in H 20)0.50.541Process Gas H2/02 Analyzers H2 Analyzer 02 Analyzer PressureOperating - inlet (psig)1212Minimum operating (psig)22 Maximum operating (psig)4040Design (psig)4040Relief valves set (psig)4040Design Temperature In (°F)104104Out (°F)104104 Analyzers Required Flow (cc/min)normal100-125100-125maximum150150Degasifier Trim Cooler Shell Side Tube Side Number1Total duty (Btu/hr)1,500,000Temperature

In (°F)95135Out (°F)110115PressureOperating inlet (psig)12585TABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 6 of 8Rev. 30Design (psig)175300Construction materialCarbon steelStainless steel Process Gas Precooler Shell Side Tube Side Number2Total duty (Btu/hr)12,900Total fluid entering (lb/hr)27.12,600 PressureOperating inlet (psig)1125Design (psig)335200Temperature In (°F)19095Out (°F)120100 Design (°F)190120Number of passes12Construction materialStainless steelStainless steel Process Gas Glycol Chiller Refrigerant Tube Side Process Gas Tube Side Number2 Total fluid entering (lb/hr)50.7PressureInlet (psig)201 Design (psig)150335Temperature In (°F)34120Out (°F)3035Construction materialStainless steelStainless steelTABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 7 of 8Rev. 30Process Gas Water Trap Number2 PressureOperating (psig)1Design (psig)335Temperature Operating (°F)35 Design (°F)150Construction materialStainless steel Degasifier Recirculation Pump Number2Capacity (gpm) @ 386 ft head150 PressureOperating-discharge (psig)186Design (psig)200Temperature Operating (°F)218 Design (°F)250Construction materialStainless steelProcess Vent Fans Number2Capacity (cfm) @ 20 in water180 Operations pressure - discharge (psig)0.4Operating temperature (°F)50Outlet air velocity (fpm)1,939Process Vent Cooler Shell Side Tube Side Number1TABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 8 of 8Rev. 30Total duty (Btu/hr)148,828Total fluid entering (lb/hr)14,828752.5 PressureOperating - inlet (psig)12514.6Design (psig)15075Temperature In (°F)45180Out (°F)5550Construction materialCarbon steelStainless steelTABLE 11.3-3RADIOACTIVE GASEOUS WASTE SYSTEM MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Portions of the system were procured to ASME SectionIII requirements and are defined in the procurement specification.(2) The overall system classification is NNS.TABLE 11.3-4CODES AND STANDARDSThe radioactive gaseous waste, the reactor plant aerated vents, and the reactor plant gaseous vents system shall be designed and constructed in accordance with the requirements of the following codes and standards:System ComponentSafety ClassCodes and Standards (1) (2)Piping, fittingsNNSASME III, Class 3 ANSI B31.1ValvesNNSASME III, Class 3 ANSI B16.5 ANSI B16.34Adsorbers, degasifier, heat exchangers, filters, water removal equipmentNNSASME VIII, Division 1Pumps and compressorsNNSASME III, Class 3FansNNSManufacturer's standardsFilter assembliesNNSASME VIII, Division 1Instrumentation and controlsN/A MotorsNEMA MG.1 MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.3-5EXPECTED RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE VIA VENTILATION VENT (HISTORICAL)

Nuclide Containment Building (Ci/yr)Auxiliary Building (Ci/yr) (1) Total (Ci/yr) Kr-83m5.7E-054.7E-014.7E-01Kr-85m2.0E-022.0E+002.0E+00Kr-851.4E+004.5E-021.5E+00Kr-877.4E-061.4E+001.4E+00 Kr-885.7E-034.0E+004.0E+00Kr-890.01.3E-011.3E-01Xe-131m8.5E-011.2E-019.6E-01 Xe-133m1.0E+008.7E-011.9E+00Xe-1331.1E+023.6E+011.5E+02X-135m0.03.4E-013.4E-01 Xe-1353.4E-014.7E+005.0E+00Xe-1370.02.3E-012.3E-01Xe-1380.01.1E+001.1E+00 I-1312.8E-064.6E-024.6E-02I-1332.8E-076.8E-026.8E-02Co-583.8E-036.0E-026.4E-02 Co-601.7E-032.7E-022.9E-02Mn-541.1E-031.8E-021.9E-02Fe-593.8E-046.0E-036.4E-03 Sr-898.5E-051.3E-031.4E-03Sr-901.5E-052.0E-042.1E-04Cs-1341.1E-031.8E-021.9E-02 Cs-1371.9E-033.0E-023.2E-02C-141.0E+000.01.0E+00Ar-412.5E+010.02.5E+01H-31.3E+026.0E+027.3E+02 MPS-3 FSAR Table 11.3-5 Continued Page 2 of 2Rev. 30NOTE:(2) In accordance with NUREG-0017, April 1976, releases from the fuel building, waste disposal building, service build ing, and engineered safety features building are combined with auxiliary building releases.

Historical, not subject to future updating. This table has been retain ed to preserve original license basis.

MPS-3 FSARPage 1 of 1Rev. 30Historical, not subject to future updating. This table has been retained to preserve original license basis.TABLE 11.3-6EXPECTED RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE FROM MILLSTONE 3 VIA MILLSTONE STACK (HISTORICAL)

NuclideMain Condenser/ Air Ejector (Ci/yr Radioactive Gaseous Waste System (Ci/yr) Total (Ci/yr) Kr-83m2.9E-010.02.9E-01Kr-85m1.2E+000.01.2E+00 Kr-852.8E-022.4E+022.4E+02Kr-878.7E-010.08.7E-01Kr-882.5E+000.02.5E+00 Kr-898.2E-020.08.2E-02Xe-131m7.3E-021.7E-012.4E-01Xe-133m5.4E-010.05.4E-01 Xe-1332.3E+010.02.3E+01Xe-135m2.1E-010.02.1E-01Xe-1352.9E+000.02.9E+00 Xe-1371.5E-010.01.5E-01Xe-1387.2E-010.07.2E-01I-1311.9E-020.01.9E-02 I-1332.8E-020.02.8E-02Co-580.00.00.0Co-600.00.00.0 Mn-540.00.00.0Fe-590.00.00.0Sr-890.00.00.0 Sr-900.00.00.0Cs-1340.00.00.0Cs-1370.00.00.0 C-140.07.0E+007.0E+00Ar-410.00.00.0H-30.00.00.0 MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.3-7EXPECTED RADIOACT IVE GASEOUS RELEASES TO ATMOSPHERE VIA TURBINE BUILDING (HISTORICAL)NuclideRoof Exhausters (Ci/yr) Steam Generator Blowdown Flash Tank Vent (2) (Ci/yr) Total (Ci/yr) Kr-83m 3.1E-05 (3)0.03.1E-05Kr-85m1.3E-00.01.3E-04Kr-853.0E-060.03.0E-06 Kr-879.2E-050.09.2E-05Kr-882.7E-040.02.7E-04Kr-898.7E-060.08.7E-06 Xe-131m8.1E-060.08.1E-06Xe-133m5.9E-050.05.9E-05Xe-1332.4E-030.02.4E-03 Xe-135m2.2E-050.02.2E-05Xe-1353.0E-040.03.0E-04Xe-1371.6E-050.01.6E-05 Xe-1387.6E-050.07.6E-05I-1312.9E-047.0E-037.3E-03I-1334.2E-049.5E-039.9E-03 Co-580.0Co-600.0Mn-540.0 Fe-590.0Sr-890.0Sr-900.0 Cs-1340.0Cs-1370.0C-140.0Ar-410.0H-30.0 MPS-3 FSAR Table 11.3-7 Continued Page 2 of 2Rev. 30NOTES:(2) Open cycle blowdown.(3) 3.1E-05 = 3.1 x 10

-5.Historical, not subject to future updating. This table has been retain ed to preserve original license basis.

MPS-3 FSARPage 1 of 4Rev. 30TABLE 11.3-8DESIGN RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE VIA VENTILATION VENT (HISTORICAL)A. CURIES PER YEAR RELEASEDNuclide Containment Building (Ci/yr)Auxiliary Building (Ci/yr) (1)Total (Ci/yr)Kr-83m1.2E-039.5E+009.5E+00Kr-85m3.6E-013.6E+013.6E+01Kr-852.3E+017.3E-012.4E+01 Kr-871.4E-042.6E+012.6E+01Kr-881.0E-017.2E+017.2E+01Kr-890.02.2E+002.2E+00 Xe-131m1.7E+002.4E-011.9E+00Xe-133m1.5E+011.3E+012.8E+01Xe-1331.7E+035.6E+022.3E+03 Xe-135m0.02.4E+012.4E+01Xe-1357.9E+001.1E+021.2E+02Xe-1370.03.5E+003.5E+00 Xe-1380.01.3E+011.3E+01I-1312.6E-054.2E-014.2E-01I-1332.7E-066.6E-016.6E-01 Co-583.8E-036.0E-026.4E-02Co-601.7E-032.7E-022.9E-02Mn-541.1E-031.8E-021.9E-02 Fe-593.8E-046.0E-036.4E-03Sr-899.9E-041.5E-021.6E-02Sr-902.6E-043.4E-033.7E-03 Cs-341.3E-022.2E-012.3E-01Cs-1371.7E-012.6+002.8E+00C-141.0E+000.01.0E+00Ar-412.5E+010.02.5E+01H-34.6E+022.1E+032.6E+03 MPS-3 FSARPage 2 of 4Rev. 30B. CONCENTRATION AND FRACTION OF MPC (2) AT SITE BOUNDARY DUE TO RELEASES FROM CONTAINMENT PURGE (3) VIA VENTILATION VENT:Nuclide Concentration

(µCi/cc)MPC Value

(µCi/cc) Fraction of MPC Kr-83m 1.02E-15 (4)3.0E-083.40E-08Kr-85m3.06E-131.0E-073.06E-06Kr-851.95E-113.0E-076.52E-05 Kr-871.19E-162.0E-085.95E-09Kr-888.50E-142.0E-084.25E-06Kr-890.03.0E-080.0 Xe-131m1.44E-124.0E-073.61E-06Xe-133m1.27E-113.0E-074.25E-05Xe-1331.44E-93.0E-074.82E-03 Xe-135m0.03.0E-080.0Xe-1356.71E-121.0E-076.71E-05Xe-1370.03.0E-080.0 Xe-1380.03.0E-080.0I-1312.21E-171.0E-102.21E-07I-1332.29E-184.0E-105.74E-09 Co-583.23E-152.0E-091.61E-06Co-601.44E-153.0E-104.82E-06Mn-549.35E-161.0E-099.35E-07 Fe-593.23E-162.0E-091.61E-07Sr-898.41E-163.0E-102.80E-06Sr-902.21E-163.0E-117.37E-06 Cs-1341.10E-144.0E-102.76E-05Cs-1371.44E-135.0E-102.89E-04C-148.50E-131.0E-078.50E-06Ar-412.12E-114.0E-085.31E-04H-33.91E-102.0E-071.95E-03 MPS-3 FSARPage 3 of 4Rev. 30C. CONCENTRATION AND FRACTION OF MPC AT SITE BOUNDARY DUE TO RELEASES FROM AUXILAIRY BUILDING (1) , (5) VIA VENTILATION VENT:Nuclide Concentration

(µCi/cc)MPC Value

(µCi/cc) Fraction of MPC Kr-83m1.20E-12 3.0E-084.01E-05Kr-85m4.55E-121.0E-074.55E-05 Kr-859.24E-143.0E-073.08E-07Kr-873.29E-122.0E-081.64E-04Kr-889.11E-122.0E-084.55E-04 Kr-892.78E-133.0E-089.28E-06Xe-131m3.04E-144.0E-077.59E-08Xe-133m1.64E-123.0E-075.48E-06 Xe-1337.09E-113.0E-072.36E-04Xe-135m3.04E-123.0E-081.01E-04Xe-1351.39E-111.0E-071.39E-04 Xe-1374.43E-133.0E-081.48E-05Xe-1381.64E-123.0E-085.48E-05I-1315.31E-141.0E-105.31E-04 I-1338.35E-144.0E-102.09E-04Co-587.59E-152.0E-093.80E-06Co-603.42E-153.0E-101.14E-05 Mn-542.28E-151.0E-092.28E-06Fe-597.59E-162.0E-093.80E-07Sr-891.90E-153.0E-106.33E-06 Sr-904.30E-163.0E-111.43E-05Cs-1342.78E-144.0E-106.96E-05Cs-1373.29E-135.0E-106.58E-04 C-140.01.0E-070.0Ar-410.04.0E-080.0H-32.66E-102.0E-071.33E-03 MPS-3 FSAR TABLE 11.3-8 (CONTINUED)

Page 4 of 4Rev. 30NOTES:(1) Releases from the fuel build ing, waste disposal building, se rvice building, and engineered safety features building, are combined with auxiliary building releases.(2) MPC values are taken from 1 0 CFR 20, Appendix B, Table II, Column I. When releasing radioactivity in air or water , instantaneous release rate limits are based on the values in Appendix B of the version of 10 CFR 20 prior to January 1, 1994 (Reference 11.3-1).(3) Containment purge is an in termittent release from ventilatio n vent, and the applicable X/Q is 2.68 x 10

-5 sec/m 3 , which is the historical value used in the original estimates of radioactive concentrations at the Site Boundary.(4) 1.02E-15 = 1.02 x 10

-15.(5) Auxiliary building releases are cont inuous, and the applicable X/Q is 3.99 x 10

-6 sec/m 3 , which is the historical value used in the original estimates of radioactive concentrations at the Site Boundary.

Historical, not subject to future updating. Has been retained to pr eserve original license basis.

MPS-3 FSARPage 1 of 3Rev. 30TABLE 11.3-9DESIGN RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE FROM MILLSTONE 3 VIA MILLSTONE STACK (HISTORICAL)A. CURIES PER YEAR RELEASEDNuclideMain Condenser / Air Ejector (Ci/yr)

Radioactive Gaseous Waste System (Ci/yr) Total (Ci/yr)Kr-83m8.2E+010.08.2E+01Kr-85m3.2E+020.03.2E+02Kr-856.3E+004.0E+034.0E+03 Kr-872.2E+020.02.2E+02Kr-886.2E+020.06.2E+02Kr-891.9E+010.01.9E+01 Xe-131m2.1E+003.5E-012.5E+00Xe-133m1.1E+020.01.1E+02Xe-1334.8E+032.4E-024.8E+03 Xe-135m2.1E+020.02.1E+02Xe-1359.3E+020.09.3E+02Xe-1373.0E+010.03.0E+01 Xe-1381.1E+020.01.1E+02I-1312.3E+000.02.3E+00I-1333.7E+000.03.7E+00 Co-580.00.00.0Co-600.00.00.0Mn-540.00.00.0 Fe-590.00.00.0Sr-890.00.00.0Sr-900.00.00.0 Cs-1340.00.00.0Cs-1370.00.00.0C-140.07.0E+007.0E+00Ar-410.00.00.0 MPS-3 FSARPage 2 of 3Rev. 30H-30.00.00.0B. CONCENTRATIONS AND FRACTION OF MPC (1) AT SITE BOUNDARY DUE TO RELEASES VIA MILLSTONE STACK (2):NuclideConcentration (µCi/cc) MPC Value

(µCi/cc) Fraction of MPC Kr-83m 3.06E-14 (4) 3.0E-081.02E-06Kr-85m1.20E-131.0E-071.20E-06K-851.50E-123.0E-074.99E-06Kr-878.23E-142.0E-084.12E-06 Kr-882.32E-132.0E-081.16E-05Kr-897.11E-153.0E-082.37E-07Xe-131m9.35E-164.0E-072.34E-09 Xe-133m4.12E-143.0E-071.37E-07Xe-1331.80E-123.0E-075.99E-06Xe-135m7.86E-143.0E-082.62E-06 Xe-1353.48E-131.0E-073.48E-06Xe-1371.12E-143.0E-083.74E-07Xe-1384.12E-143.0E-081.37E-06 I-1318.61E-161.0E-108.61E-06I-1331.38E-154.0E-103.46E-06Co-580.02.0E-090.0 Co-600.03.0E-100.0Mn-540.01.0E-090.0Fe-590.02.0E-090.0Sr-890.03.0E-100.0Sr-900.03.0E-110.0 Cs-1340.04.0E-100.0A. CURIES PER YEAR RELEASEDNuclideMain Condenser / Air Ejector (Ci/yr)

Radioactive Gaseous Waste System (Ci/yr) Total (Ci/yr)

MPS-3 FSAR TABLE 11.3-9 CONTINUEDPage 3 of 3Rev. 30NOTES:(1) MPC values are taken from 10 CFR 20, Appendix B, Table II, Column I. When releasing radioactivity in air or water , instantaneous release rate limits are based on the values in Appendix B of the version of 10 CFR 20 prior to January 1, 1994 (Reference 11.3-1)(2) This is continuous, and th e applicable X/Q is 1.18 x 10

-8 sec/m 3 , which is the historical value used in the original es timates of radioactive concentrations at the Site Boundary.(3) 3.06E-14 = 3.06 x 10

-14.Historical, not subject to future updating. This table has been retain ed to preserve original license basis.Cs-137 0.0 5.0E-10 0.0C-14 2.62E-15 1.0E-07 2.62E-08Ar-41 0.0 4.0E-08 0.0H-3 0.0 2.0E-07 0.0B. CONCENTRATIONS AND FRACTION OF MPC (1) AT SITE BOUNDARY DUE TO RELEASES VIA MILLSTONE STACK (2):NuclideConcentration (µCi/cc) MPC Value

(µCi/cc) Fraction of MPC MPS-3 FSARPage 1 of 4Rev. 30TABLE 11.3-10 DESIGN RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE VIA TURBINE BUILDING ROOF (HISTORICAL) A. CURIES PER YEAR RELEASEDNuclide Roof Exhausters (Ci/yr) Steam Generator Blowdown Flash Tank Vent (1) (Ci/yr)Total (Ci/yr)Kr-83m 8.7E-03 0.0 8.7E-03Kr-85m 3.3E-02 0.0 3.3E-02Kr-85 6.7E-04 0.0 6.7E-04Kr-87 2.4E-02 0.0 2.4E-02Kr-88 6.5E-02 0.0 6.5E-02Kr-89 2.1E-03 0.0 2.1E-03 Xe-131m 2.2E-04 0.0 2.2E-04 Xe-133m 1.2E-02 0.0 1.2E-02 Xe-133 5.1E-01 0.0 5.1E-01 Xe-135m 2.2E-02 0.0 2.2E-02 Xe-135 1.0E-01 0.0 1.0E-01 Xe-137 3.2E-03 0.0 3.2E-03 Xe-138 1.2E-02 0.0 1.2E-02 I-131 3.7E-02 8.65E-01 9.0E-01 I-133 5.6E-02 1.30E+00 1.4E+00Co-58 0.0 0.0 0.0Co-60 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0Fe-59 0.0 0.0 0.0Sr-89 0.0 0.0 0.0Sr-90 0.0 0.0 0.0 Cs-134 0.0 0.0 0.0 Cs-137 0.0 0.0 0.0C-14 0.0 0.0 0.0 MPS-3 FSAR Table 11.3-10 ContinuedPage 2 of 4Rev. 30Ar-41 0.0 0.0 0.0H-3 0.0 0.0 0.0B. CONCENTRATIONS AND FRACTION OF MPC (2) AT SITE BOUNDARY DUE TO RELEASES FROM TURBINE BUILDING VENTILATION (3):Nuclide Concentration (µCi/cc) MPC Value (µCi/cc)Fraction of MPCKr-83m 2.79E-15 (4) 3.0E-08 9.29E-08Kr-85m 1.06E-14 1.0E-07 1.06E-07Kr-85 2.15E-16 3.0E-07 7.15E-10Kr-87 7.69E-15 2.0E-08 3.84E-07Kr-88 2.08E-14 2.0E-08 1.04E-06Kr-89 6.73E-16 3.0E-08 2.24E-08 Xe-131m 7.05E-17 4.0E-07 1.76E-10 Xe-133m 3.84E-15 3.0E-07 1.28E-08 Xe-133 1.63E-13 3.0E-07 5.44E-07 Xe-135m 7.05E-15 3.0E-08 2.35E-07 Xe-135 3.20E-14 1.0E-07 3.20E-07 Xe-137 1.02E-15 3.0E-08 3.42E-08 Xe-138 3.84E-15 3.0E-08 1.28E-07 I-131 1.18E-14 1.0E-10 1.18E-04 I-133 1.79E-14 4.0E-10 4.48E-05Co-58 0.0 2.0E-09 0.0Co-60 0.0 3.0E-10 0.0 Mn-54 0.0 1.0E-09 0.0Fe-59 0.0 2.0E-09 0.0A. CURIES PER YEAR RELEASED Nuc lide Roof Exhausters (Ci/yr) Steam Generator Blowdown Flash Tank Vent (1) (Ci/yr)Total (Ci/yr)

MPS-3 FSAR Table 11.3-10 ContinuedPage 3 of 4Rev. 30Sr-89 0.0 3.0E-10 0.0Sr-90 0.03.0E-11 0.0 Cs-134 0.0 4.0E-10 0.0 Cs-137 0.0 5.0E-10 0.0C-14 0.0 1.0E-07 0.0Ar-41 0.0 4.0E-08 0.0H-3 0.0 2.0E-07 0.0C. CONCENTRATIONS AND FRACTION OF MPC AT SITE BOUNDARY DUE TO RELEASES FROM STEAM GENERATOR BLOWDOWN (5):Nuclide Concentration (µCi/cc) MPC Value (µCi/cc)Fraction of MPCKr-83m 0.0 3.0E-08 0.0Kr-85m 0.0 1.0E-07 0.0Kr-85 0.0 3.0E-07 0.0Kr-87 0.0 2.0E-08 0.0Kr-88 0.0 2.0E-08 0.0Kr-89 0.0 3.0E-08 0.0 Xe-131m 0.0 4.0E-07 0.0 Xe-133m 0.0 3.0E-07 0.0 Xe-133 0.0 3.0E-07 0.0 Xe-135m 0.0 3.0E-08 0.0 Xe-135 0.0 1.0E-07 0.0 Xe-137 0.0 3.0E-08 0.0 Xe-138 0.0 3.0E-08 0.0 I-131 1.33E-12 1.0E-10 1.33E-02 I-133 2.00E-12 4.0E-10 4.99E-03Co-58 0.0 2.0E-09 0.0B. CONCENTRATIONS AND FRACTION OF MPC (2) AT SITE BOUNDARY DUE TO RELEASES FROM TURBINE BUILDING VENTILATION (3):NuclideConcentration (µCi/cc) MPC Value (µCi/cc)Fraction of MPC MPS-3 FSAR Table 11.3-10 ContinuedPage 4 of 4Rev. 30NOTES:(1) Open cycle blowdown.(2) MPC values are taken from 10 CFR 20, Appendix B, Table II, Column I. When releasing radioactivity in air or water , instantaneous release rate limits are based on the values in Appendix B of the version of 10 CFR 20 prior to January 1, 1994 (Reference 11.3-1.)(3) This is a continuous release, and the applicable X/Q is 1.01 x 10

-5 sec/m 3 , which is the historical value used in the original estimates of radioactive concentrations at the Site Boundary.(4) 2.79E-15 = 2.79 x 10

-15.(5) This is an intermittent release, and the applicable X/Q is 4.84 x 10

-5 sec/m 3 , which is the historical value used in the original estimates of radioactive concentrations at the Site Boundary.Historical, not subject to future updating. This table has been retain ed to preserve original license basis.Co-60 0.0 3.0E-10 0.0 Mn-54 0.0 1.0E-09 0.0Fe-59 0.0 2.0E-09 0.0Sr-89 0.0 3.0E-10 0.0Sr-90 0.03.0E-11 0.0 Cs-134 0.0 4.0E-10 0.0 Cs-137 0.0 5.0E-10 0.0C-14 0.0 1.0E-07 0.0Ar-41 0.0 4.0E-08 0.0H-3 0.0 2.0E-07 0.0C. CONCENTRATIONS AND FRACTION OF MPC AT SITE BOUNDARY DUE TO RELEASES FROM STEAM GENERATOR BLOWDOWN (5):NuclideConcentration (µCi/cc) MPC Value (µCi/cc)Fraction of MPC MPS-3 FSARPage 1 of 5Rev. 30TABLE 11.3-11 DESIGN RADIOACTIVE GASEOUS RELEASES TO ATMOSPHERE FROM MILLSTONE 3 (HISTORICAL)A. TOTAL RELEASES:Nuclide Millstone Stack (Ci/yr)Ventilation Vent (Ci/yr)Turbine Building (Ci/yr) Condensate Polishing Building (1) (Ci/yr)Total (Ci/yr)

K-83m 8.2E+01 (2)9.5E+00 8.7E-03 4.3E-02 9.2E+01Kr-85m 3.2E+02 3.6E+01 3.3E-02 1.7E-01 3.6E+02K-85 4.0E+03 2.4E+01 6.7E-04 3.3E-03 4.0E+03Kr-87 2.2E+02 2.6E+01 2.4E-02 1.2E-01 2.5E+02Kr-88 6.2E+02 7.2E+01 2.4E-02 3.3E-01 6.9E+02Kr-89 1.9E+01 2.2E+00 2.1E-03 1.0E-02 2.1E+01 Xe-131m 2.5E+00 1.9E+00 2.2E-04 1.1E-03 4.4E+00 Xe-133m 1.1E+02 2.8E+01 1.2E-02 5.9E-02 1.4E+02 Xe-133 4.8E+03 2.3E+03 5.1E-01 2.5E+00 7.1E+03 Xe-135m 2.1E+02 2.4E+01 2.2E-02 1.1E-01 2.3E+02 Xe-135 9.3E+02 1.2E+02 1.0E-01 5.0E-01 1.1E+03 Xe-137 3.0E+01 3.5E+00 3.2E-03 1.6E-02 3.4E+01 Xe-138 1.1E+02 1.3E+01 1.2E-02 5.8E-02 1.2E+02I-131 2.3E+00 4.2E-01 9.0E-01 2.8E-02 3.6E+00I-133 3.7E+00 6.6E-01 1.4E+00 4.2E-02 5.8E+00Co-58 0.0 6.4E-02 0.0 0.0 6.4E-02Co-60 0.0 2.9E-02 0.0 0.0 2.9E-02 Mn-54 0.0 1.9E-02 0.0 0.0 1.9E-02Fe-59 0.0 6.4E-03 0.0 0.0 6.4E-03Sr-89 0.0 1.6E-02 0.0 0.0 1.6E-02Sr-90 0.0 3.7E-03 0.0 0.0 3.7E-03 Cs-134 0.0 2.3E-01 0.0 0.0 2.3E-01 Cs-137 0.0 2.8E+00 0.0 0.0 2.8E+00C-14 7.0E+00 1.0E+00 0.0 0.0 8.0E+00Ar-41 0.0 2.5E+01 0.0 0.0 2.5E+01 MPS-3 FSAR TABLE 11.3-11 CONTINUEDPage 2 of 5Rev. 30H-3 0.0 2.6E+03 0.0 0.0 2.6E+03A. TOTAL RELEASES:

Nuc lide Millstone Stack (Ci/yr)Ventilation Vent (Ci/yr)Turbine Building (Ci/yr) Condensate Polishing Building (1) (Ci/yr)Total (Ci/yr)

MPS-3 FSAR TABLE 11.3-11 CONTINUEDPage 3 of 5Rev. 30B. CONCENTRATIONS AND FRACTION OF MPC (3) AT SITE BOUNDARY DUE TO RELEASES FROM CONDEN SATE POLISHING BUILDING (4):IsotopeConcentration (µCi/cc) MPC Value (µCi/cc) Fraction of MPC K-83m3.08E-143.0E-081.03E-06Kr-85m1.22E-131.0E-071.22E-06Kr-852.36E-153.0E-077.88E-09 Kr-878.60E-142.0E-084.30E-06Kr-882.36E-132.0E-081.18E-05K-897.17E-153.0E-082.39E-07 Xe-131m7.88E-164.0E-071.97E-09Xe-133m4.23E-143.0E-071.41E-07Xe-1331.79E-123.0E-075.97E-06 Xe-135m7.88E-143.0E-082.63E-06Xe-1353.58E-131.0E-073.58E-06Xe-1371.15E-143.0E-083.82E-07Xe-1384.16E-143.0E-081.39E-06I-1312.01E-141.0E-102.01E-04 I-1353.01E-144.0E-107.52E-05Co-580.02.0E-090.0Co-600.03.0E-100.0 Mn-540.01.0E-090.0Fe-590.02.0E-090.0Sr-890.03.0E-100.0 Sr-900.03.0E-110.0Cs-1340.04.0E-100.0Cs-1370.05.0E-100.0 C-140.01.0E-070.0Ar-410.04.0E-080.0H-30.02.0E-070.0 MPS-3 FSAR TABLE 11.3-11 CONTINUEDPage 4 of 5Rev. 30C. FRACTION OF MPC AT SITE BOUNDARY DUE TO ALL RELEASES (5):NuclideMillstone Stack Ventilation Vent Turbine Building Condensate Polishing Building Total Fraction of MPCKr-83m 1.02E-06 4.01E-05 9.29E-08 1.03E-06 4.22E-05Kr-85m 1.20E-06 4.86E-05 1.06E-07 1.22E-065.11E-05Kr-85 4.99E-06 6.55E-05 7.15E-10 7.88E-09 7.05E-05Kr-87 4.12E-06 1.64E-04 3.84E-07 4.30E-06 1.73E-04Kr-88 1.16E-05 4.59E-04 1.04E-06 1.18E-05 4.83E-04Kr-89 2.37E-07 9.28E-06 2.24E-08 2.39E-07 9.78E-06 Xe-131m 2.34E-09 3.69E-06 1.76E-10 1.97E-09 3.69E-06 Xe-133m 1.37E-07 4.80E-05 1.28E-08 1.41E-07 4.83E-05 Xe-133 5.99E-06 5.06E-03 5.44E-07 5.97E-06 5.07E-03 Xe-135m 2.62E-06 1.01E-04 2.35E-07 2.63E-06 1.06E-04 Xe-135 3.48E-06 2.06E-04 3.20E-07 3.58E-06 2.13E-04 Xe-137 3.74E-07 1.48E-05 3.42E-08 3.82E-07 1.56E-05 Xe-138 1.37E-06 5.48E-05 1.28E-07 1.39E-06 5.77E-05 I-131 8.61E-06 5.31E-04 1.34E-02 2.01E-04 1.41E-02 I-133 3.46E-06 2.09E-04 5.03E-03 7.52E-05 5.32E-03Co-58 0.0 5.41E-06 0.0 0.0 5.41E-06Co-60 0.0 1.62E-05 0.0 0.0 1.62E-05 Mn-54 0.0 3.22E-06 0.0 0.0 3.22E-06Fe-59 0.0 5.41E-07 0.0 0.0 5.41E-07Sr-89 0.0 9.13E-06 0.0 0.0 9.13E-06Sr-90 0.0 2.17E-05 0.0 0.0 2.17E-05 Cs-134 0.0 9.72E-05 0.0 0.0 9.72E-05 Cs-137 0.0 9.47E-04 0.0 0.0 9.47E-04C-14 2.62E-08 8.50E-06 0.0 0.0 8.53E-06Ar-41 0.0 5.31E-04 0.0 0.0 5.31E-04H-3 0.0 3.28E-03 0.0 0.0 3.28E-03 MPS-3 FSAR TABLE 11.3-11 CONTINUEDPage 5 of 5Rev. 30NOTES:(1) Releases from turbine gland sealing system exhaust.(2) 8.2E+01 = 8.2 x 10 1.(3) MPC values are taken from 10 CFR 20, Appendix B, Table II, Column I. When releasing radioactivity in air or water , instantaneous release rate limits are based on the values in Appendix B of the version of 10 CFR 20 prior to January 1, 1994 (Reference 11.3-1).(4) This is a continuous release, and the applicable X/Q is 2.26 x 10

-5 sec/m 3 , which is the historical value used in the original estimates of radioactive concentrations at the Site Boundary.(5) The values provided in this tabl e are based on historical X/Q values.

Historical, not subject to future updating. This table has been retain ed to preserve original license basis.

MPS-3 FSARPage 1 of 1Rev. 30NOTE:(1) The bypass of the process gas charcoal bed adsorbers releases the gaseous activity discharged from the degasifier for 60minutes and a fraction of the activity on the beds.TABLE 11.3-12ASSUMPTIONS USED FOR THE PROCESS GAS CHARCOAL BED ADSORBER BYPASS ANALYSIS (1)Design ReleaseLetdown flow to degasifier (gpm)75Reactor coolant activityTable 11.1-2

Charcoal bed adsorber holdup timeKr (days)6.1Xe (days)142Fraction of noble gas released from bedTable 11.3-13Duration of release (min.)60 MPS-3 FSARPage 1 of 1Rev. 30NOTE:(1) 7.65E+00 = 7.65 x 10 0TABLE 11.3-13RADIOISOTOPE RELEASES FROM THE PROCESS GAS CHARCOAL BED ADSORBER AND ASSOCIATED PIPING Isotope DesignDischarge from the degasifier (Ci)Inventory in the Charcoal Bed (Ci)Fraction Released from BedTotal Release (Ci)Kr-83m 7.65E+00 (1)2.06E+011.0002.83E+01Kr-85m2.93E+011.86E+010.9982.15E+02Kr-855.88E-018.64E+010.1091.00E+01Kr-872.11E+013.84E+011.0005.95E+01Kr-885.76E+012.32E+021.0002.90E+02 Kr-891.80E+001.38E-011.0001.94E+00Xe-131m1.91E-017.91E+010.0957.69E+00Xe-133m1.04E+0.18.15E+020.4103.44E+02 Xe-1334.48E+028.27E+040.2021.72E+04Xe-135m1.96E+017.39E+001.0002.70E+01Xe-1358.74E+011.16E+031.0001.25E+03 Xe-1372.84E+002.62E-011.0003.10E+00Xe-1381.02E+013.48E+001.0001.37E+01 MPS-3 FSAR Rev. 30FIGURE 11.3-1 (SHEETS 1-2) P&ID RADIOACTIVE GASEOUS WASTE SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS3 UFSAR11.4-1Rev. 3011.4SOLID WASTE MANAGEMENT The radioactive solid waste system is designed to collect, hold, process, dewater or solidify, package, handle, and temporarily store radioactive materials prior to their shipment offsite and ultimate disposal.The radioactivity values provided in this section are the design basis values used for the design of the Solid Waste System. As such, they are considered historical and not subject to future updating.

The information is retained to avoid loss of the original design bases. Volumes and radioactivity content, including specific nuclide percentages of actual shipments can be found in the annual radioactive effluent release re ports as submitted to the NRC.11.4.1DESIGN BASESThe radioactive solid waste system is designe d in accordance with the following criteria.1.The system design parameters are ba sed on radionuclide concentrations and volumes consistent with reactor operating experience for similar designs and with the source terms of Section 11.1.2.All wet solid wastes including bulk liquids , sludges, or spent resin are dewatered, solidified or otherwise treated (as require d). Procedures are used to ensure the absence of free liquid in the containers and control other appropriate waste form characteristics. The plant is compatible with any mobile processing equipment.

Plant procedures are used to ensure that mobile equipment meets federal and state disposal regulations. Dewa tering equipment is portabl e and described in FSAR Section 11.4.2.2 and Figure 11.4-1 (2 of 2).Processing equipment is sized to handle the design inputs without the need to ship bulk liquids. In-plant waste storage facilities provide sufficient temporary storage capacity to allow time for shipping delays. Tanks accumulating spent resins have the capability of accommodating at least 60 days' waste generation at normal generation rates. Temporary storage capac ity exceeds 30 days' waste generation at expected generation rates. Longer term temporary storage is described in FSAR Section 11.4.2.5.3.Solid waste containers, shipping casks, and methods of packaging meet applicable federal regulations, e.g., 10 CFR Part 71, and wastes are to be shipped to a licensed burial site in accordance with appl icable NRC, e.g., 10 CFR Part 61, and Department of Transportation regulat ions, e.g., 49 CFR 171-178. Solid waste treatment design is in compliance with the relevant requirements of 10 CFR Part 20, sections 105 and 106 (version prior to January 1, 1994) as it relates to radioactivity in effluents to unrestricted areas.

MPS3 UFSAR11.4-2Rev. 304.Permanent plant components and piping sy stems are designed in accordance with the provisions of Regulatory Guide 1.143 and Branch Technical Position ETSB 11-3.5.The permanent plant system contains pr ovisions to reduce leakage and facilitate operations and maintenance in accordance with the provisi ons of Regulatory Guide 1.143 and Branch Technical Position, ETSB 11-3 (July 1981).6.The filling of containers, the dewatering, the solidification, and/or the storage of radioactive solid wastes conforms to 10 CFR 20 and 10 CFR 50 requirements and Regulatory Guide 8.8 guidelines in terms of "as low as is reasonably achievable" (ALARA) doses to plant personnel and the general public.7.The temporary waste storage facilities in the waste storage building are shielded to provide protection of operating person nel in accordance with the radiation protection design consider ations in Section 12.1.8.Radiation protection pers onnel conducting surveys use por table radiation detectors to determine radiation levels outside and inside shielded areas. These surveys are performed to ensure that area guidelines are met and to ensure adequate personnel protection through area posting and access limitations.9.This system is not safety-related and is classified as n onnuclear safety (NNS).10.The portion of the solid waste building' s foundation and adjacent walls up to a height sufficient to contain the liquid i nventory in the building is designed to the seismic criteria used for Millstone 3 and to the quality assurance criteria of Section 6 of Regulatory Guide 1.143.11.Portions of the system that handle radi oactive liquid waste meet the applicable design bases of Section 11.2.1.12.The system is designed to process, handle, and store the waste types, and quantities described in Section 11.4.2 below, which are generated as a result of normal operation.13.The system is able to reduce the volum e of selected input streams to minimize packaged quantities for transport and disposal.11.4.2SYSTEM DESCRIPTIONThe radioactive solid waste system is shown on Figure 11.4-1.

The layout of the solid waste bui lding, including packaging, storage, and shipping areas, is shown in Figure 3.8-74.

MPS3 UFSAR11.4-3Rev. 30 Radioactive solid waste system components are listed in Table 11.4-2, which summarizes design and operating conditions.11.4.2.1System Inputs Materials handled as solid wastes may include any of the following: concentrated waste solutions from the waste evaporator, con centrated boric acid discarded from the boron evaporator in the boron recovery system (Section 9.3.5), spent resin from radioactive proce ss demineralizers and exchangers, spent filter cartri dges, and miscellaneous sludges.Figure 11.4-2 is a flow chart of the expected quantities and activity levels of the radioactive solid waste generated by the plant. Gross activities and weights or volumes for radioactive solid waste sources are also given on Figure 11.4-2. Figure 11.4-3 provide s the design volumes and activity levels of radioactive solid wastes and waste sources.11.4.2.1.1Spent ResinsFigure 11.4-2 provides the estimated volumes of spent demineralize r and ion exchanger resins per year. These resins come from a variety of different services, a nd the total volume estimated is based on the individual resin bed volumes and the expected frequency of replacement (Table 11.4-1).11.4.2.1.2Waste Evaporator Bottoms Evaporator bottoms from the waste evaporator in the radioactive liquid waste system may be transferred to the solid waste system. The estimated volume of bottoms solution to be shipped off site is given on Figure 11.4-2. The calculated ac tivity of these bottoms, also shown on Figure 11.4-2, is based on the input of the radioactive liquid waste system (Section 11.2).11.4.2.1.3Regenerant Chemical Evaporat or Bottoms (Removed From Service)11.4.2.1.4Boron Evaporator Bottoms The boron evaporator in the boron recovery syst em processes reactor coolant letdown to the boron recovery system for separation of boric acid and water. The estimated volume given on Figure 11.4-2 for the boron evaporator bottoms is a 1 year average of the contents requiring processing for eventual off site disposal. The activity shown for these bottoms, given on Figure 11.4-2, is based on the expected performance of the boron recovery system (Section 9.3.5). Boric acid may be either recycled to the plant or transferred to the solid waste system depending upon operational considerations.11.4.2.1.5Miscellaneous Radioactive Solid Wastes

It is estimated that approximately 4,000 cubic feet per year of additional waste requiring disposal is processed as radioactive solid waste. This volume of spent filters, c ontaminated cloths, and MPS3 UFSAR11.4-4Rev. 30 other radioactive material and their activity, given on Figure 11.4-2, was originally estimated from operating experience at other nuclear facilities.This category includes waste charcoal absorber me dia from filtration units. Charcoal is removed by an external vacuum system outlined in CVI Topical Report No. CVI-TR-7301 (February 1975).11.4.2.2Equipment DescriptionThe radioactive solid waste system equipmen t is operated on a batch basis. Individual components are designed to support the system's rated capacity.

The permanently installed system consists of a control panel, spent resin dewatering and hold tanks, and the piping, pumps, and pr ocess equipment modules required for transfer of wastes to a shipping container. The solid waste equi pment requires a minimum manual action and, in conjunction with the building la yout, is designed to minimize o ccupational radiation exposures.

The portable portion of the system is designed to permit filling an d dewatering of the resin slurry in the shipping container. This portion consists of a fill/dewatering head, a dewatering pump, a control panel, and other process e quipment for drying the resin to suitable levels for shipment off site.Shipping containers are approved for the pr ocess for which they are to be used.When slurries are being processed, the fill head is lifted by the overhead bridge crane and lowered onto the shipping container having the integral dewa tering line. Connected to the fill head is a waste fill line, vent line, decont amination lines, and a dewatering li ne that extends from the fill head. These lines are flexible, with flanged fittings to faci litate periodic replacement or maintenance. When the fill head is lowered onto the shipping container, it is aligned with the internal dewatering line a nd secured prior to filling.Boric acid concentrated in the boron evaporator in the boron recovery system is typically recycled. However, the capability exists to transfer waste concentrated from the waste evaporator in the radioactive liquid waste system to the solid waste system.Resins sluiced from demineralizers and ion exchangers are stored in the spent resin hold tank. The resins are then slurried to the shipping container, where they are allowed to settle. Excess water is removed by the spent resin dewatering pump within the portable dewatering unit and transferred to the spent resin dewatering tank.

If solidification were desired, it would be performed by an approved vendor whose Process Control Program was reviewed and SORC approved.

The solidification will be performed with a mobile solidification system provided by the approved vendor.Spent filters, contaminated tool s, and other incompressible contaminated solid wastes can be inserted into a shipping container prior to filli ng. If solidification is required, a vendor will be utilized. Containers a nd shields are handled by overhead bridge crane.

MPS3 UFSAR11.4-5Rev. 30Waste containers are sealed in various ways depending on the system chosen from a number of solid waste system suppliers.The systems are designed to prevent external c ontamination of containers by use of reliable container sealing, appropriate system flushing into the container, and necessary mechanical design or instrumentation interlock signals that prevent overfil ling of containers.The sealed disposable container is transported to an interim on-site storage facility or a disposal site. In general, the disposal container ma y hold spent resin, spent filters, and other incompressible waste in add ition to evaporator bottoms.

Compressible dry solid waste (e.g., contaminated clothing, wipeup toweling) are collected and forwarded for processing.11.4.2.2.1Boron, Waste Evaporator Bottoms

Concentrated liquids from the evaporators may be pumped to th e waste bottoms tank for holdup until they are to be solidified. The tank is insu lated, heat traced, and is recirculated to prevent crystallization, stra tification, settling, and accumulation of undissolved solids.

The amount of waste liquid allowed per container is determined by prior an alysis of the waste.

Radiation levels are also monito red while filling the container to ensure DOT ship ping limits are not exceeded. After the container is filled and solidified, it is sealed and stored. Prior to shipment, it is shielded, as necessary.

Shipments are made in accordance with NRC regulations 10 CFR 20, 10 CFR 50, and 10 CFR 71, and Department of Tran sportation regulations 49 CFR 171 through 178. The shipping containers are stored temporarily in the solidification area (Section 11.4.2.5) or an interim on-site storage facility prior to off-site disposal.11.4.2.2.2Spent Resin Handling Resin in a demineralizer or ion exchanger is considered spent when the decontamination factor falls below a predetermined value, when the demineralizer or ion exchanger surface dose approaches a predetermined limit, or when the resin bed pressu re drop becomes excessive. The demineralizer or ion exchanger is then isolated. Water from the spent resin dewatering tank is used to flush the spent resin into the spent resi n hold tank utilizing the spent resin transfer pump. The flush water passes out of th e spent resin hold tank, through a screened element to prevent any resin carryover, and is recirculated. When the resin is to be packaged for disposal, it is pumped as a resin-water slurry to the shi pping container in the processing area. Excess water is dewatered from the shipping container, via the portable de watering unit, and returned to the spent resin dewatering tank. Radiation levels are monitored to ensure that DOT shipping limits are not exceeded. After dewatering, the cont ainer is sealed, labeled, and te mporarily stored until shipped off site. Prior to shipment the container is shielded as necessary.

MPS3 UFSAR11.4-6Rev. 3011.4.2.2.3Filter Handling Cartridge filter elements are removed from service when the su rface dose on the filter housing reaches a predetermined level or when the element pressure drop becomes excessive. The operation is carried out using remote handling equipment and a filter removal shield, when required. Radiation protection pe rsonnel conduct surveys during filter changeouts as required, using portable radiation detector s and take action as appropriate in accordance with ALARA guidelines.11.4.2.2.4Incompressible Waste Handling Contaminated metallic materials and solid objects are placed in disposab le shipping containers.11.4.2.2.5Waste Compaction OperationContaminated compressible materials are temporar ily stored in suitably labeled containers in different plant locations. These materials are then transported for processing or packaging.11.4.2.3Expected VolumesTable 11.4-1 presents a listing of the expected volumes of spent resins from various sources entering the radioactive solid waste system. (Note that resi n from the condensate polishing demineralizers indicated on Table 11.4-1 is pr ocessed by Millstone Unit 2 - See Figure 10.4-5.) Figure 11.4-2 presents gross activities and weights or volumes for radioact ive solid waste sources.11.4.2.4PackagingBased on the gross activities supplied on Figure 1 1.4-2 and DOT Low Specific Activity Limits, container activity varies between negligible for most compressible or compacted wastes to less than 300 µCi/cc for reactor water purif ication demineralizer resins. The specific radionuclide content of the solid wastes is available, as discussed in Section 11.4.

The filling of containers and the storage of radioactive solid wastes conform with 10 CFR 20 and 10 CFR 50 requirements. Packages meet shipping regulations of 49 CFR 171-178 and 10 CFR 71 as applicable.

Complete solidification and abse nce of free liquid is ensured by the implementation of a process control program and preoperational testing.11.4.2.5Temporary On-site Storage Facilities The processing area is located on the ground floor of the solid waste building. Wastes are packaged to allow for shipment after processing. Therefore, no decay time is assumed in the given estimate of package cont ents and activity levels.Storage capacity and storage time are based on anticipated operational factors.

MPS3 UFSAR11.4-7Rev. 30 In addition to the above, interim on-site storage facilities accept waste from Millstone Units 1, 2 and 3 in accordance with the following criteria.1.The Millstone Radwaste Storage Facility and the On-Site Storage Containers (OSSC's) are used to store liners that contain wastes like dewatered resin and filters. Safety Evaluations performed on the Radwaste Storage Facility and OSSC's conclude that these stru ctures meet the criteria set forth in Section 3.5.1.4.2.In addition to being used for the so rting, processing, loading and shipping of radioactive materials, the MRRF may be used to stor e dry activated waste. The capacity of the MRRF is dependent upon:a.the waste generated from unit/site activities; andb.waste volume reduction techniques employed.3.The Unit 2/Unit 3 Condensate Polishing Facility (CPF) wast e processing area may be used to store mixed waste (radioact ive and hazardous combined) material.11.4.2.6Shipment The shipment of radioactive solid waste conforms with 10 CFR 20, 10 CFR 50, and 10 CFR 61 requirements and 10 CFR 71 and 49 CFR 171 through 178. Solid waste is transferred either directly to a licensed disposal contractor or to a common carrier for delivery to a licensed burial site, or secondary processor as appropriate.Table 11.4-4 summarizes the annual number of shipments and shipped containers for the expected and design cases.11.4.2.7Protection Against Uncontrolled Releases Protection against uncontrolled releases of radioactive material from the radioactive solid waste system is achieved through the use of alar ms, interlocks, and a retaining structure.

The spent resin dewatering tank, evaporator bot toms tank, spent resin transfer pump, and spent resin recycle pump are located in curbed cubicles where any leakage is retained. The walls and floors are suitably finished to facilitate decontaminat ion. The disposal wast e shipping container and the resin fill and dewater head are located on the 24 foot 6 inch elevation of the waste solidification building. In the even t of spillage of radioactive li quid in this area, the liquid is collected by a network of floor drains. The floor is pitched toward the floor drains. The drains are piped to the waste disposal building sumps (FSAR Figure 3.8-74) which are collected by the aerated drains system (FSAR Section 9.3.3) and forwarded to the radioactive liquid waste system (FSAR Section 11.2) for processing.

MPS3 UFSAR11.4-8Rev. 30 The spent resin dewatering tank and the evapor ator bottoms tank are provided with low level alarms at the solid waste panel. For an alar m condition, a visual inspection by the operator is conducted.Health physics personnel, equipped with portable radiation detectors, are present at the fill area to ensure that the radiation levels are within the design levels.

Portable dewatering equipment receives solid waste from the plant and transfers excess water back to the plant.

The single worst operator error or equipment failure would result in the spillage of spent resin and/or evaporator bottoms tank contents. The drainage system in the waste solidification building is designed to handle such an event.

MPS-3 FSARPage 1 of 1Rev. 30NOTE:(1) The condensate polishing demineralizer resin will be processed by Millstone2.TABLE 11.4-1VOLUME OF SPEN T RESIN GENERATED ANNUALLY Demineralizers Number of BedsVolume ft 3/ bedFrequency of Replacement (beds/yr) Fuel pool demineralizer1151Cesium removal ion exchangers2354Boron demineralizer2352 Waste demineralizer2352 Condensate polishing demineralizers (1)81961Mixed bed demineralizers2306Cation bed demineralizers1202 Thermal regeneration demineralizers5742 MPS-3 FSARPage 1 of 2Rev. 30TABLE 11.4-2RADIOACTIVE SOLID WASTE SYS TEM COMPONENT DATASpent Resin Hold Tank Parameters Number1Capacity (gal)2,000Operating pressure50 Design temperature (°F)200Material of constructionSSSpent Resin Dewatering Tank Number1 Capacity (gal)500Operating pressureAtmospheric Design temperature (°F)200Material of constructionSS or FRPBinder Storage Tank (not used)

Number1 Capacity (gal)6,000Operating pressureAtmospheric Design temperature (°F)100Material of constructionSS Spent Resin Recycle Pump Number1 Capacity (gpm)125Operating pressure (psig)75 Design temperature (°F)100Material of constructionSS MPS-3 FSARPage 2 of 2Rev. 30NOTE:(1) Designed in accordance with ASMEVIII, DivisionI.Spent Resin Transfer Pump Number1Capacity (gpm)120Operating pressure (psig)146 Design temperature (°F)100Metal of constructionSS Binder Pump (not used)

Number1 Capacity (gpm)65 Design temperature (°F)150Material of constructionCast ironSpent Resin Transfer Pump Filter Parameters Number1Capacity (gpm)150 Design pressure (psig)150 Design temperature (°F)200Material of constructionSSTABLE 11.4-2RADIOACTIVE SOLID WASTE SYS TEM COMPONENT DATA MPS-3 FSARPage 1 of 1Rev. 30TABLE 11.4-3OMITTED MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Condensate polishing spent resins are designed to be processed by MP2.(2) Miscellaneous compressible solids for Millstone3 are processed at the MRRF.(3) If evaporators are used to process waste instead of resins, then expected shipments will be 56containers/56shipments.TABLE 11.4-4RADIOACTIVE SOLID WASTE ANNUAL SHIPMENTS (1)(2)Type of Waste and PackagingExpectedDesignSolidified / Dewatered Wastes (3)Containers / shipments13 containers/13 shipments33 containers / 33 shipmentsMiscellaneous IncompressibleSolids 128 cu ft crates / shipments4 crates / 1 shipment8 crates / 1 shipment MPS-3 FSAR Rev. 30FIGURE 11.4-1 (SHEETS 1-2) P&ID RADIOACTIVE SOLID WASTE The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS3 UFSAR11.5-1Rev. 3011.5PROCESS, EFFLUENT, AND AIRBORNE RADIATION MONITORING SYSTEMS11.5.1DESIGN BASESThe process, effluent, and airborne radiation monitoring system (RMS) is designed in accordance with NRC General Design Criterion (GDC) 64 (Section 3.1.2.64) and American National Standards Institute (ANSI) N13.1-1969.Normal and potential paths for release of radioactive materials, during both normal operation and anticipated operational occurrences , are continuously monitored to ensure compliance with the requirements of 10CFR20, 10CFR50, and the guidelines of Regulatory Guide 1.21 (Section1.8). Sections 11.5.2 and 11.5.3 descri be the design features provided to ensure agreement with Regulatory Guide 1.21. Potential pathways for rel ease of radioactive materials during accident conditions are continuously monitored to ensure agreement with the guidelines of Regulatory Guide 1.97 (Sections 1.8, 7.5).

Continuous monitoring means that the system ope rates essentially uninterrupted for extended periods during normal plant operation, but does not preclude periods when individual monitors may be out of service for maintenance, repair, calibration, etc.The reactor coolant system (RCS) is monitored continuously for gross activity level. Maintaining the RCS activity within acceptable levels ensures that the activity levels in the normally radioactive auxiliary systems are at acceptable levels. Nonradioactive systems which may become contaminated by leaks from radioactive systems are also monitored continuously. This monitoring ensures that no conditions develop that are potenti ally hazardous to the operating personnel or to the general public.

In the event of an accident re leasing radionuclides, the process, effluent, and airborne RMS and the area RMS (Section 12.3.4), provide informati on on the concentration and dispersion of radioactivity throughout the plant. This enables operating personnel to evaluate the severity and to mitigate the consequences of the accident.The following automatic actions are initiated by the process and effluent RMS to mitigate both the consequences of postulated accidents and excessive releases dur ing normal operations.1.Containment purge (Section 9.4.7) is automa tically terminated in the event of a high radiation alarm.2.Liquid effluent discharges from the tu rbine building drains are diverted to the radioactive liquid waste systems in the event of a high radionuclide concentration alarm (Section 11.2).3.Waste gas holdup system effluent releas e is terminated automatically upon a high radionuclide concentration alarm (Section 11.3.3).

MPS3 UFSAR11.5-2Rev. 304.Liquid waste system discharge to the circulating water discharge tunnel is automatically halted on a high radionuclide concentration alarm (Section11.2).5.Control building ventilation intake is te rminated on high radiation alarm, and the entire control building ventilation system is isolated (Section 9.4.1).6.Discharge from the condensate regenerant demineralizer (removed from service) to the environment was designed to be dive rted to the regenerant evaporator feed tank (removed from service) upon a hi gh radionuclide concentration alarm (Section 11.5.2.3.8).7.Following a high radionuclide concentrati on alarm from the auxiliary condensate monitor, effluent from the auxiliary c ondensate flash tank is diverted from the auxiliary condensate feed tank to the auxiliary building sump.8.A high radionuclide concentration al arm from either hydrogen recombiner ventilation monitor automa tically shuts down ventilation from its cubicle (Section9.4.11).9.Waste neutralization sump effluent discharge to the circulating water discharge tunnel is redirected back to the waste neutralization sump on a high radionuclide concentration alarm (Section 10.4.6.5).10.Steam generator blowdown is automati cally terminated on high radionuclide concentration alarm on the steam generator blowdown monitor.11.5.2SYSTEM DESCRIPTION 11.5.2.1InstrumentationThe process, effluent, and airborne RMS consists of separate and independent monitors, which incorporate the following features:1.Each liquid monitor has one detector. E ach particulate and gas monitor has one gaseous detector and either a fixed pape r disc filter for specific radionuclide laboratory analysis, or a moving filter assembly with an internal particulate detector for gross activity measurement.2.Each monitor is equipped with a dedi cated microprocessor monitoring all its functions. The RMS computer system polls each microprocessor every few seconds. Radiation alarms are displayed and annunciated in the main control room.

Many monitors also have local alarms in addition to the control room.3.For specified monitors, a record of radiological events is printed.

MPS3 UFSAR11.5-3Rev. 30 Alert/alarm setpoints are establis hed to allow observation of change s in radioactivity and/or to ensure that release rates are within guidelines established in 10CFR20 for effluent pathways.

The applicable guidelines are those in Appendix B of 10CFR20.

The dedicated microprocessor sends an alarm message to the RMS computer system in the event of local power loss, loss of sample flow, filter failure, or other conditions specific to a given monitor.Means are provided to purge each fixed volume sample chamber in the system with clean fluid to minimize contamination of the ch ambers. Sample lines and all su rfaces of each sampler exposed to the sample are stainless steel and are run to minimize fixed contamination.

Operability of many detectors can be checked usi ng an installed check source remotely activated from the control room. Tests and calibrations of the radiation monitors are performed at specified intervals. For ease in calibration and maintenance, each detector is located in an easily accessible area and is provided with a local readout device or connects to a plug-in portable indication and control (PIC) module.

All monitors and microprocesso rs are powered by 120 V AC buses. Monitor skids designed in accordance with Regulatory Guide 1.97 and safety-rel ated monitors, such as the control building inlet and ventilation monitors, use the safety-r elated Class 1E buses (Section 8.3.1.1.2). All other monitors use regular 120 V AC buses (Section 8.3.1.1.1).Additionally, all monitors with pumps require 480 V AC power. Safety-related monitors with pumps use safety-related Class 1E buses, and all others use regular 480 V AC buses.N-16 and Fission Product Main Steam Line Monito r alarms are processed by the plant process computer. All other alarms are di splayed visually on the RMS workstations in the control room.

These include equipment malfunctions, alarm/high radiation levels, conductivity, sample flow, and sample temperature. Local and control r oom annunciation is provide d for high radiation alarms, as well as an interfa ce to the main plant annunciato r panel. All alarms can be acknowledged locally or in the control room.Additionally, those variables designated Class 1E are also displayed on the Class 1E control room panels, as required by Regulator y Guide 1.97. A digital display a nd control module is provided for each Class 1E monitor, as well as a dedicated two-pen recorder. In order to record this data at the RMS computer along with that from the non-Class 1E devices, these cabinets are connected to the RMS computer via electronic isolators.

Processes Monitored Table 11.5-1 gives numbers and location of gaseous process and effluent radiation monitors, and Table 11.5-2 gives numbers and locations of liquid process and effluent radiation monitors.

MPS3 UFSAR11.5-4Rev. 30 Sensitivities and Ranges Each detector has sufficient shielding to ensure that the required sensitivity is achieved for the maximum expected background radiati on level at the detector location.

All monitors measure gross concen trations. The output is typically measured in microcuries per cubic centimeter (µCi/cc), with a minimum range of five decades. 11.5.2.2Process and Effluent Monitors 11.5.2.2.1Ventilation Vent Monitors-Normal RangeEach skid has its own microprocessor. The two microprocessors ar e linked by a dedicated interface. When the normal range skid senses the upper level of its range, control is automatically passed to the high range skid. When the radionuclide concentrati on returns to normal, the reverse happens. Both skids use the same isokineti c nozzle. During high range operation, the normal range gas sampler is au tomatically isolated.The sample line and all surfaces of the sampler exposed to the sample are stainless steel. A sample pump provides isokinetic sample fl ow via a flow control valve which modulates proportional to the flow signal from the process flow transmitter. Isokinetic samp le flow is maintained through the normal range skid even after the detectors are isolated duri ng high range operation.

The ventilation vent normal range monitor takes a continuous effl uent sample from an isokinetic nozzle common to this normal and high range monitor and draws the gas sample through a particulate and charcoal filter a nd then to a gas sampling assembly where activity is measured by a beta scintillation detector. Lead shielding is provided to reduce the background radiation to a level that minimizes interference with the detector sensitivity. After leaving the gas sampler, the sample is retu rned to the duct downstrea m of the sample point. A purge system for flushing the sample volume is also provided.

The ventilation vent normal rang e monitor detector output is transmitted via the dedicated microprocessor to the RMS computer system Class 1E panels located in the control room. Here, the activity level is displayed and also recorded on a two-pen strip chart recorder. Activity levels are also digitally displayed locally at the microprocessor location. Alarm conditions, such as high activity or monitor failure, are indicated by audible and visible al arms in the control room and by visible alarm locally.11.5.2.2.2Ventilation Vent Monitor-High Range The vent sample point is common to that of the ventilation ve nt normal range monitor and is located downstream of the last poin t where radioactivity is introduced to the flow stream prior to its release via the ventilation vent stack. Sa mple flow through the high range monitor is maintained constant via a set hand control valve. Isokinetic flow through this monitor depends on continued operation of the normal range monitor pump to maintain isokinetic flow conditions MPS3 UFSAR11.5-5Rev. 30 through the portion of sample line which is common to both monitors. The sample flow is drawn through a fixed 0.3micron filter paper. The activity of the deposited partic ulate is continuously monitored by a Geiger-Mueller detector. An alarm will automatically direct flow to another filter assembly at a level which still allows analysis of the filter in a laboratory GeLi detector. An adequate amount of lead shielding around the detector assembly reduces the background radiation to a level that minimizes interference with the detector sensitivity. Afte r passing through the filter paper, the sample passes through an inline easily removable charcoal filter cartri dge arrangement and then into a fixed and shielded volume wher e it is monitored by mid-range and high range monitors. 11.5.2.2.3Hydrogenated Vent MonitorThe hydrogenated vent monitor continuously monitors the effluent from the gaseous waste system downstream of the charcoal decay beds and prior to their release via the ventilation vent stack. The gas detector, a beta scintillator, is located in the well of an inline gas sampler. Four pi lead shielding is provided in order to reduce the background radiation to a level that minimizes interference with the detector sensitivity.This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of all non-Class 1E process monitors. When high acti vity is present, gas fl ow to the ventilation vent stack is isolated. The sample line and all su rfaces of the sampler exposed to the sample line are stainless steel. Due to the presence of free hydrogen in this effluent, this monitor is purged with nitrogen.11.5.2.2.4Containment Fuel Drop Monitors The gross radionuclide concentration entering the containment purge air vent is monitored by two redundant containment fuel drop monitors. These monitors, each consist of an ion chamber detector measuring dose rates just above the surface of the refueling canal. These are safety-related, Class 1E, monitors. Due to high radiat ion levels inside th e containment, their microprocessors are located in the auxiliary building.

The outputs of these monitors are transmitted, indicated, reco rded, and alarmed in a manner similar to that of the ventilation vent monitor. A high activity indication from either of these monitors automatically isolates containment purge (Section 9.4.7.2) based on the assumption that the high dose rates are due to high airborne activity. 11.5.2.2.5Supplementary Leak Collecti on and Release System Monitor The supplementary leak collection and releas e system (SLCRS) norma l range and high range monitors are identical to the ventilation vent m onitors. Using an isokinetic nozzle, the monitor withdraws a sample from the SLCRS exhaust vent prior to its discharge to the Millstone stack.The SLCRS collects, filters, and releases the le akage from the containment enclosure building and contiguous areas to the atmosphere follow ing a design basis accide nt (DBA) (Section 6.2.3).

During normal operation, this detector monitors the discharge point for th e containment vacuum MPS3 UFSAR11.5-6Rev. 30pumps, the condenser air removal system, the react or plant gaseous vents, and the reactor plant aerated vents. The supplementary leak collecti on monitor is located dow nstream of the SLCRS filters and prior to the exhaust discharge point. The SLCRS monitor functions as a final effluent monitor. This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of the ventilation vent monitors. An alarm from this monitor warns the operator of a potential problem so that he may take appropriate action, which may include the manual transfer of flow to a standby SLCRS filter bank.11.5.2.2.6Condenser Air Ejector MonitorThe condenser air ejector monitor continuously analyzes the gaseous effluents from the condenser air ejector discharge (Sec tion 10.4.2). A gamma scintillation detector is inserted into an in-line air well. The detector is shielded with lead to reduce the background radiation to a level which minimizes interference with the detector sensitivity. The monitor's output is transmitted, indicated, reco rded, and alarmed in a manner similar to that of the hydrogenated vent monitor. Activity re adings are indicative of primary-to-secondary leakage.11.5.2.2.7Control Building Inlet Ventilation Monitors The two control building inlet monitors are located in the upper level of the control building in the inlet plenum. Here, they continuously analyze the ventilation being supplied to the control building by measuring gross activity. Each monitor consists of a beta scintillation detector. These are Class1E, safety-related monitors.

The detector outputs are tr ansmitted, indicated, and alarmed in a manner similar to that of the ventilation vent monitor. A hi gh activity condition initiates a c ontrol building isolation signal which will isolate the control building atmosphere from the ou tside atmosphere (Section 6.4).11.5.2.2.8Hydrogen Recombiner Ventilation MonitorsFollowing a loss-of-coolant accident (LOCA), th e hydrogen recombiners were originally installed to be used to eliminate the free hydrogen in the containment atmosphere. However, the containment atmosphere following a LOCA also contains large amounts of gaseous radionuclides. Should a leak develop in a recombiner, these radionuclides are collected and discharged to the atmosphere via the hydrogen recombiner ventilation system (Section 9.4.11).

Therefore, one safety-related, post-accident, beta scintillation radi ation detector is placed within the exhaust duct of each ventilation system, and its dedicated microprocessor is located in the hydrogen recombiner control room. Upon a high radiation alarm, a signal from the monitor automatically shuts down the respective hydrogen recombiner package sy stem and activates closure of the supply and exhaust duct dampers securing the ventilation.

Recombiner operators can then manually start th e other recombiner unit.

MPS3 UFSAR11.5-7Rev. 30 The output of these monitors is transmitted, indicat ed, and alarmed in a manner similar to that of the ventilation vent monitor.11.5.2.2.9Normal Range Partic ulate and Gas Monitors In addition to the ventilation ve nt and SLCRS monitors, the RMS also contains 11 normal range particulate and gas monitors. They are of a single, basic skid design. The particulate channels employ either a remotely contro lled, variable speed moving paper filter with an integral beta detector, or a fixed paper disc fi lter for specific radionuclide anal ysis. The gas channel is a beta detector. Each monitor includes a removable charcoal filter. These monitors are as follows.1.Reactor Plant Heating and Ventilation System (Section 9.4.3)Eight of these monitors are located in the reactor plant heating and ventilation system upstream of the ventilation vent monitor such that any effluent sampled by any one of them is also sampled by the ventilation vent monitor before it is released. These monitors allow operati ng personnel to locate the source of radionuclide leakage into the air of the re levant spaces. The areas sampled are in auxiliary building, the fuel buildi ng, and the waste disposal building.

These monitors are all designated non-sa fety related, and their outputs are indicated, recorded, and annunciated similarl y to that of the condenser air ejector monitor.2.The Engineered Safety Features Building Heating and Ventilation System (Section9.4.5)

The engineered safety features (ESF) building heating and ventilation system normally discharges directly into the atmosphere. This monitor samples this effluent before release and employs th e fixed paper filter disc design for a radionuclide analysis in compliance w ith Regulatory Guide 1.21. This monitor is also designated non-safety related, and its output is indicated, recorded, and annunciated similarly to that of the condenser air ejector monitor. Following an accident, this system is secured and th e ESF building is ventilated by the SLCRS system, which has its own extended range radiation monitor (Section 11.5.2.2.5). 3.The Control Building Heating and Ventilation System (Section 9.4.1)

Since there are no radionuclide sources within the control building, this system is not a release point. However, this monito r provides an indication of radionuclide concentration within the control buildi ng. It provides greater sensitivity for detecting airborne activity in the cont rol room than the control building inlet monitors.

This monitor is also desi gnated non-safety related, a nd its output is indicated, recorded, and annunciated similarly to that of the condenser air ejector monitor.

MPS3 UFSAR11.5-8Rev. 304.The Containment Atmosphere Monitoring System This monitor continually withdraws a sa mple of the containment atmosphere, analyzes it, and returns it to the contai nment, using dedicated sample lines. These lines are heat traced to prevent conde nsation and slope backwards toward the containment structure. This prevents th e sample lines from becoming sources of radiation. The sample li nes contain valves to is olate this monitor upon a containment isolation signal. The removabl e charcoal filter is not used, Iodine samples are collected using a temporary sampler.11.5.2.2.10Main Steam Relief Line Monitors These four ion chamber monitors, located in th e main steam valve building, measure the gross radionuclide concentration in the main steam lines in the event of a steam generator tube failure.

They are used to meet the intent of Regulatory Guide 1.97 Rev. 2. Lead shielding is provided to reduce the background radiation to a level which does not interfere with the detector sensitivity.The main steam relief line monito rs' output is transmitted via thei r dedicated microprocessors to the RMS workstations located in the control room where the activity level is digitally displayed.

A high activity level is indicated by audible and visible alarms in the main control room.11.5.2.2.11Turbine Driven Auxiliary Feedwater Pump Discharge MonitorAn alternate effluent path for radionuclides entr ained in the main steam is monitored by this detector located in the ESF building. This is an ion chamber gross detector used to meet the intent of Regulatory Guide 1.97, Rev. 2. Lead shielding is provided to reduce th e background radiation to a level which does not interfere with the detector sensitivity.

The turbine driven auxiliary feedwater pump discharge monitor's output is transmitted via its dedicated microprocessor to the RMS workstations located in the control room where the activity is digitally displayed. A high activity level is in dicated by audible and visi ble alarms in the main control room.11.5.2.2.12Main Steam Line Moni tor; N-16 and Fission ProductThe monitor has four unshielded gamma scintillation detectors, one mounted on each main steam line at the containment penetration in the ma in steam valve building. Each detector has two channels, one to measure high energy N-16 activity and one to measure lower energy fission product activity. This monitor inputs all eight ch annels and an equipment failure signal to the plant process computer. The plant process computer records, disp lays, and provides main control board annunciation for this monitor. This monitoring system is Non-QA, designed to support Millstone's Primary-to-Sec ondary leak rate program.

MPS3 UFSAR11.5-9Rev. 3011.5.2.3Liquid Process MonitorsTable11.5-2 lists the locations of liquid proc ess monitors and the streams being monitored.11.5.2.3.1Containment Recirculation Cooler Service Water Outlet MonitorsDuring the recirculation phase of emergency core cooling, the containment recirculation cooler service water outlet monitors c ontinuously measure the radionuclide concentration in the service water effluent from each pair of containment recirculation coolers (one monitor per pair of coolers). These are Class 1E, on-line, gamma sc intillation, gross detect ors, placed atop the discharge pipes just outside the ESF building. As the pipes are underground, only the detectors are so mounted, with their dedi cated microprocessors locat ed in the fuel building.

The containment recirculation cooler service water outlet monitor output is transmitted via the dedicated microprocessor to the RMS computer syst em Class 1E cabinets located in the control room. Here, the concentration is digitally displayed and also recorded on a strip chart recorder. The activity level is also digitally displayed locally at the micr oprocessor location. A high activity level is indicated by audible an d visible alarms locally and in the main control room. These monitors warn of a leak into the service wate r system, within the containment recirculation coolers. 11.5.2.3.2Liquid Waste Monitor The liquid waste monitor continuously analyzes the liquid waste effluent discharge pipe (Section 11.2) downstream of the last possible poi nt of radioactive liqui d addition. The detector assembly consists of a gamma scintillation detector inserted into the well of a four-pi, lead shielded liquid sampler. All surfaces in contact wi th the liquid sample are of stainless steel. Lead shielding is provided in order to reduce the background radiation to a level which does not interfere with the detector sensitivity. The detector output is transmitte d, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor. A high activity situation initiates closure of a discharge valve, thereby preventing the discharge of effluent to the environment in excess of dose limits. The high activity alarm and isolat ion function are based on a time averaged activity to verify consistently high levels.11.5.2.3.3Steam Generator Blowdown Sample Monitor The steam generator blowdown sample monitor analyzes the steam generator blowdown effluent (Section 10.4.8) for radioactivity which would be indicative of primary-to-secondary leakage.

Samples from each of the four steam generator bottoms are mixed in a common header. This common sample is continuously monitored by a ga mma scintillation detect or inserted into the well of a four-pi, lead-shielded liquid sampler. All surfaces in c ontact with the liquid sample are of stainless steel. Lead shielding is provided in order to reduce the background radiation to a level which does not interfere with the detector sensitivity.

MPS3 UFSAR11.5-10Rev. 30This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor. If signifi cant activity is det ected, the monitor will automatically isolate the steam generator blowdown. 11.5.2.3.4Auxiliary Condensate Monitor The auxiliary condensate monitor continuously analyzes samples drawn from the discharge of the auxiliary condensate flash tank (Section10.4.10). The detector asse mbly consists of a gamma scintillation detector inserted into the well of a four pi, lead-shielded liquid sampler. All surfaces in contact with the liquid sample are of stainless steel. Lead sh ielding is provided in order to reduce the background radiation to a level which does not interfere with the detector sensitivity.

An inline conductivity element provides sample specific conductivity me asurement, indication, and alarm. A sample cooler reduc es the sample temperature to 140

°F or less, and inlet and outlet solenoid valves isolate the det ector assembly and stop the sa mple pump automatically on high sample temperatures.The detector output is transmitte d, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor. During nor mal operation, activities significantly above background are indicative of a leak into the auxiliary steam system (Section 10.4.10) from one of the systems containing radioactive fluids which exchange heat with the auxiliary steam system. A high radiation alarm will automatically di vert flow to auxiliary building sumps. 11.5.2.3.5Turbine Building Floor Drains MonitorThe turbine building floor drains monitor analyzes a sample from the turbine building floor drains discharge line (Section 11.2), downstream of any possible fluid addition to the discharge line. The detector assembly consists of a gamma scintillation detector inserted into the well of a four-pi, lead shielded liquid sampler. All surfaces in cont act with the liquid sample are of stainless steel. An adequate amount of lead shielding is provided in order to reduce the background radiation to a level which does not interfere with the detector sensitivity.This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor. A high act ivity level initiates valve action to divert the effluent to the liquid waste system, thereby preventing the discharge of effluent to the environment in excess of dose limits.11.5.2.3.6Reactor Plant Component Cooling Water System MonitorThe reactor plant component cooling water system monitor continuously analyzes the component cooling water for radioactivity (Section 9.2.2.1).

A sample is continuously withdrawn from the reactor plant component cooling water system downstream of the reactor plant component cooling pumps discharge. The sample is monitored by a gamma scintillation detector inserted into the well of a four-pi, lead shielded liquid sampler. All surfaces in contact with the liquid sample are of stainless steel. Lead shielding is provided in order to reduce the background radiation to a level which does not interfere with the detector sensitivity.

MPS3 UFSAR11.5-11Rev. 30This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor. During normal operation, activities significantly above background are indicative of a leak into the reac tor plant component cooling water system from one of the systems containing ra dioactive fluids which exchange heat with the reactor plant component cooling water subsystem.11.5.2.3.7Deleted by FSARCR 05-MP3-015 11.5.2.3.8Regenerant Evaporator M onitor (Removed from Service)

Located in the condensate demineralizer liqui d waste system (removed from service) the regenerant evaporator monitor was designed to measure gross radionuclide concentration in the distillate discharged from the regenerant evaporator. Following a high concentration alarm, the effluent would be automatically isolated.

An evaluation has been performed which has determined that the LWC is not needed. This evaluation is documented by change to the Radiological Effluent Monitoring Offsite Dose Calculation Manual (ref. REMODCM CR# 95-7).11.5.2.3.9Waste Neutralization Sump Monitor Located in the condensate polishing facility, the waste neutralizat ion sump monitor measures the radionuclide concentration in the effluent from a sump located below potentially contaminated condensate polishing equipment. The sample is monitored by a gamma scintillation detector inserted into the well of a four-pi, lead shielded liquid sampler.

All surfaces in contact with the liquid sample are of stainless steel. Lead shielding is provided in order to reduce the background radiation to a leve l which does not interfere with the detector sensitivity.This monitor's output is transmit ted, indicated, recorded, and alarmed in a manner similar to that of the condenser air ejector monitor.11.5.2.4Inservice Inspection, Calibration, and Maintenance Most channels of the process and effluent RMS are checked routinely with an installed check source or manually at the monitor. Periodic tests are also used to verify th e operability of alarms and automatic actions.

Calibration of monitors is condu cted periodically in accordance wi th the technical specifications.Process and effluent monitors were initially isotopically calibrated at the vendor and concurrently cross calibrated to a secondary National Institute of Science and Tec hnology (NIST) traceable source in a fixed field geometry.

Field calibration is accomplishe d using these secondary sources.

MPS3 UFSAR11.5-12Rev. 30Repair of failed electronic components is accomplished in the units' instrument and control shop. Sufficient spare parts, staffing, and training are being provided to support this effort.

Radiochemical analysis takes place in the chemis try laboratory counting room. The facility is equipped with a gamma spectromet er traceable to NIST for th e purpose of isotopic analysis.11.5.2.5Sampling Section 9.3.2 discusses the various process and effl uent samples taken periodically for chemical and radiochemical analysis.Table 11.5-3 lists liquid process and effluent samples to be take n periodically and monitored for radioactivity. Those not covered in Section 9.3.2 are included in the indi vidual system designs.

Sampling of these fluid systems is via local sa mpling connections, e.g., the fuel pool cooling and purification system (Section 9.1.3).

Prior to collecting a sample, liquid sample lines are purged of stagnant water and undissolved solids for a sufficient time to ensure that a representative sample is obtained.

Sample taps suitable for connection to a sampling chamber are provided at all off-line process monitors to obtain a sample for la boratory gamma spectrum analysis.

Gas samples are collected in a sample vessel wi th valves on each end. After adequate purging of the sample vessel, the gas sample is collected by closing valves at both ends of the sample vessel.

The isokinetic nozzles used for obtaining uniform samples (Section 11.5.2.2) are designed according to the guide lines of ANSI N13.1.

A sampling room (Section 9.3.2) is provided fo r remote sampling. Recirculation loops are separated from the sample taps by shielding. Sample sink drains are collec ted and sent to various systems, depending on the nature of the sample being taken, for reclai ming or processing as necessary. Sample sinks are provide d with ventilation exhaust hoods.11.

5.3REFERENCES

FOR SECTION 11.511.5-1ANSI N13.1 - 1969. Guide to Sampling Air borne Radioactive Materials in Nuclear Facilities, American National Standards Institute (1974).11.5-2EPRI TR104788-R2 PWR Primary-To-Sec ondary Leak Guidelines - Revision 2.

MPS-3 FSAR Page 1 of 3 Rev. 30TABLE 11.5-1GASEOUS MONITORSMonitorMark Number (1)Number of ChannelsMediumLocationMeasurement RangeVentilation Vent (2)3HVR*RE10AirAux Bldg 66'-6"Normal Range1Normal OPSHigh Range1Post AccidentHydrogenated Vent (2)3GWS-RE481GasAux Bldg 43'-6"Normal OPS Fuel Drop (3)3RMS*RE41 2 (4)AirContainment 51'-4"Normal OPS 3RMS*RE42 Supplementary Leak Collection (2)3HVR*RE19AirAux Bldg 66'-6"Normal Range1Normal OPSHigh Range1Post Accident Condenser Air Ejector (5)3ARC-RE211AirTurbine Bldg 38'-6"Normal OPS -- Will Alarm With Tube Failure Control Building InletVentilation (5)3HVC*RE16A 2 (4)AirControl Bldg 64'-6"Post Accident 3HVC*RE16B Containment Atmosphere (2)3CMS*RE22AirAux Bldg 66'-6" Particulate1Normal OPSGas1Normal OPS Auxiliary Building (2)3HVR-RE11AirAux Bldg 66'-6" &

43'-6" MPS-3 FSAR Page 2 of 3 Rev. 30Particulate3HVR-RE126Normal OPS Gas3HVR-RE136Normal OPS 3HVR-RE14 3HVR-RE15 3HVR-RE16 Fuel Building (2)3HVR-RE17AirAux Bldg 66'-6"Particulate1Normal OPS Gas1Normal OPSWaste Disposal (2)3HVR-RE18AirAux Bldg 66'-6"Particulate1Normal OPS Gas1Normal OPS Control Building (2)3HVC-RE91AirControl Bldg 64'-6"Particulate1Normal OPS Gas1Normal OPS ESF Building (2)3HVQ-RE49AirESF Bldg 36'-6"ParticulateNote 6Normal OPS Gas1Normal OPS Hydrogen (5) Recombiner Cubicle Vent 3HVZ*RE09A 2 (4)AirHR Bldg 37'-6"Post Accident 3HVZ*RE09BTABLE 11.5-1GASEOUS MONITORSMonitorMark Number (1)Number of ChannelsMediumLocationMeasurement Range MPS-3 FSAR Page 3 of 3 Rev. 30NOTES:(1) A and B used to indicate redundant monito rs powered from sepa rate safety trains.(2) Offline monitors.(3) The fuel drop monitors are configured as high range area m onitors, having a minimum sensitivity of 0.1R/hr and a range of 6decades.(4) Redundant monitors.(5) Inline monitors.(6) Offline laboratory radionuclide analysis.Main Steam Relief Line (5)3MSS-RE75 to 784SteamMSVB 70'-6"Post AccidentTurbine Driven Auxiliary Feedwater Pump Discharge (5)3MSS-RE791SteamESF 36'-6"Post AccidentMain Steam Line3MSS-RE80AN-163MSS-RE80B4SteamMSVB 70'6"Normal OPSFission Product3MSS-RE80C4SteamMSVB 70'6"Normal OPS 3MSS-RE80DTABLE 11.5-1GASEOUS MONITORSMonitorMark Number (1)Number of ChannelsMediumLocationMeasurement Range MPS-3 FSARPage 1 of 1Rev. 30(1) A and B used to indicate redundant monito rs powered from sepa rate safety trains.(2) Inline monitors.(3) Offline monitors.TABLE 11.5-2 LIQUID PROCESS MONITORSMonitor Mark Number (1)Number of Channels Medium LocationMeasurement Range Containment Recirculat ion Cooler Service Water Outlet (2) 3SWP*RE60A 2WaterYard Post Accident 3SWP*RE60BLiquid Waste (3)3LWS-RE70 1Water Aux Bldg 4'-6" Normal OPSSteam Generator Blowdown Sample (3) 3SSR-RE08 1Water Aux Bldg 43'-6" Normal OPS Auxiliary Condensate (3) 3CNA-RE47 1Water Aux Bldg 4'-6" Normal OPSTurbine Building Floor Drains (3)3DAS-RE50 1WaterTurbine 14'-6" Normal OPSReactor Plant Component Cooling Water Subsystem (3)3CCP-RE31 1Water Aux Bldg 43'-6" Normal OPS Regenerant Evaporator (3) (Removed from Service)3LWC-RE65 1WaterWarehouse 5 4'-6" Normal OPSWaste Neutralization Sump (3)3CND-RE07 1Water Condensate Polishing Facility 14'-6" Normal OPS MPS-3 FSARPage 1 of 1Rev. 30TABLE 11.5-3RADIOLOGICAL SAMPLES TA KEN AT REACTOR PLANT SAMPLE SINKSample LocationsNumberReactor Coolant System (Chapter 5) Loop No. 1 (Hot Leg)1Loop No. 3 (Hot Leg)1 Pressurizer Vapor Space1 Residual Heat Removal System (Chapter 5.4.7) RHR Heat Exchanger Outlet2Reactor Plant Component Cooling Water System (Section 9.2.2.1)Pump Discharge2Primary Grade Water System (Section 9.2.8) Primary Grade Water Tanks2Chemical and Volume Control System (Section 9.3.4) Letdown Heat Exchanger Outlet2 Reactor Coolant Filter Inlet1Volume Control Tank1Boron Thermal Regeneration Outlet1 Boron Recovery System (Section 9.3.5)Boron Recovery Tanks2Boron Test Tanks2Steam Generator Blowdown System (Section 10.4.8) Blowdown Sample4Radioactive Liquid Waste System (Section 11.2) Waste Test Tank2Radioactive Gaseous Waste System (Section 11.3) Degasifier Condenser Process Effluent1 MPS-3 FSAR Rev. 30APPENDIX 11APART I -

SUMMARY

OF ANNUAL RADIATION DOSES (HISTORICAL)PART II - DOSE CALCULATION MODELS AND ASSUMPTIONS (HISTORICAL)PART III - COST-BENEFIT ANAlLYSIS (HISTORICAL)

MPS-3 FSAR11A.1-1Rev. 30PART I -

SUMMARY

OF ANNUAL RADIATION DOSES (HISTORICAL)

The dose evaluation presented in this Appendix wa s developed in support of the original license and portions were updated during the MPS-3 rest art. These dose estimates are considered historical and not s ubject to future upda ting. This information is retained to avoi d loss of original design basis.As stated in Section 11.0, the Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) provides guidance requirement s for system operation, dose calculations, and monitoring requirements to ensure MPS-3 compliance with effluent limits. Actual measured concentrations of radioa ctivity released and real time di lution and dispersion estimates are utilized to verify compliance with effluent limits. Therefore, MPS-3 operation within the requirements of the REMODCM ensures compliance within effluent limits, rather than operations within the nominal assumptions utilized in the dose evaluation presented in this Chapter.

The calculated annual radiation doses to th e maximum individual fr om liquid and gaseous pathways are presented in Tables11A-1 through 11A-8 and 11A-13 through 11A-15. Table11A-16 demonstrates that the calculate d annual radiation doses are below the design objectives of 10CFR50, AppendixI.The maximum calculated organ dos e per reactor for an individual from gaseous releases (particulates and radioiodines) is 4.4mRem/yr to an infant's thyroid. This represents a hypothetical infant living at the residence 2.4km north-northeast of the site consuming milk from a goat at the same location.

The calculated external exposure to the whole body and skin from immersion in noble gases is 3.8E-02 and 6.9E-02 mRem/yr, respectively. These represent the maximum values which occur at the site boundary in the direction of the maximum overland/Q, 650meters east-northeast of Millstone3. The maximum calculated beta a nd gamma air doses from noble gas releases are 6.6E-02 and 8.6E-02 mrad/yr, respectively. These were also calculated 650meters east-northeast of Millstone3.

For liquid releases, the maximum individual wa s assumed to consume aquatic foods whose principal habitat is the edge of the initial mixing zone (EIMZ). This location was also conservatively used in calculating doses fr om boating. Doses from swimming and shoreline recreation were calculated at the nearest resident's beach, located 1.1km from the point of discharge. The maximum calculated whole body dos e for an individual from liquid pathways is 2.4E-02mRem/yr in the adult age group. The maximum calculated organ dose for an individual from liquid pathways is 4.4E-01mRem/yr to an adult's GI-UI and 1.8E-01 mRem/yr to a child's thyroid. These doses were primary due to consumption of aquatic foods.

The calculated annual gaseous and liquid doses from the population residing within an 80km radius of the site are presented in Table11A-17. For liquid effluents, the calculated population dose commitment within 80km for whole body and thyroid are 1.7E+00 and 1.6E+01 man-Rem/yr, respectively.

MPS-3 FSAR11A.1-2Rev. 30For the gaseous effluents, the calculated population doses within 80km from noble gas effluents and radioiodines and particulates are 4.8man-Rem/yr whole body and 7.8man-Rem/yr thyroid.

Population doses were calculated for a projected population of 3.3 million people residing within 80km of the site in the year2010.

The calculated annual gaseous and liquid doses to the continguous U.S. population are also presented in Table11A-17. For liquid effluents, the calculated dose to the contiguous U.S.

population is 1.7E

+00 man-Rem whole body and 1.6E

+01 man-Rem thyroid. For gaseous effluents, the calculated dose to the contiguous U.S. population is 2.2E

+01 man-Rem whole body and 2.5E+01 man-Rem thyroid.

MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.7E+00 = 1.7 x 10 0TABLE 11A.1-1ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS (Residence 0.81 km ENE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.7E+00 (1)2.0E+001.7E+001.7E+001.7E+001.7E+001.7E+001.7E+00Inhalation1.2E-010.08.6E-031.2E-012.5E-011.1E-011.4E-011.1E-01Fresh vegetation9.7E-020.08.3E-021.2E-011.5E+006.2E-023.5E-026.0E-02Stored vegetation5.7E-010.04.7E-016.7E-012.2E-013.3E-012.2E-013.2E-01Total dose2.5E+002.0E+002.3E+002.6E+003.7E+002.2E+002.1E+002.2E+00 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.7E+00 = 1.7 x 10 0TABLE 11A.1-2ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS (Residence 0.81 km ENE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.7E+00 (1)2.0E+001.7E+001.7E+001.7E+001.7E+001.7E+001.7E+00Inhalation1.2E-010.01.1E-021.2E-012.9E-011.2E-011.6E-011.2E-01Fresh vegetation5.6E-020.07.1E-029.7E-021.2E+004.9E-022.7E-023.8E-02Stored vegetation5.9E-01 0.04.7E-011.0E+002.8E-014.7E-013.0E-013.7E-01 Total dose2.5E+002.0E+002.5E+002.9E+003.5E+002.3E+002.2E+002.2E+00 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.7E+00 = 1.7 x 10 0TABLE 11A.1-3ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS (Residence 0.81 km ENE) Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.7E+00 (1)2.0E+001.7E+001.7E+001.7E+001.7E+001.7E+001.7E+00Inhalation1.0E-010.01.4E-021.1E-013.1E-011.0E-011.4E-011.0E-01Fresh vegetation5.0E-020.01.2E-011.2E-011.8E+005.9E-023.1E-023.1E-02Stored vegetation6.9E-010.0 1.6E+001.7E+004.8E-017.6E-014.7E-014.3E-01Total dose2.5E+002.0E+003.4E+003.6E+004.3E+002.6E+002.3E+002.3E+00 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.7E+00 = 1.7 x 10 0TABLE 11A.1-4ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS (Residence 0.81 km ENE) Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.7E+00 (1)2.0E+001.7E+001.7E+001.7E+001.7E+001.7E+001.7E+00Inhalation5.9E-020.09.1E-036.2E-022.5E-016.0E-028.2E-025.8E-02Total dose1.8E+002.0E+001.7E+001.8E+001.9E+001.8E+001.8E+001.8E+00 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.8E-01 = 1.8 x 10

-1TABLE 11A.1-5ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.8E-01 (1)2.2E-011.8E-011.8E-011.8E-011.8E-011.8E-011.8E-01Inhalation3.6E-020.03.4E-033.7E-026.2E-023.6E-024.4E-023.6E-02Fresh vegetation2.0E-020.01.4E-022.2E-021.9E-011.4E-021.1E-021.4E-02Stored vegetation1.2E-010.08.1E-021.3E-016.4E-028.2E-026.6E-028.1E-02Goat milk1.8E-010.01.2E-012.4E-015.8E-011.1E-016.6E-025.0E-02Total dose5.4E-012.2E-014.0E-016.1E-011.1E+014.2E-013.7E-013.6E-01 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.8E-01 = 1.8 x 10

-1TABLE 11A.1-6ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.8E-01 (1)2.2E-011.8E-011.8E-011.8E-011.8E-011.8E-011.8E-01Inhalation3.7E-020.04.6E-033.8E-026.9E-023.7E-024.8E-023.6E-02Fresh vegetation1.2E-020.01.2E-021.8E-021.5E-011.1E-027.7E-039.3E-03Stored vegetation1.3E-010.01.3E-011.9E-018.2E-021.1E-018.7E-029.8E-02Goat milk1.9E-010.02.2E-013.9E-019.0E-011.7E-011.0E-01 6.5E-02Total dose5.5E-012.2E-015.5E-018.2E-011.4E+005.1E-014.2E-013.9E-01 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.8E-01 = 1.8 x 10

-1TABLE 11A.1-7ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.8E-01 (1)2.2E-011.8E-011.8E-011.8E-011.8E-011.8E-011.8E-01Inhalation3.2E-020.06.1E-033.4E-027.0E-023.3E-024.2E-023.2E-02Fresh vegetation1.2E-020.02.1E-022.2E-022.2E-011.3E-029.3E-039.3E-03Stored vegetation1.7E-010.02.9E-013.2E-011.4E-011.8E-011.4E-011.4E-01Goat milk2.0E-010.05.2E-016.5E-011.8E+002.8E-011.6E-011.0E-01Total dose5.9E-012.2E-011.0E+001.2E+002.4E+006.9E-015.3E-014.6E-01 MPS-3 FSAR Page 1 of 1 Rev. 30NOTE:(1) 1.8E-01 = 1.8 x 10

-1TABLE 11A.1-8ANNUAL DOSES TO MAXIMUM INDIVI DUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)

Annual Dose (mRem/yr)PathwayTotal BodySkinBoneLiverThyroidKidneyLungGI-tract Contaminated ground 1.8E-01 (1)2.2E-011.8E-011.8E-011.8E-011.8E-011.8E-011.8E-01Inhalation1.9E-020.04.1E-032.0E-025.3E-021.9E-022.5E-021.9E-02Goat milk2.4E-010.08.5E-011.2E+004.2E+004.4E-012.6E-011.5E-01 Total dose4.4E-012.2E-011.0E+001.4E+004.4E+006.4E-014.6E-013.5E-01 MPS-3 FSARPage 1 of 1Rev. 30TABLE 11A.1-9OMITTED MPS-3 FSARPage 1 of 1Rev. 30TABLE 11A.1-10OMITTED MPS-3 FSARPage 1 of 1Rev. 30TABLE 11A.1-11OMITTED MPS-3 FSARPage 1 of 1Rev. 30TABLE 11A.1-12OMITTED MPS-3 FSAR Page 1 of 1 Rev. 30TABLE 11A.1-13ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM LIQUID EFFLUENTS Annual Dose (mRem/yr)PathwaySkinBoneLiverTotal BodyThyroidKidneyLungGI-LLIFish0.00E+007.29E-031.40E-021.04E-027.72E-025.15E-031.90E-031.02E-02Invertebrate0.00E+001.97E-021.34E-021.05E-021.02E-011.01E-016.33E-044.26E-01Shoreline3.18E-032.73E-032.73E-032.73E-032.73E-032.73E-032.73E-032.73E-03Swimming0.00E+002.89E-052.89E-052.89E-052.89E-052.89E-052.89E-052.89E-05Boating0.00E+001.80E-051.80E-051.80E-051.80E-051.80E-051.80E-051.80E-05Total3.18E-032.98E-023.02E-022.37E-021.82E-011.09E-015.31E-034.39E-01 MPS-3 FSAR Page 1 of 1 Rev. 30TABLE 11A.1-14ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM LIQUID EFFLUENTS Annual Dose (mRem/yr)PathwaySkinBoneLiverTotal BodyThyroidKidneyLungGI-LLIFish0.00E+007.63E-031.43E-026.28E-037.24E-025.16E-032.11E-037.29E-03Invertebrate0.00E+002.09E-021.38E-021.02E-029.63E-021.05E-017.12E-043.01E-01 Shoreline3.18E-032.73E-032.73E-032.73E-032.73E-032.73E-032.73E-032.73E-03Swimming0.00E+002.89E-052.89E-052.89E-052.89E-052.89E-052.89E-052.89E-05Boating0.00E+00 1.80E-051.80E-051.80E-051.80E-051.80E-051.80E-051.80E-05 Total3.18E-033.13E-023.09E-021.93E-021.71E-011.12E-015.60E-033.11E-01 MPS-3 FSAR Page 1 of 1 Rev. 30TABLE 11A.1-15ANNUAL DOSES TO MAXI MUM INDIVIDUAL IN THE CHILD GROUP FROM LIQUID EFFLUENTS Annual Dose (mRem/yr)PathwaySkinBoneLiverTotal BodyThyroidKidneyLungGI-LLIFish0.00E+009.43E-031.23E-022.95E-037.57E-024.34E-031.67E-032.74E-03Invertebrate0.00E+002.73E-021.27E-021.14E-021.06E-019.31E-025.94E-049.47E-02 Shoreline1.78E-031.53E-031.53E-031.53E-031.53E-031.53E-031.53E-031.53E-03Swimming0.00E+001.62E-051.62E-051.62E-051.62E-051.62E-051.62E-051.62E-05Boating0.00E-001.00E-051.00E-051.00E-051.00E-051.00E-051.00E-051.00E-05 Total1.78E-033.83E-022.66E-021.59E-021.84E-019.90E-023.82E-039.90E-02 MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Per reactor.(2) 8.6E-02 = 8.6 x 10

-2.(3) Site boundary 650 meters ENE of Millstone 3.(4)Infant thyroid dose at resi dence with goat, 2.4 km NNE.(5) Adult GI-UI dose is calculated to be the highest organ dose.TABLE 11A.1-16COMPARISON OF MA XIMUM CALCULATED DOSES FROM MILLSTONE 3 NUCLEAR PLANT WITH APPENDIX I DESIGN OBJECTIVES Criterion Appendix I Design Objective (1)Calculated DoseGaseous EffluentsGamma air dose10 mrad/yr 8.6E-02 (2) mrad/yr (3)Beta air dose20 mrad/yr 6.6E-02 mrad/yr (3)Noble gas - total body5 mRem/yr 3.8E-02 mRem/yr (3)Noble gas - skin15 mRem/yr 6.9E-02 mRem/yr (3)Iodines and particulates, any organ 15 mRem/yr 4.4E+00 mRem/yr (4)Liquid Effluents:Total body3 mRem/yr2.4E-02 mRem/yr Any organ10 mRem/yr 4.4E-01 mRem/yr (5)

MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1)Natural Radiation Exposure in the United States, U.S. Environmental Protection Agency, ORP-SID-72-1 (June1972), using the average state background dose (100mRem/yr), and year 2010 projected population of 3.3million.(2) 3.3E+05 = 3.3 x 10 5(3) Carbon-14 and tritium have been added to this category.TABLE 11A.1-17CALCULATED POPULATION DOSE80-KM POPULATION DOSEAnnual Dose Per Reactor UnitTotal Body (man-Rem)Thyroid (man-Rem)

Natural radiation background (1)3.3E+05 (2)3.3E+05Liquid effluents1.7E+001.6E+01Noble gas effluents1.3E-011.3E-01 Radioiodines a nd particulates (3)4.7E+007.7E+00CONTIGUOUS U.S. POPULATION DOSEAnnual Dose Per Reactor UnitTotal Body (man-Rem)Thyroid (man-Rem)Liquid effluents1.7E+001.6E+0 Noble gas effluents1.5E-04.0E-0 Radioiodines a nd particulates (3)2.2E+02.5E+0 MPS-3 FSAR11A.2-1Rev. 30PART II-DOSE CALCULATION MODELS AND ASSUMPTIONS (HISTORICAL)

Doses to Humans Calculation of dose rates to the maximum indi vidual and to the population residing within an 80-km radius of the site are based on the met hodology and equations of U.

S. Nuclear Regulatory Guide 1.109 "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for The Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I, Revision1, October 1977." Stone & Webster Engineering Corporation Computer Programs IND1109E, POP1109E, NG1109E, DUCKMANE, and NEPA were used in the design case analysis of the maximum individual and population doses. These computer codes have been verified by comparing the Stone & Webster program outputs to results obtained from th e U.S. NRC Computer Codes GASPAR and LADTAP and to hand calculations, where appropriate. The U.S. NRC Computer Code, LADTAPII was used for the expected case analysis. The U.S. NRC Computer Code RABFIN was used to analyze the elevated release of noble gases. NRC defa ult values have been used in lieu of site specific data, where site data was unavailable. The site specific data that was used for this analysis is listed in Tables 11A.2-1, 11A.2-2, and 11A.2-3.

The following sections present the equations used in the analysis for each pathway considered.

Doses from Liquid Pathways The generalized equation for cal culating radiation doses to humans via liquid pathways is:

R aipj = (C ip)(U ap)(D aipj) (11A.2-1) where: R aipj = the annual dose to organ j of an individual of age group a fr om nuclide i via pathway p, in mRem/yr C ip = the concentration of nuclide i in the me dia 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, 1/yr, or kg/yr (as appropriate)

D aipj = the dose factor, specific age group a, radionuclide i, pathway p, and organ j, in mRem/pCi ingested or mRem per hr/pCi per sq m from ex posure to deposited activity in sediment or on the ground1.Aquatic Foods (11A.2-2)

Rapj 1100 U ap M p F----------------

-Q i B ip D aipji t p-()exp i=

MPS-3 FSAR11A.2-2Rev. 30 where: B ip =the equilibrium bioaccumulation factor for nuclide i in pathway p, expressed as the ratio of the concen tration in biota (in pCi/kg) to the radionuclide concentration in water (in pCi/l), in l/kg M p =the mixing ratio (reciprocal of the d ilution factor) at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food), dimensionlessF = the flow rate of the liquid effluent in cu ft/s

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 cons tant of nuclide i, in hr t p =the average transit time required for nuclides to reach the point of exposure. For internal dose, t is the total time elapsed between release of the nuclides and ingestion of food or water, in hours 1,100 = the factor to convert from (Ci/yr)/(cu ft/s) to pCi/l All the other symbols are as previously defined.2.Doses from Shoreline Deposits Foods (11A.2-3) where: W = the shoreline width factor that describes the geometry of the exposure, dimensionless T i = the radiological half-lif e of nuclide i, in days t b = the period of time for which sediment or soil is exposed to the contaminated water, in hours110,000 = the factor to convert from (Ci/yr)/(cu ft/s) to pCi/l and to account for the proportionality constant used in the sediment radioactivity model All other symbols are as previously defined.3.Doses from Swimming and Boating Rapj110000 U ap M p W F--------------

---

Q i T i Daipji t p-()exp[]1i t b-()exp-[]i,=

MPS-3 FSAR11A.2-3Rev. 30 The doses from swimming and boating were calculated using the methodology described in WASH 1258 (Atomic Energy Commission 1973).

The equation for calculation of the external dose to skin and the total body dose from swimming (water immer sion) or boating (water surface) is: (11A.2-4) where: K p = geometry correction factor equal to 1 fo r swimming and 2 for boa ting, dimensionless (no credit is taken for the shielding provided by the boat)

All other symbols are as previously defined.

Doses from Air Pathways 1.Gamma and Beta Doses from Noble Gas Discharged to the Atmospherea.Annual Gamma and Beta Air Doses from Noble Gas Releas es (ground level) (11A.2-5) where: D(r,), D(r,) = the annual gamma and beta air doses at the distance r in the sector at angle from the discharg e point in mrad/year Q i = the release rate of the radionuclide i, in Ci/year

[/Q](r,) = the annual average gaseous dispersion fa ctor at the distance r in sector in sec/cu m DFi, DFi = the gamma and beta air dose factors for a uniform semi-infinite cloud of radionuclide i, in mrad-cu m/pCi-yr 3.17 x 10 4 = the number of pCi per Ci divided by the numbe r of seconds per yearb.Annual Total Body Dose from N oble Gas Releases (ground level) (11A.2-6)

Rapj 1100 U ap M p FK p----------------

-Q i D aipji t p-()exp i=Dr, ()orDr, ()3.1710 4xQ iQ[]r, ()DFiorDFi ()i=D Tr, ()S Fi r, ()DFB i i=

MPS-3 FSAR11A.2-4Rev. 30 where: D T (r,) = the total body dose due to immersion in a semi-infinite cloud at the distance r in sector , in mRem/year S F = the attention factor that accounts for dose reduction due to shielding provided by residential structures, dimensionlessi (r,) = the annual average ground-level concentr ation of radionuclide i at the distance r in sector , in pCi/cu m DFB i = the total body dose factor for a semi-inf inite cloud of the radionuclide i which includes the attenuation of 5 g/sq cm of tissue, in mRem-cu m/pCi-yrc.Annual Skin Dose from Noble Gas Releases (ground level) (11A.2-7) 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 factor for a semi-infinite cloud of radionuclide i, which includes the attenuation by the outer "dead" laye r of the skin, in mRem-cu m/pCi-yr1.11 = the average ratio of tissue to air energy absorption coefficients All other parameters are as previously defined.d.Annual Gamma Air Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High (11A.2-8)

D Sr, ()1.11S Fi r, ()DFii r, ()DFS i i+i=Dr, ()260 r()--------------

-1 u n-------f nsµa E k ()E kIHusz E k,,,, ()Q D ni A ki iksn=

MPS-3 FSAR11A.2-5Rev. 30 where: A ki is the photon yield for gamma-ray photons in energy group k from the decay of radionuclide i, in photons/disintegration D(r,) is the annual gamma air dose at a distance r (meters) in the sector at an angle in mrad/yr Ek is the energy of the kth photon energy group, in MeV/photon f ns is the joint frequency of occurrence of stabi lity class s and wind spee d class n for sector , dimensionless I(H,u,s,z ,E k) is the dimensionless num erical integration cons tant accounting for the distribution of radioactivity according to meteorological conditions of wind speed (u) and atmospheric stability (s) which in part determine the effective stack height (H) and the vertical plume standard deviation (z). See Regulatory Guide 1.109 for deviation Q D ni is the release rate of radionuclide i, corr ected for decay during tr ansit to the distance r under wind speed u n , in Ci/yr u n is the mean wind speed of wind speed class n, in m/sec is the sector width over which atmosphe ric conditions are average, in radians

µa (E k) is the air energy absorption coefficient for the kth photon energy group, in m

-1 260 is the conversion factor to obtain D (r,), in mrad/yr, and has the units of mrad-radians-m 3 - disintegration/sec-MeV-Cie.Annual Total Body Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High (11A.2-9)

D T r, ()1.11S F Dk r,()µT a E k ()t d-[]exp k=

MPS-3 FSAR11A.2-6Rev. 30 where: D T(r,) is the annual total body dose at the distance r in sector , in mrem/yr Dk (r,) is the annual gamma air dose associated with the kth photon energy group at the distance r in sector , in mrad/yr S F is the attenuator factor that accounts for the dose reduction due to shielding provided by residential structures, dimensionless t d is the product of tissue density and depth us ed to determine a total body dose, in g/cm 2µT a (E k) is the tissue energy absorption coefficient, in cm 2/g; and1.11 is the average ratio of tissue to air energy absorption coefficientsf.Annual Skin Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High (11A.2-10) where: DFS i is the beta skin dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuator by the outer "dead" layer of the skin, in mrem-m 3/pCi-yr D s(r,) is the annual skin dose at the distance r in sector , in mrem/yr All other parameters are as de fined in preceding paragraphs.2.Doses from Radioiodines and Other Radi onuclides (not including Noble Gases) Released to the Atmospherea.Annual Organ Dose from External Irr adiation from Radionuc lides Deposited onto the Ground Surface (11A.2-11) where: D G j (r,) = the annual dose to the 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, dimensionless D s r, ()1.11S F Dr, ()3.1710 4xQ iQ[]D r, ()DFS i i+=D G j r, ()8760S F C G i r, ()DFG ij i=

MPS-3 FSAR11A.2-7Rev. 30 C G i (r,) = the ground plane concentration of radionuclide i at distance r in sector , in pCi/sq m DFGij = the open field ground plane dose conversion factor for organ j from radionuclide i, in mRem-sq m/pCi-hr8,760 = the number of hours in a yearb.Annual Organ Dose from Inhalation of Radionuclides in Air (11A.2-12) where: D A ja (r,) = the annual dose to organ j of an i ndividual 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 cu m/yri (r,) = the annual average concentration of radionuclide i in air at location (r,), in pCi/cu m DFAija = the inhalation dose factor for radionuclide i, organ j, and age group a, in mRem/pCic.Annual Organ Dose from Ingestion of Atmospherically Released Radionuclides in Food (11A.2-13) where: C v i (r,), C m i (r,) = the concentrations of radionuclide i in produce C L i (r,), C F i (r,) = (non-leafy vegetables, fruits, and grains), milk, leafy vegetables, and meat, respectively, 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 D A ja r, ()R ai r, ()DFAija i=D D ja r, ()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, ()+++[]i=

MPS-3 FSAR11A.2-8Rev. 30 DFIija = the ingestion dose factor for radionuclide i, organ j, and age group a in mRem/pCi f g , f 1 = 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/yr General Expression for Population Doses The general expression for calculating the annual population-integrated dose is: (11A.2-14) where: D P j = the annual population-integrated dose to organ j (total body or thyroid), in man-Rems or thyroid man-Rems P d = the population associated with subregion d D jda = the annual dose to organ j (total body or thyroid) of an average individual of age group a in subregion d, in mRem/yr f da = the fraction of the population in s ubregion d that is in age group a0.001 = the conversion factor from mRem to Rem The above equation used in conjunction with the preceding equations and average usage factors for each age group was used to calculate the population doses.

For further refinements on the preceding equati ons used to calculate the doses to man, see Regulatory Guide 1.109, Revision 1.

Doses to Biota Other Than Man Calculation of dose rates to biota other than man was performed by means of the computer programs ARRRG and CRITER (Soldat et al., 1974), developed at the Pacific Northwest Laboratory of Battelle Memorial Institute under contract to the Atomic Energy Commission (NRC). Bioaccumulation factors used in ARRR G and CRITER have been updated to correspond to the latest published values in Regulator y Guide 1.109, Revision 0 (plants) and Regulatory Guide 1.109, Revision 1 (all others). Site specific data used in this analysis are presented in Table 11A.2-4.The following sections provide a summary of the dose models used in the analysis for each pathway considered.

D P j0.001P d D jda f da ad=

MPS-3 FSAR11A.2-9Rev. 30Internal Doses to Aquatic OrganismsAquatic organisms were considered to receive an internal dose rate from uptake and concentration of radiochemicals in the water and from exposure through the food chain. Dose rates to primary organisms were calculated directly from radioisotopic concentrations in discharge water and from bioaccumulation factors. The dose rate through the food chain was estimated for secondary organisms such as muskrats and raccoons feeding on primary organisms whose radionuclide content was estimated in the first calculation.

Equations used by the program CRITER fo r these calculations are as follows: (11A.2-15) where: (DR)i = dose rate for radionuclide i (mrad/yr)

E i = effective absorbed energy in organ of interest (MeV/dis) b i = specific body burden of nuclide i (pCi/kg)A = conversion factor = and b i = C iw B i where: Ciw = concentration of nuclide i in water (pCi/l)

B i = bioaccumulation factor for nuclide i (pCi/kg per pCi/l)

The concentration in water C iw is calculated from: (11A.2-16) where: Q i = release rate of nuclide i (Ci/yr)

R i = reconcentration factor to es timate recycling of effluent M p = mixing ratio at point of exposure (1/dilution factor)

DR ()i AE i b i=0.0187diskg-mrad-pCiyr-MeV----------------------------------------

C iw 1119 Q i R i M p F-------------------i t p-()exp=

MPS-3 FSAR11A.2-10Rev. 30F = flow rate of the liquid effluent (cu ft/s)i = radiological decay cons tant of nuclide i (hr

-1)t p = transit time for nuclides to reach point of exposure (hr)1,119 = constant to convert Ci/yr per cu ft/s to pCi/lThe total body dose rate to secondary organisms wa s calculated as follow s (Soldat et al., 1974):

DR'i = 0.365 b i P' D'i (11A.2-17) where: DR'i = total body dose rate to secondary or ganisms due to nuc lide i (mrad/yr) 0.365 = kg-day/g-yr D i (man) = total body dose conversion factor for man for radionuclide in mRem/pCi e i (man) = effective absorbed energy for ma n for radionuclide i (meV/disintegration) e'i = effective absorbed energy for s econdary organism for radionuclide i (meV/disintegration) m i = mass of secondary organism (grams)

P i = consumption rate of primary organism s by the secondary organism (grams/day) 70,000 = total body mass of adult (grams)

The actual equation used by CRITER was of the form: (11A.2-18) where: DR' = total body dose rate to secondary organisms (mrad/yr) n = 136, number of isotopes 2.86x10 7 = (0.365) (1119) (70,000)

All other parameters are as previously defined.

D'i70000 D i man ()e i man ()---------------------

-e'i m'-----,=DR'2.8610 7xM p P'Fm'------------

Q i R i B i e'ii t p-()D i e i[]man ()exp i1=n=

MPS-3 FSAR11A.2-11Rev. 30 Exposure to Shoreline Deposits (11A.2-19) where: (DR)pr = dose rate to organ r (total body or skin) from pathway P (mrad/yr)

W f = shore width factor = 0.5 (ocean shoreline)

T i = radiological half-life of isotope i (days) n = 136, number of isotopes111,900 = constant to convert (Ci/y r)/(cu ft/sec) to pCi/literDose for Swimming and Water Surface Exposure (11A.2-20) where: K p = hemispherical correcti on constant = 1 for swimming and 2 for boating n = 136, number of isotopesDose from Immersion in Gaseous Effluents These doses were calculated in the same manner as doses to human s with appropriate changes in use factors as shown in Table 11A.2-1.References for Appendix 11A Part II - Do se Calculation Mode ls and Assumptions Atomic Energy Commission 1973. Final Environmental Statement Concerning Proposed Rule Making Action; Numerical Guides for Design Obje ctives and Limiting Conditions for Operation to Meet the Criterion ("as low as practicable") for Radioactive Material in Light Water Cooled Nuclear Power Reactor Effluents. Washington, D.C.Soldat, S.K.; Robinson, N.M.; and Baker, D.A. 1974. Models and Computer Codes for Evaluating Environmental Radiation Doses. Battelle P acific Northwest Laborat ories BNWL-1754, Richland, Wash.DR ()pr111900 U p M p W f F--------------

---

Q i R i T ii t p-()1i t ()D ipr exp-()exp i1=n,=DR'pr 1119 U p M p FK p----------------

Q i R i Dipri t p-()exp i1=n=

MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Location used to calculate doses to maximum offsite individual from ingestion of aquatic foods and boating.(2) Location used to calculate doses to maximum offsite individual from shoreline recreation and swimming.(3) Location used to calculate doses to popula tion from ingestion of aquatic foods, boating, swimming, and shoreline recreation. The travel time and dilution factor for the edge of the initial mixing zone radius is conservatively ap plied to the entire 80 km radius. It is also assumed that the entire 80 km radius popul ation participates in swimming and boating.(4) 3.3E+06 = 3.3 x 10 6.TABLE 11A.2-1DILUTION FACTORS, TRAVEL TIMES FROM THE SITE, AND POPULATION SERVED Location of AnalysisApproximate Distance from Site (km)

Dilution Factor Transit Time to Point of Analysis (hr)

Population Served Edge of initial mixing zone (1)030.0 (assumed)-Closest accessible shoreline (2)1.17.20.0 (assumed)-

Edge of initial mixing zone (3)030.0 (assumed) 3.3E+06 (4)

MPS-3 FSARPage 1 of 2Rev. 30TABLE 11A.2-2PARA METERS AND ASSUMPTIONS US ED IN EQUATIONS FOR ESTIMATING DOSES TO HUMANS All parameters and assumptions used are recommended values to be us ed, in lieu of site specific data, from Regulatory Guide 1.109, Revision 1.The following are site specific parameters or parameters for which there is no recommended value: F = normal circulation flow rate (f or 3-unit operation) = 4,160 cu ft/sec T p = transit time = see Table 11A.2-1 Note: T p used in calculations wa s increased, where appropriate, by the distribution or holdup time recommended by Regula tory Guide 1.109, Revision 1.

P = fractional equilibrium ratio of C 14 = 1 (continuous release); = 0.0073 (intermittent containment release); = 0.062 (intermitt ent steam generato r blowdown release)

Q i = annual release rate of radionuclide i, Ci/yr (Tables 11.2-7 and 11.3-1) f p = fraction of year animals gr aze on pasture = 0.67 (8 months) f s = fraction of daily feed which is pastur e grass when animal is grazing = 1 (100%)

H = absolute humidity of atmosphere at location of analysis 9.86 g/cu m U ap = usage factor (hr/year of exposure):

Maximum Individua lAdultTeenChildSwimming10010056 Boating525229 80 km RadiusAdultTeenChildSwimming3.51012Boating292916.5 MPS-3 FSAR TABLE 11A.2-2 CONTINUEDPage 2 of 2Rev. 30 V p = Total commercial U.S. fi sh harvest = 1.1E+09 kg/yr (1)V p ' = Total commercial U.S. sh ellfish harvest = 5.2E+08 kg/yr V dp = 80 km commercial fi sh harvest = 9.7E+06 kg/yr V dp ' = 80 km sports fish harvest = 9.7E+06 kg/yr V dp = 80 km invertebrate harvest = 2.8E+06 kg/yr V dp = 80 km milk production = 5.6E+08 1/yr V dp ' = 80 km meat production = 1.7E+07 kg/yr V = 80-km vegetation production = 1.8E+08 kg/yrNOTE:(1) 1.1E+09 = 1.1 x 10 9

MPS-3 FSARPage 1 of 2Rev. 30TABLE 11A.2-3METEOROGICAL DATA Radiological Release Points:1.Millstone stack (continuous release)2.Ventilation vent (intermittent release)3.Ventilation vent (continuous release)4.Turbine building vent (continuous release)5.Steam generator blowdown vent (intermittent release)6.Condensate polishing buildi ng vent (continuous release)

Meteorological Parameters - c/Q = Sec/m3; D/Q = m-2LocationResidentAnnual Average Growing/Grazing Season810 m ENEMaximum resident/Q 1 (1)4.03E-08 (2)4.59E-08 D/Q 12.16E-091.70E-09/Q 25.24E-065.70E-06 D/Q 26.22E-085.40E-08/Q 33.50E-063.98E-06 D/Q 34.15E-083.72E-08/Q 41.22E-051.54E-05 D/Q 47.57E-087.86E-08/Q 51.95E-052.45E-05 D/Q 51.21E-071.28E-07/Q 61.19E-051.51E-05 D/Q 67.39E-087.86E-082,400 m NNEMaximum goat/Q 16.60E-088.00E-08 D/Q 17.04E-107.58E-10/Q 23.08E-063.56E-06 D/Q 21.48E-081.55E-08 MPS-3 FSARPage 2 of 2Rev. 30NOTES:(1) Numerical superscripts correspond to release point.(2) 4.03E-08 = 4.03 x 10

-8.LocationResidentAnnual Average Growing / Grazing Season/Q 37.96E-091.04E-06 D/Q 33.94E-094.87E-09/Q 49.71E-061.28E-06 D/Q 43.81E-094.62E-09/Q 52.41E-063.02E-06 D/Q 59.84E-091.12E-08/Q 69.20E-071.23E-06 D/Q 63.76E-094.59E-09650 m ENEMaximum site boundary/Q 11.27E-08-D/Q 12.34E-09-/Q 28.08E-06-D/Q 29.70E-08-/Q 34.81E-06-D/Q 36.29E-08-/Q 41.90E-05-D/Q 41.22E-07-/D 53.06E-05-D/Q 51.96E-07-/Q 61.86E-05-D/Q 61.19E MPS-3 FSARPage 1 of 1Rev. 30NOTE:(1) Edge of mixing zone and nearest shoreline.TABLE 11A.2-4PARAMETERS AND ASSUMPTIONS USED IN ESTIMATING DOSES TO BIOTA ParameterValues Assigned Primary OrganismsSecondary Organisms(Fish, Crustaceans, Mollusks, Algae)MuskratHeronDuckRaccoon R i (recirculation factor00000F (flow rate, cfs) 4,1604,1604,1604,1604,160 M p (mixing ratio)

(1)0.3330.3330.3330.3330.333 W f (shore width factor)-0.50.50.50.5K (water immersion)-1---(water surface)--22-Effective radius (cm)2611514M mass (kg)-14.6112 P food consumption (gpd)aquatic plants-100-100-fish--600--invertebrates----200 U p usage (hr/yr)shoreline-2,9222,9224,3832,191water immersion-2,922---water surface--2,9224,383-t holdup time (hr)00000Residence time (mo)1212121212 MPS-3 FSAR11A.3-1Rev. 30PART III-COST-BENEFIT ANALYSIS (HISTORICAL)

This appendix presents the results of cost-benef it analyses performed in accordance with Section II. D of 10CFR50, Appendix I.Augments to the liquid and gaseous effluent sy stems and respective potential reductions to the annual population exposure are taken from the U.S. NRC Regulatory Guide 1.110, Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (Regulatory Guide 1.110, 1976). The beneficial savings of eac h augment were calculated by multiplying the calculated dose reduction by $1,000 per man-Rem or $1,000 per man-thyroid-Rem. The cost of borrowed money was conservatively assumed to be 10 percent. The equations and site specific data required for the dose calculations ar e presented in Part I of this appendix.Augments to the Liquid Effluent Treatment System Table 11A.3-1 presents the calculated base case annual total body dose (man-Rem) and thyroid dose (man-thyroid-Rem) associated with the operation of the plant liquid radwaste system for the population expected to live within an 80-km radi us of the plant for the year 2010. Assuming that each augment is capable of reducing the popula tion doses to zero (an ex tremely conservative assumption), the maximum benefit to be derived from any augment would be $1,700 for reducing man-Rem exposures to zero and $16,000 for re ducing man-thyroid-Rem exposure to zero.

In an analysis of the annualized procurement, installation, operation, and maintenance costs, the least expensive liquid radwaste augment was f ound to be $19,000 per year for a plant located in the northeastern United States. Since the benef it from this augment would be less than the corresponding total annualized cost, the cost-benef it ratio is greater than 1. The operation of additional equipment for the purpose of reducing the annual population dos e would not be cost effective. Therefore, the most cost-beneficial system has been included in the current plant design.Augments to the Gaseous Effluent Treatment System Table 11A.3-2 presents the calculated base case annual total body man-Rem and thyroid man-Rem associated with the operation of the ga seous radwaste system for the 80-km radius population.Assuming that each augment is capable of redu cing the population doses to zero, the maximum benefit to be derived from any augment would be $4,800 for reducing man-Rem exposures to zero and $7,800 for reducing man-thyroi d-Rem exposures to zero.

In an analysis of the annualized procurement, installation, operation, and maintenance costs, the least expensive gaseous radwaste augment was foun d to be $8,700 per year for a plant located in the northeastern United States. Since the benef it from this augment would be less than the corresponding total annualized cost, the cost-benef it ratio is greater than 1. The operation of additional equipment for the purpose of reducing the annual population dos e would not be cost MPS-3 FSAR11A.3-2Rev. 30effective. Therefore, the most cost-beneficial system has been included in the current plant design.Reference for Appendix 11A Part III - Cost-Benefit Analysis Regulatory Guide 1.110, 1976. Cost-Benefit Analysis for Radwaste Systems for Light Water-Cooled Nuclear Power Reactor. March 1976.

MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Total annual dose from all existi ng pathways for Millstone 3 operation.(2) 1.2E-00 = 1.2 x 10 0TABLE 11A.3-1BASE CASE ANNUAL PO PULATION DOSES DUE TO LIQUID EFFLUENTS (1) PathwayTotal Body Dose (man-Rem)Thyroid Dose (man-thyroid-Rem)

Ingestion of fish 1.2E-00 (2)9.2E+00Ingestion of other seafood 2.5E-016.5E+00Shoreline recreation 2.4E-012.4E-01 Swimming 3.1E-033.1E-03Boating 6.2E-036.2E-03Total 1.7E+001.6E+01 MPS-3 FSARPage 1 of 1Rev. 30NOTES:(1) Total annual dose from all existi ng pathways for Millstone 3 operation(2) 1.3E-01 = 1.3 x 10

-1TABLE 11A.3-2BASE CASE ANNUAL POPU LATION DOSES DUE TO GASEOUS EFFLUENTS (1)PathwayTotal Body Dose (man-Rem)Thyroid Dose (man-thyroid-Rem)

Submersion 1.3E-01 (2)1.3E-01Inhalation8.8E-011.9E+00Standing on contaminated ground3.3E+003.3E+00

Ingestion of fruits, grains, and vegetation 8.6E-022.6E-01Ingestion of cow milk 3.9E-012.2E+00Ingestion of meat 2.1E-022.7E-02 Total 4.8E+007.8E+00 MPS-3 FSAR FIGURE 11.2-3 EXPECTED RADIOACTIVE LIQUID WASTE SOURCE AND DISCHARGE PATHS D.F.=1.0 O.F.=100 1, OTl£RS D.F.=2 Ca, Rb w (;)a:<t::I: o V).....o:l: w(J)V>426,'411 CPO C148 x 10 3*/HRl FROJA rorrAltN:Kl" 6UIilliNG SU\4P*-WASTE WASTE TO PRIMARYDEMINDEMIN I---GRADE WATER.0 GPO FILTERFROU AUXILIARY OOILDlHC StJ.fP*-200 GPO O.F.:: 1.0..*..D.F.=LO D.F.=10 D.r.=1.0 REACTOR PlANl SANPlE SINKS HlGHLEVEl WASTE WASTE...WASTE t.t.WASTETEST r--PULtPDIJlIteUl-OOOt£AAl-..£VAPOOATOR TANK IZER 35 GPO DRAIN TANK lZf.R tILlER LADORA my YfASl£S*26,000 GAL 35 GPN 24,000 GAL 150 GP10I 400 GPO---.UISCElUt£OOS HIGH-LEVEL WASTE TO PRIMARY 660 GPO D.F..=10 I.Br D.F.=10 LOr I.8r GRADE WATER O.F.=2C60 Rb D.F.=2Cs.RbD.f.=10 AU.0Tl£RS D.F.=].0 OJ".=10 AU.0ll0S D.F.::;'J.o D.f.=1 AU.0Tl£RS OF.::;1.0 O.F.=LO O.F.=1.0 REACTOR COOlANT BlEED 10 OBCICAl C£SlUU6000N RECOVERY.&VOLlH:OEGASlFIER H REMlVAl r-.RECOV£RY I-tP1..NP r---TESTDOOJ£RAl-H FR-TER-...CONTID.100 EVAPORATOO TAm..1440 CPO SYSlEU EXCHANGER TANK IZER 150.000 GAL 25GPN 50 CPM 12.000 GAl REACTOR PlANT GAS8lUS ORAlNS 300 Q>O D.F.=LO REGtNERANT CtOOCAl.S 3400 GPO FILTER D.f.::;1.0 D.F.::;1.0 Ul$C£lLAtEOOS lOfH.EV£l WASTE-40 GPO...LOI LEVEL H EfflLENT WASTEPLI.lP FILlERDRAIN TAN(llRllNE PlANT tEMACE.10 SUN?4 000GL 50 CPf,£7200 CPO'A TO CONDENSER STEAM GfNERATffi 8l0WDQWN<<PEN CYCLE)11 10 8 4 7 5 9 3 2 1 6 NOTES*-VIA REACTOR PLANT AERATED ORA.lNS SYSTEM.**-NORMALLY DISCHARGED OIRECTL Y TO ENVIRONMENT.***-NOT REQUIRED TO MEET 10 CFR 50 APPENDIX I GUIDELINES.

D.F.-DECONAMINA TION FACTORS ARE CONSISTENT WITH NUREG-0017.

SECTION 2.2.21.GPO-ESTIMATED OR EXPECTED FLOWS ARE PROVIDED FOR ILLUSTRA nON OF PROCESSED WASTES.FLOW ARE NOT NECESSARILY CONSISTENT WITH ACTUAL OPERATING FLOW DATA.FIGURE 11.2-3 EXPECTED RADIOACTIVE LIQUID WASTE\SOURCE AND DISCHARGE PATHS.CAD FILE: 1123.dqn/1l23.clt MAY 1998 May 1998 Rev.20.3 MPS-3 FSARFIGURE 11.3-2 VENTILATION SYSTEM COMPOSITE DRAWING NORMAL OPERATION.Rev. 16 MPS-3 FSAR FIGURE 11.4-2 (HISTORICAL)

RADIOACTIVE SOLID WASTE SYSTEM EXPECTED OUANTITIES Historical, not subject to future updating.Has been retained topreserveoriginal design basis.OFf SITE SINDER PROMOTE/<Co CATI\LYS T on ACCEPTA8LE ALTERNATE

'2 i*04Ci/l'H$PENT'U6Ei 01Cl/FTJ RESINS (II-"'" jHIPPING r-----+2.GOE+01 le:;FT-'/YR CONTAINER 1.30E-01Ci/FPi 02CI/YR SHIELDING MISCELLAN EOU S CASKS AS 1.69Ci/FT-'INCOMPRESSIBLEREQUIRED WASTE (2)900 fT3/YR 500 FP/YR...SPENT 6.16Ci/YR RESINS (1)(51...2.04E-02cI/FT3 400 FT::l/YR BORON EVAPORATOR 1.27cl/YRBOTTOMS (3)8.47-C3Ci/rT FT3/YR p-I OPERATION 6.MAINTENANCE NEGLIGIBLE ACTIVITY OTHER SERVICE RADIOACTIVE LIQUID WASTE SYSTEM BORON RECOVERY SYSTEM REACTOR PLANT SERVICE SPENT FILTERS'-CONDENSATE CO:.lDE:-':Sf.TE i POLISHING POL..lSHING SPENT RESINS FACILITY-j 0 FT 3 IYR (41 I MILLSTON£UNIT 2 PROCE5SlEG FACILITY t.MAINTENANCE 4'MISCELLANEOUS COMPHESSIBLE WASTES (2)NEGL!GIBLE ACTIVITY 1 MILLSTO"lE RAOWASF RED...CTl01'FACILITY NOTES: 1.Ci FT J{YR VALUES BASED UPON VOLUME OF RAW SPENT RESIN.2.Ci F-r{YR VALUES BASED UPON VOLUME OF PACKAGED WASTE.3.Ci FT J IYR VALUES BASED UPON VOLUME OF RAW BOTTOMS.4.NO CONDENSATE POLISHING SPENT RESINS ARE EXPECTED TO BE GENERA TED FOR NORMAL EXPECTED RADIATION LEVELS.5.ALTERNATE METHOD WOULD PRODUCE APPROXIMA TEL Y 3.000 FT J IYR OF RAW EVAPORATOR BOTTOMS, THIS METHOD WOULD NOT BE THE NORMAL OR PREFERRED METHOD OF DISPOSAL.Rev.21.3 MPS-3 FSAR FIGURE 11.4-3 (HISTORICAL)

RADIOACTIVE SOLID WASTE SYSTEM DESIGN OUANTITIES Historical, not subject to future updating.Has been retained topreserveoriginal design basis.OFf SITE BINDER PROMOTER C CATALYST OR ACCEPTABLE ALTERNATE PROCESS 69E-05 Ci/¥R l SPENT.92 E.02Ci/FT RESINS (I)24E-04 Ci/YR 1600 SHIPPING r---+..CONTAINER.'-01 Ci/FTs t.'("04 C'/YR SHIELOING.63 E-Ol CI/FT3 MISCELLANEOUS CASKS AS INCOMPRESSIBLE REQUIRED WASTES (2)p..t600 FTs/YR 00 FT'/YR-.SSE-OlCI/YR SPENT RESINS (1)(:\)...76 E-02 CI/FT r 830 frs/YR BORON EVAPORATOR.SSE-OICi/VR BOTTOMS (3)...25E-02Ci/FT 600 FrS/YR....10 I.9.2 OPERATION£.MAINTENANCE NEGLIGIBLE ACTIVITY OTHER SERVICE REACTOR PLANT SERVICE RADIOACTIVE LIQUID WASTE SYSTEM 2 SPENT FILTERS Z BORON 5 RECOVERY SYSTEM 9 CONDENSATE POLISHING rACILlTY 1.56£..02 C'/YR..CONDENSATE POLISHING SPENT RESINS 10.000 fTllYR MILLSTONE UNIT 2-.-PROCE'SSING FACILITY OPERATION e tAAINTENANCE NEGLIGIBLE ACTIVITY MISCELLANEOUS COMPRESSI6LE WASTES(Z)6000 fT.5/YR 1-----..;-

..MILLSTONE RAOWASTE....REDUCTION fACILITY NOTES: 1.Ci FT 3 IVR VALUES BASED UPON VOLUME OF RAW SPENT RESIN.2.Ci FT 3/YRVALUES BASED UPON VOLUME OF PACKAGED WASTE.3.ci FT 3/YR VALUES BASED UPON VOLUME OF RAW BonOMS.4.WHEN RADIATION LEVELS IN THE CONDENSATE REQUIRE THE PROCESSING OF RESIN 5.ALTERNATE METHOD WOULD PRODUCE APPROXIMATELY 6.000 FT 3 IYR OF RAW EVAPORATOR BOTTOMS.THIS METHOD WOULD NOT BE THE NORMAL OR PREFERRED METHOD OF DISPOSAl.Rev.21.3