ML17212A081
| ML17212A081 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/2017 |
| From: | Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17212A038 | List:
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| References | |
| 17-208 | |
| Download: ML17212A081 (219) | |
Text
Millstone Power Station Unit 3 Safety Analysis Report Chapter 11
Table of Contents tion Title Page BACKGROUND ...................................................................................... 11.0-1 SOURCE TERMS .................................................................................... 11.1-1
.1 RADIONUCLIDE INVENTORY IN THE CORE .................................. 11.1-1
.2 RADIONUCLIDE INVENTORY IN FUEL ELEMENT GAP ............... 11.1-2
.3 PRIMARY COOLANT EQUILIBRIUM ACTIVITIES.......................... 11.1-3
.3.1 Fission Product Activities ......................................................................... 11.1-3
.3.2 Tritium Activity ........................................................................................ 11.1-4 1.3.3 Corrosion Products ................................................................................... 11.1-7
.3.4 Nitrogen-16 Activity................................................................................. 11.1-7
.4 RADIOACTIVITY IN THE SECONDARY SIDE.................................. 11.1-7
.5 REFERENCES FOR SECTION 11.1....................................................... 11.1-8 LIQUID WASTE MANAGEMENT SYSTEMS ..................................... 11.2-1
.1 DESIGN BASES ...................................................................................... 11.2-1
.2 SYSTEM DESCRIPTION........................................................................ 11.2-3
.2.1 Radioactive Liquid Waste System (LWS)................................................ 11.2-3
.2.2 Condensate Demineralizer Liquid Waste System (LWC) ........................ 11.2-5
.2.3 Other Systems Discharging Radioactive Liquid Waste............................ 11.2-6
.3 RADIOACTIVE RELEASES .................................................................. 11.2-7
.3.1 Radioactive Liquid Waste System Leak or Failure (Atmospheric Release).............................................................................. 11.2-7
.3.2 Liquid Containing Tank Failure ............................................................... 11.2-8
.4 REFERENCE FOR SECTION 11.2 ......................................................... 11.2-8 GASEOUS WASTE MANAGEMENT SYSTEMS ................................ 11.3-1
.1 DESIGN BASES ...................................................................................... 11.3-1
.1.1 Design Objective....................................................................................... 11.3-1
.1.2 Design Criteria .......................................................................................... 11.3-2
.1.3 Cost Benefit Evaluation ............................................................................ 11.3-3
.1.4 Equipment Design Criteria ....................................................................... 11.3-3
.1.5 Building Ventilation Systems ................................................................... 11.3-4
.2 SYSTEM DESCRIPTIONS ..................................................................... 11.3-4
.2.1 Radioactivity Inputs and Release Points................................................... 11.3-4
.2.2 Degasifier Subsystem of Radioactive Gaseous Waste System................. 11.3-5
.2.3 Process Gas Subsystem of Radioactive Gaseous Waste System....................................................................................................... 11.3-6 11-i Rev. 30
tion Title Page
.2.4 Process Vent Portion of Radioactive Gaseous Waste System .................. 11.3-6
.2.5 Steam and Power Conversion System ...................................................... 11.3-7
.2.6 System Instrumentation Requirements ..................................................... 11.3-7
.2.6.1 Radioactive Gaseous Waste System ......................................................... 11.3-7
.2.6.2 Ventilation Systems .................................................................................. 11.3-8
.2.7 Seismic Design Provisions of the Radioactive Gaseous Waste System............................................................................................ 11.3-8
.2.8 Quality Control ......................................................................................... 11.3-8
.2.9 Welding..................................................................................................... 11.3-9
.2.10 Materials ................................................................................................... 11.3-9
.2.11 Construction of Process Systems .............................................................. 11.3-9
.2.12 System Integrity Testing ........................................................................... 11.3-9
.3 RADIOACTIVE RELEASES ................................................................ 11.3-10
.3.1 Radioactive Gaseous Waste System Failure........................................... 11.3-11
.4 REFERENCE FOR SECTION 11.3 ....................................................... 11.3-12 SOLID WASTE MANAGEMENT .......................................................... 11.4-1
.1 DESIGN BASES ...................................................................................... 11.4-1 4.2 SYSTEM DESCRIPTION........................................................................ 11.4-2
.2.1 System Inputs............................................................................................ 11.4-3
.2.1.1 Spent Resins.............................................................................................. 11.4-3
.2.1.2 Waste Evaporator Bottoms ....................................................................... 11.4-3
.2.1.3 Regenerant Chemical Evaporator Bottoms (Removed From Service) ..................................................................................................... 11.4-3
.2.1.4 Boron Evaporator Bottoms ....................................................................... 11.4-3
.2.1.5 Miscellaneous Radioactive Solid Wastes ................................................. 11.4-3
.2.2 Equipment Description ............................................................................. 11.4-4
.2.2.1 Boron, Waste Evaporator Bottoms ........................................................... 11.4-5
.2.2.2 Spent Resin Handling ............................................................................... 11.4-5
.2.2.3 Filter Handling .......................................................................................... 11.4-6
.2.2.4 Incompressible Waste Handling ............................................................... 11.4-6
.2.2.5 Waste Compaction Operation ................................................................... 11.4-6
.2.3 Expected Volumes .................................................................................... 11.4-6
.2.4 Packaging.................................................................................................. 11.4-6
.2.5 Temporary On-site Storage Facilities ....................................................... 11.4-6
.2.6 Shipment ................................................................................................... 11.4-7
.2.7 Protection Against Uncontrolled Releases ............................................... 11.4-7 11-ii Rev. 30
tion Title Page PROCESS, EFFLUENT, AND AIRBORNE RADIATION MONITORING SYSTEMS...................................................................... 11.5-1
.1 DESIGN BASES ...................................................................................... 11.5-1 5.2 SYSTEM DESCRIPTION........................................................................ 11.5-2
.2.1 Instrumentation ......................................................................................... 11.5-2
.2.2 Process and Effluent Monitors.................................................................. 11.5-4
.2.2.1 Ventilation Vent MonitorsNormal Range............................................. 11.5-4
.2.2.2 Ventilation Vent Monitor-High Range ..................................................... 11.5-4
.2.2.3 Hydrogenated Vent Monitor ..................................................................... 11.5-5
.2.2.4 Containment Fuel Drop Monitors............................................................. 11.5-5
.2.2.5 Supplementary Leak Collection and Release System Monitor ................ 11.5-5
.2.2.6 Condenser Air Ejector Monitor ................................................................ 11.5-6
.2.2.7 Control Building Inlet Ventilation Monitors ............................................ 11.5-6
.2.2.8 Hydrogen Recombiner Ventilation Monitors ........................................... 11.5-6
.2.2.9 Normal Range Particulate and Gas Monitors ........................................... 11.5-7
.2.2.10 Main Steam Relief Line Monitors ............................................................ 11.5-8
.2.2.11 Turbine Driven Auxiliary Feedwater Pump Discharge Monitor .............. 11.5-8
.2.2.12 Main Steam Line Monitor; N-16 and Fission Product ............................. 11.5-8 5.2.3 Liquid Process Monitors ........................................................................... 11.5-9
.2.3.1 Containment Recirculation Cooler Service Water Outlet Monitors.................................................................................................... 11.5-9
.2.3.2 Liquid Waste Monitor............................................................................... 11.5-9
.2.3.3 Steam Generator Blowdown Sample Monitor.......................................... 11.5-9
.2.3.4 Auxiliary Condensate Monitor ............................................................... 11.5-10
.2.3.5 Turbine Building Floor Drains Monitor ................................................. 11.5-10
.2.3.6 Reactor Plant Component Cooling Water System Monitor.................... 11.5-10
.2.3.7 Deleted by FSARCR 05-MP3-015 ......................................................... 11.5-11
.2.3.8 Regenerant Evaporator Monitor (Removed from Service)..................... 11.5-11
.2.3.9 Waste Neutralization Sump Monitor ...................................................... 11.5-11
.2.4 Inservice Inspection, Calibration, and Maintenance............................... 11.5-11
.2.5 Sampling ................................................................................................. 11.5-12
.3 REFERENCES FOR SECTION 11.5..................................................... 11.5-12 11-iii Rev. 30
List of Tables mber Title
-1 Iodine and Noble Gas Inventory in Reactor Core - Original License Basis (1)
(historical) 1-2 Reactor Coolant Equilibrium Concentrations - Original License Basis (HISTORICAL)
-2A Design Reactor Coolant Equilibrium Concentrations at 3723 MWt
-3 Parameters Used in the Calculation of Reactor Coolant, Secondary Side Liquid, and Secondary Side Steam Fission and Activation Product Activity - Original License Basis
-3A Parameters Used in the Calculation of Design Reactor Fission and Activation Product Activity
-4 Tritium Production
-5 Reactor Coolant N-16 Activity (1)
-6 Secondary Side Liquid Equilibrium Concentrations Original License Basis (HISTORICAL)
-7 Secondary Side Steam Equilibrium Concentrations - Original License Basis (HISTORICAL)
-1 Liquid Waste Management System Daily Input Flows 2-2 Liquid Waste Management System Design Data
-3 Tank Overflow Protection
-4 Expected Radioactive Liquid Concentrations From Each Liquid Release Stream (Ci/ml) (1) Following Treatment (HISTORICAL)
-5 Expected Annual Radioactive Liquid Releases Prior to Dilution in the Circulating Water Discharge System and Prior to Inclusion of Anticipated Operational Occurrences (1) (HISTORICAL)
-6 Expected Annual Radioactive Liquid Releases After Dilution in the Circulating Water Disharge System and Inclusion of Anticipated Operational Occurrences (1) (HISTORICAL)
-7 Design (1) Radioactive Liquid Concentrations From Each Liquid Release Stream (Ci/gm) Following Treatment (2)
-8 Design (1) Annual Radioactive Liquid Releases Prior to Addition of Anticipated Operational Occurances and Dilution in the Circulating Water Discharge System 11-iv Rev. 30
mber Title
-9 Design (1) Annual Radioactive Liquid Releases Following Addition of Anticipated Operational Occurances and Dilution in the circulating Water Discharge System
-10 Fraction of MPC Released - Design Case (1) (HISTORICAL)
-11 Assumptions Used for the Radioactive Liquid Waste System Failure (Release to Atmosphere) and for the Liquid Containing Tank Failure
-12 Boron Recovery Tank Concentrations (Ci/cc)
-13 Activity Released to Atmosphere from a Radioactive Liquid Containing Tank Failure (Boron Recovery Tank)
-14 Radioactive Concentrations in Groundwater Entering Niantic Bay Following a Rupture of Boron Recovery Tank 3-1 Total Expected Radioactive Gaseous Released to Atmosphere from Millstone 3 (HISTORICAL)
-2 Radioactive Gaseous Source Term Parameters (HISTORICAL) 3-3 Radioactive Gaseous Waste System
-4 Codes and Standards
-5 Expected Radioactive Gaseous Releases to Atmosphere via Ventilation Vent (HISTORICAL)
-6 Expected Radioactive Gaseous Releases to Atmosphere from Millstone 3 via Millstone Stack (HISTORICAL)
-7 Expected Radioactive Gaseous Releases to Atmosphere via Turbine Building (HISTORICAL)
-8 Design Radioactive Gaseous Releases to Atmosphere via Ventilation Vent (HISTORICAL)
-9 Design Radioactive Gaseous Releases to Atmosphere from Millstone 3 via Millstone Stack (HISTORICAL)
-10 Design Radioactive Gaseous Releases to Atmosphere via Turbine Building Roof (HISTORICAL)
-11 Design Radioactive Gaseous Releases to Atmosphere from Millstone 3 (HISTORICAL)
-12 Assumptions Used for the Process Gas Charcoal Bed Adsorber Bypass Analysis (1) 11-v Rev. 30
mber Title
-13 Radioisotope Releases from the Process Gas Charcoal Bed Adsorber and Associated Piping
-1 Volume of Spent Resin Generated Annually
-2 Radioactive Solid Waste System Component Data
-3 Omitted
-4 Radioactive Solid Waste Annual Shipments (1)(2)
-1 Gaseous Monitors
-2 Liquid Process Monitors
-3 Radiological Samples Taken at Reactor Plant Sample Sink
.1-1 Annual Doses to Maximum Individual in the Adult Group from Gaseous Effluents
.1-2 Annual Doses to Maximum Individual in the Teen Group from Gaseous Effluents
.1-3 Annual Doses to Maximum Individual in the Child Group from Gaseous Effluents
.1-4 Annual Doses to Maximum Individual in the Infant Group from Gaseous Effluents
.1-5 Annual Doses to Maximum Individual in the Adult Group from Gaseous Effluents
.1-6 Annual Doses to Maximum Individual in the Teen Group from Gaseous Effluents
.1-7 Annual Doses to Maximum Individual in the Child Group from Gaseous Effluents
.1-8 Annual Doses to Maximum Individual in the Infant Group from Gaseous Effluents
.1-9 Omitted
.1-10 Omitted
.1-11 Omitted
.1-12 Omitted
.1-13 Annual Doses to Maximum Individual in the Adult Group from Liquid Effluents 11-vi Rev. 30
mber Title
.1-14 Annual Doses to Maximum Individual in the Teen Group from Liquid Effluents
.1-15 Annual Doses to Maximum Individual in the Child Group from Liquid Effluents
.1-16 Comparison of Maximum Calculated Doses from Millstone 3 Nuclear Plant with Appendix I Design Objectives
.1-17 Calculated Population Dose
.2-1 Dilution Factors, Travel Times from the Site, and Population Served
.2-2 Parameters and Assumptions Used in Equations for Estimating Doses to Humans
.2-3 Meteorogical Data
.2-4 Parameters and Assumptions Used in Estimating Doses to Biota
.3-1 Base Case Annual Population Doses Due to Liquid Effluents (1)
.3-2 Base Case Annual Population Doses Due to Gaseous Effluents (1) 11-vii Rev. 30
List of Figures mber Title
-1 Radioactive Liquid Waste and Aerated Drains
-2 Condensate Demineralizer Liquid Waste
-3 Expected Radioactive Liquid Waste Source and Discharge Paths
-1 Radioactive Gaseous Waste
-2 Ventilation System Composite Drawing Normal Operation
-1 Radioactive Solid Waste
-2 (HISTORICAL) Radioactive Solid Waste System Expected Quantities
-3 (HISTORICAL) Radioactive Solid Waste System Design Quantities 11-viii Rev. 30
BACKGROUND estimate of the radioactive effluents and public dose is provided that documents projected lic dose consequences are within 10 CFR 50 Appendix I radioactive release criteria. The mates were based on nominal assumptions and generic models and based on plant operations core power level of 3636 MWt. These dose estimates were developed in support of the inal license and updated during MPS-3 restart. These dose estimates are historical and not ject to future update. This information is retained to avoid loss of original design basis.
Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) vide guidance requirements for system operation, dose calculations, and monitoring uirements, and to ensure compliance with effluent limits. Actual measured concentrations of oactivity released and real time dilution or dispersion estimates are required to verify pliance with effluent limits. Therefore, it is operation within the requirements of the MODCM that ensures compliance with effluent limits, rather than operation to the nominal mptions in this chapter.
Annual Radioactive Effluent Release Report and Annual Radiological Environmental rating Report, which are required to be submitted in May of each year, document that ration of the plant complies with effluent limits and appropriate regulations. Technical cifications requires implementation of the REMODCM and Radiation Environmental nitoring Program.
11.0-1 Rev. 30
source of all radioactivity occurring in the process streams of the various radioactive systems e radionuclides generated in the reactor core and neutron activation of nuclides in the reactor lant system (RCS) and the air surrounding the reactor vessel.
ioactive liquid and gaseous releases are described in Sections 11.2 and 11.3, respectively. All uction factors and decontamination factors associated with these systems are discussed in the ropriate sections.
.1 RADIONUCLIDE INVENTORY IN THE CORE discussion on core inventory presented below represents information used by the original nse application to establish plant shielding and radwaste effluent assessments. It is retained for orical purposes to avoid loss of original design basis and is not subject to future update. The inventory used for accident analyses is presented in Chapter 15.
specific activity of fission products in the core is calculated using the computer program TIVITY2. This program calculates the contribution from parent, daughter, and nddaughter isotopes by solving the following differential equations:
- 1. First order nuclides:
dN ci ( t )
= F i - ( i + h i + i )N ci ( t ) (11.1-1) dt
- 2. Second order nuclides:
dN cj ( t )
= F j + ( i f ij N ci ( t ) ) - ( j + h j + j )N cj ( t ) (11.1-2) dt
- 3. Third order nuclides:
dN ck ( t )
= F k + i f ik N ci ( t ) + j f jk N cj ( t ) (11.1-3) dt
- ( k + h k + k )N ck ( t )
re:
i, j, k = indicate first, second, and third order nuclide parameters N c i ( t ) = concentration of nuclide i per fuel region at time t (atoms/region) t = time (sec) 11.1-1 Rev. 30
i = fission yield for isotope i (atoms/fission) i = decay constant for isotope i (sec-1) i = escape rate coefficient (sec-1) i = a th i = r a th = burnup rate (sec-1) i i fij = branching fraction from i to j h = fraction of failed fuel program has a basic library of 167 nuclides with a capability of 200 nuclides. Library data ude decay scheme information, production information, purification factors for typical ineralizers, and fuel escape rate coefficients. The library also contains decay gamma spectra even energy groups. Input data include time intervals, initial source inventory in the fuel, tron flux, power level, fraction of fuel defects, and density of reactor coolant. The program cribes the system analyzed, as well as the operating history, the activities, and associated ma spectral information as a function of time. Fuel assembly source terms for shielding gn, given in Chapter 12, are calculated by this method.
calculation of the core iodine fission product inventory is consistent with the inventories n in TID-14844 (DiNunno et al., 1962). The core iodine and noble gas fission product entories in Table 11.1-1 are based on continuous operation of the unit at 3,636 MWt (design output plus 2.0 percent instrumentation error). The core inventory source terms used for pter 15 radiological accident analyses have been revised to reflect requirements as described Regulatory Guide 1.183 and is discussed in Section 15.0.9.1.
core inventory information listed in Table 11.1-1 is being retained for historical purposes ause that was the basis for original plant shielding design as described in Section 12.2.1.
l element heat loadings and stresses as well as fuel operating experience are presented in pter 4.
.2 RADIONUCLIDE INVENTORY IN FUEL ELEMENT GAP gap activity is that fraction of the gaseous activity in the core that diffuses to the fuel gaps.
fuel element gap source terms used for Chapter 15 radiological accident analyses have been sed to reflect requirements as described by Regulatory Guide 1.183 and is discussed in tion 15.0.9.2.
11.1-2 Rev. 30
.3.1 Fission Product Activities design basis fission product activities in the reactor coolant resulting from fuel defects ciated with 1-percent power are also calculated with the ACTIVITY2 program. The owing differential equations are used:
- a. First order nuclides:
dN wi hn PF EQi Q 1 T
= ----------i N ci ( t ) - i + --------------------
- + i ---- N wi ( t ) (11.1-4) dt Vw Vw T 2
- b. Second order nuclides:
dN wj hn j
= ---------- N cj ( t ) + i f ij N wi ( t ) (11.1-5) dt Vw PF EQj Q 1 T1
- j + --------------------
- + j ------ N wj ( t )
V T 2 w
- c. Third order nuclides:
dN wk hn k
= ----------- N ck ( t ) + i f ik N wi ( t ) + j f jk N wj ( t ) (11.1-6) dt Vw PF EQk Q 1 T1
- k + ---------------------
- + k ------ N wk ( t )
V T 2 w
re:
N wi ( t ) = concentration of nuclide i in the main coolant at time t (atoms/cm3) n = total number of fuel regions Vw = volume of main coolant (cm3)
PFEQ = equivalent purification factor (fraction) for i i = ai th = Burnup rate (sec-1)
Q1 = equivalent flow into purification stream (cm3/sec)
T1 = coolant residence time in core (sec)
T2 = coolant circulation time (sec) 11.1-3 Rev. 30
lication to establish plant shielding and radwaste effluent assessments. This data is retained for orical purposes to avoid loss of original design basis and is not subject to future update.
hese calculations, the defective fuel rods were assumed to be present in the initial core and ormly distributed throughout the core. Thus, the fission product escape rate coefficients were ed upon average fuel temperature. Calculations are performed using the average temperature he reactor coolant. The reactor coolant density correction of 1.4 was made in order to obtain correct radionuclides concentration downstream of the letdown heat exchanger.
o included in Table 11.1-2 are the expected equilibrium concentrations for the RCS. These lts were based on measured and calculated concentrations given in NUREG-0017 (USNRC
- 6) and the parameters listed in Table 11.1-3.
expected reactor coolant activities were used to develop the source terms for gaseous and id effluents in Sections 11.2 and 11.3.
h expected and design reactor coolant activities were used to evaluate the ventilation design in pter 12.
methodology to calculate the current design basis primary coolant activity concentrations is ilar to that used during original licensing basis with a minor modification. The design basis inventory is calculated by industry computer code ORIGEN instead of ACTIVITY2. Since source of primary coolant fission product activity is the leakage of core activity via the ctive fuels, the primary coolant activity concentrations calculated by ACTIVITY2 are sted by the ratio of the core inventory developed by ORIGEN, and presented in Chapter 15, to core inventory calculated by ACTIVITY2. The coolant concentration for a given isotope ends on the core inventory, the escape coefficients of the isotope and its precursor isotopes, the depletion rate of the isotopes. For each isotope in the primary coolant, the decay chain and escape coefficients are examined, and those isotopes that have significant impact on the centration of the referenced isotope in the coolant are identified. Amongst these major tributors, the ratios of the ORIGEN core activity to the ACTIVITY2 core activity are erally very close to each other, and the maximum scaling factor is used to adjust the TIVITY2 based coolant concentration to reflect the core developed by ORIGEN.
current design basis reactor coolant equilibrium activities presented in Table 11.1-2A, are ed on parameters listed in Table 11.1-3A.
.3.2 Tritium Activity re are two principal contributors to tritium production within the pressurized water reactor em: the ternary fission source and the dissolved boron in the reactor coolant. Additional tributions are made by Li-6, Li-7, and deuterium in the reactor water. Tritium is also produced uclear reactions with boron contained in burnable poison rods. Tritium production from erent sources is shown in Table 11.1-4.
11.1-4 Rev. 30
s tritium is formed within the fuel material and may:
- 1. Remain in the fuel rod uranium matrix
- 2. Diffuse into the cladding and become fixed there, as zirconium tritide
- 3. Diffuse through the cladding and be released into the primary coolant, or
- 4. Be released to the coolant through microscopic cracks or failures in the fuel cladding vious WNES fuel design has conservatively assumed that the ratio of fission tritium released the coolant to the total fission tritium formed was approximately 0.30 for Zircaloy clad fuel.
operating experience at the R.E. Ginna Plant of the Rochester Gas and Electric Company, and ther operating pressurized water reactors using zircaloy clad fuel, has shown that the tritium ase through the Zircaloy fuel cladding is less than the earlier estimates. Consequently, the ase fraction has been revised downward from 30 to 10 percent based on these data CAP-8253 1974).
ic Acid Source irect contribution to the reactor coolant tritium concentration is made by neutron reaction with boron in solution. The concentration of boric acid varies with core life and load follow so that is a steadily decreasing source during core life. The principal boron reactions are B-10(n,2) and B-10(n,)Li-7(n,n)H-3 reactions. The Li-7 reaction is controlled by limiting the overall um concentration to approximately 2 ppm during operation.
nable Poison Rod Source se rods are in the core only during the first operating cycle and their potential tritium tribution is only during this period.
ium and Deuterium ium and deuterium reactions contribute only minor quantities to the tritium inventory (Table
-4). These sources are due to the activation of the lithium and deuterium in the RCS as they s through the reactor. Lithium-6 is essentially excluded from the system by using 99.9 percent ign Bases design intent is to reduce the tritium sources in the RCS to a practical minimum to permit ger retention of the reactor coolant within the plant without compromising operator exposures.
uction of source terms is provided by using hafnium or Ag-In-Cd control rods instead of B4C 11.1-5 Rev. 30
ign Evaluation le 11.1-4 compares a typical design basis tritium production which has been used in the past to blish system and operational requirements of the plant and present expected values. There are principal contributors to the expected tritium release to the RCS: ternary fission source and dissolved boron in the reactor coolant.
ause of the importance of the ternary fission source on the operation of the plant, WNES has n closely following operating plant data at the R. E. Ginna Plant. The R. E. Ginna Plant has a aloy clad core with silver-indium-cadmium control rods. The operating levels of boron centration during the startup of the plant are approximately 1,100 to 1,200 ppm of boron. In ition, burnable poison rods in the core contain boron which contributes some tritium to the lant, but only during the first cycle. Data during the operation of the plant have very clearly cated that the present design sources were conservative. The tritium released is essentially m the boron dissolved in the coolant and a ternary fission source which is less than 10 percent.
ddition to these data, other operating plants with Zircaloy clad cores have also reported low um concentrations in the RCS after considerable periods of operation.
s quantity of tritium becomes uniformly distributed in the RCS, the primary grade water em, and the boron recovery system. During refueling operations, the tritium is further diluted n the refueling canal is filled from the borated refueling water storage tank.
expected tritium concentration in the primary coolant is based on the value provided in REG-0017 (USNRC 1976). For radioactive liquid waste analysis, it is assumed that 50 percent he activity produced and entering into the coolant in 1 year is released in the radioactive liquid te system effluent. The remaining 50 percent is released in the radioactive gaseous waste em and ventilation effluents.
design tritium concentration in the primary coolant is selected to allow limited access to the tainment during normal operation. The water management plan controls tritium centrations to design levels and also allows for continuous containment access during eling with operation of the containment ventilation system.
ed 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 full power operation is reduced by using hafnium control rods.
11.1-6 Rev. 30
rosion products in the reactor coolant become activated when they pass through the core. The t important corrosion products are Cr-51, Mn-54, Fe-55, Fe-59, Co-58, and Co-60. The osion product activity is dependent on many factors, including the type of plant and the erials of construction. The mass transport process is complex and stochastic, and calculational hods to predict corrosion product activity accurately have not been successfully correlated h operational data (Bartlett 1969). Analytical predictions of the corrosion product activity ls are approximations. Therefore, design corrosion product levels are assumed to be the es measured at operating reactors. The design corrosion product levels are based on REG-0017 (USNRC 1976) values modified by the appropriate adjustment factor described ein.
corrosion product activities in the reactor coolant are given in Table 11.1-2, for the original nse and Table 11.1-2A for current conditions.
.3.4 Nitrogen-16 Activity ogen-16 is a concern only during reactor operation because of its short half-life (7.1 seconds).
ogen-16 is produced in circulating primary coolant entering the core region and irradiated by trons. Reactions with all three oxygen isotopes 0-16 (99.76 percent), 0-17 (0.037 percent), and 8 (0.204 percent) result in the production of N-16.
ogen-16 emits high energy gammas in 75-percent of the disintegrations (70-percent at 6.13 V and 5 percent at 7.11 MeV).
N-16 activity at various points in the RCS is given in Table 11.1-5.
.4 RADIOACTIVITY IN THE SECONDARY SIDE concentrations of principal radioisotopes in the secondary side of the steam generators are d for both the design and expected cases in Table 11.1-6 for liquid and Table 11.1-7 for steam.
se tables present secondary coolant and steam data used by the original license application to blish plant shielding and radwaste effluent assessments. This data is retained for historical poses to avoid loss of original design basis and is not subject to future update.The design lts for fission and activation products, based on parameters in Table 11.1-3, were calculated h the computer program IONEXCHANGER which solves the following differential equations secondary liquid activities.
- 1. First order nuclides:
dN i Q
= R i - i + ------B- N i ( t ) (11.1-7) dt V
- 2. Second order nuclides:
11.1-7 Rev. 30
V
- 3. Third order nuclides:
dN k Q
= R k + i f ik N i ( t ) + j f jk N j ( t ) - k + ------B- N k ( t ) (11.1-9) dt V re:
Ni = number of atoms of nuclide i (atoms)
Ri = feed rate of nuclide i (atoms/sec) i = radioactive decay constant for nuclide i (sec-1) fij = branching fraction from i to j QB = steam generator radioactivity removal rate (cm3/sec)
V = volume of steam generator liquid (cm3) t = time in seconds ondary side steam activities are obtained by using the following relationship:
A i = P i A oi (11.1-10) re:
Ai = steam equilibrium activity for isotope i (Ci/gm)
A oi = liquid equilibrium activity for isotope i (Ci/gm) expected secondary liquid and steam activities are based on the concentrations reported in REG-0017 and the parameters listed in Table 11.1-3.
.5 REFERENCES FOR SECTION 11.1
-1 Barlett, J.W. 1969. Stochastics of Coolant Crud. In: ANS Transactions, Vol 12.
-2 DiNunno, 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.
-3 U.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.
11.1-8 Rev. 30
11.1-9 Rev. 30 TABLE 11.1-1 IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE -
ORIGINAL LICENSE BASIS (1) (HISTORICAL)
Isotope Core (Ci)
I-131 9.1E+07 I-132 1.3E+08 I-133 2.0E+08 I-134 2.4E+08 I-135 1.9E+08 Kr-83m 1.6E+07 Kr-85m 4.0E+07 Kr-85 8.8E+05 Kr-87 7.7E+07 Kr-88 1.1E+08 Kr-89 1.4E+08 Xe-131m 8.0E+04 Xe-133m 4.9E+06 Xe-33 2.0E+08 Xe-135m 5.5E+07 Xe-135 5.4E+07 Xe-137 1.8E+08 Xe-138 1.8E+08 TES:
Based on 650 days of operation at 3,636 MWt. Historical, not subject to future updating.
This table has been retained to preserve original license basis.
Page 1 of 1 Rev. 30
TABLE 11.1-2 REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS -
ORIGINAL LICENSE BASIS (HISTORICAL)
Nuclide Design (Ci/g) Expected (Ci/g) ble Gases 83m 4.5E-01 (1) 2.2E-02 85m 1.7E+00 9.2E-02 85 3.4E-02 2.1E-03 87 1.2E+00 6.6E-02 88 3.4E+00 1.9E-01 89 1.1E-01 6.2E-03 131m 1.1E-02 5.5E-03 133m 6.1E-01 4.1E-02 133 2.6E+01 1.7E+00 135m 1.2E+00 1.6E-02 135 5.1E+00 2.2E-01 137 1.7E-01 1.1E-02 138 6.0E-01 5.4E-02 btotal: 4.1E+01 2.4E+00 logens 83 9.0E-02 5.8E-03 84 4.1E-02 3.2E-03 85 5.7E-03 3.8E-04 130 7.2E-03 2.4E-03 131 2.6E+00 2.9E-01 132 9.3E-01 1.2E-01 133 4.2E+00 4.3E-01 134 5.8E-01 5.8E-02 135 2.2E+00 2.2E-01 btotal: 1.1E+01 1.1E+00 Page 1 of 3 Rev. 30
Nuclide Design (Ci/g) Expected (Ci/g) rrosion Products 51 2.0E-03 2.0E-03 54 3.3E-04 3.3E-04 55 1.7E-03 1.7E-03 59 1.1E-03 1.1E-03 58 1.7E-02 1.7E-02 60 2.1E-03 2.1E-03 btotal: 2.4E-02 2.4E-02 her Nuclides 86 2.7E-04 9.1E-05 88 3.4E+00 2.5E-01 89 4.3E-03 3.7E-04 90 1.7E-04 1.0E-05 91 2.0E-03 7.5E-04 90 2.1E-04 1.3E-06 91m 1.1E-03 4.5E-04 91 6.9E-04 6.7E-05 93 3.4E-04 3.9E-05 95 7.1E-04 6.3E-05 95 7.4E-04 5.3E-05 99 3.4E+00 9.0E-02 99m 1.9E+00 5.6E-02 103 3.4E-04 4.7E-05 106 3.3E-05 1.0E-05 103m 3.4E-04 5.6E-05 106 3.3E-05 1.3E-05 125m 9.0E-05 3.0E-05 127m 2.1E-03 2.9E-04 127 1.1E-03 9.8E-04 Page 2 of 3 Rev. 30
Nuclide Design (Ci/g) Expected (Ci/g) 129m 3.9E-02 1.5E-03 129 2.2E-02 2.0E-03 131m 2.3E-02 2.7E-03 131 1.1E-02 1.4E-03 132 2.7E-01 2.9E-02 134 3.3E-01 2.7E-02 136 1.7E-01 1.4E-02 137 1.7E+00 1.9E-02 137m 1.5E+00 2.0E-02 140 4.4E-03 2.3E-04 140 1.5E-03 1.6E-04 141 7.0E-04 7.4E-05 143 5.2E-04 4.4E-05 144 5.0E-04 3.5E-05 143 6.8E-04 5.3E-05 144 5.0E-04 4.1E-05 239 3.9E-03 1.3E-03 btotal: 1.4E+01 5.2E-01 3 3.5E+00 1.0E+00 tal (Excluding H-3) 5.8E+01 4.0E+00 tal (Including H-3) 6.2E+01 5.0E+00 TE:
4.5E-01 = 4.5 x 10-1 torical, not subject to future updating. This table has been retained to preserve original license basis.
Page 3 of 3 Rev. 30
BLE 11.1-2A DESIGN REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS AT 3723 MWt Nuclide Primary Coolant Activity Concentration (Ci/g)
-83m 3.53E-01
-85m 1.09E+00
-85 1.33E-01
-87 8.49E-01
-88 2.22E+00
-89 6.96E-02
-131m 1.93E-01
-133m 7.68E-01
-133 2.54E+01
-135m 9.45E-01
-135 5.50E+00
-137 1.94E-01
-138 6.54E-01 83 7.05E-02 84 3.50E-02 85 3.68E-03 87 1.89E-03 29 1.79E-07 30 4.39E-02 31 2.67E+00 32 1.09E+00 33 4.06E+00 34 6.19E-01 35 2.39E+00 36 6.73E-03 81 5.84E-07 83 7.94E-07 84 4.60E-07 Page 1 of 5 Rev. 30
Nuclide Primary Coolant Activity Concentration (Ci/g)
-86 1.46E-01
-88 2.32E+00
-89 1.45E-01
-90 1.12E-02
-91 5.70E-03
-92 3.88E-04 89 3.02E-03 90 1.96E-04 91 1.30E-03 92 9.45E-04 93 4.41E-05 94 7.49E-06 90 3.84E-04 91m 7.77E-04 91 1.43E-02 92 1.12E-03 93 6.13E-04 94 2.72E-05 95 1.14E-05 95 5.72E-04 97 3.67E-04
-95m 6.58E-06
-95 5.78E-04
-97m 3.48E-04
-97 3.91E-04
-99 5.27E+00
-101 2.10E-02
-102 1.52E-02
-105 7.22E-04 Page 2 of 5 Rev. 30
Nuclide Primary Coolant Activity Concentration (Ci/g) 99m 2.73E+00 101 2.05E-02 102 1.53E-02 105 7.66E-04
-103 5.49E-04
-105 1.34E-04
-106 2.02E-04
-107 1.90E-06
-103m 5.50E-04
-105m 3.82E-05
-105 3.44E-04
-106 2.25E-04
-107 1.14E-05
-127 2.41E-06
-128 5.04E-06
-130 8.31E-07
-127 2.77E-05
-128 2.66E-06
-129 3.73E-05
-130 2.67E-06
-131 1.16E-05
-132 8.85E-07
-133 1.04E-06 125m 3.98E-04 127m 3.11E-03 127 1.10E-02 129m 1.32E-02 129 1.35E-02 131m 3.34E-02 Page 3 of 5 Rev. 30
Nuclide Primary Coolant Activity Concentration (Ci/g) 131 1.26E-02 132 2.77E-01 133m 1.91E-02 133 8.60E-03 134 2.91E-02
-134m 4.72E-02
-134 2.31E+01
-136 3.52E+00
-137 1.62E+01
-138 1.00E+00
-139 8.99E-02
-140 9.09E-03
-142 1.08E-04
-137m 1.52E+01
-139 7.79E-02
-140 3.75E-03
-141 1.21E-04
-142 1.66E-04
-140 1.26E-03
-141 2.61E-04
-142 2.44E-04
-143 1.38E-05
-141 5.65E-04
-143 4.19E-04
-144 4.46E-04
-145 2.09E-06
-146 7.74E-06 143 5.20E-04 144 4.50E-04 Page 4 of 5 Rev. 30
Nuclide Primary Coolant Activity Concentration (Ci/g) 145 1.49E-04 146 2.04E-05
-147 2.22E-04
-149 2.28E-05
-151 1.66E-06
-147 1.13E-04
-149 1.97E-04
-151 5.48E-05
-151 7.50E-07
-153 1.55E-04 51 2.00E-03
-54 3.30E-04 55 1.70E-03 59 1.10E-03
-58 1.70E-02
-60 2.10E-03
-239 1.98E-02 3 3.5E+00 Page 5 of 5 Rev. 30
TABLE 11.1-3 PARAMETERS USED IN THE CALCULATION OF REACTOR OOLANT, SECONDARY SIDE LIQUID, AND SECONDARY SIDE STEAM FISSION AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS Parameter Value (1) re thermal power 3,636 MWt cludes 2% instrumentation error) rcent fuel defects (design case) 1.0 rcent fuel defects (expected case) 0.12 sion 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) actor coolant liquid mass 4.70 x 105 lb ithout pressurizer) actor coolant liquid volume 10,920 ft3 ithout pressurizer) actor coolant liquid volume 12,000 ft3 ith pressurizer) actor coolant full power average 590°F perature rification flow rate, normal 75 gpm xed bed demineralizer decontamination tors Noble Gases, N-16, H-3 1.0 Cations Design: Cs, Mo, Y 1.0 Expected: Cs, Rb 2.0 All other nuclides including activation 10.0 products Page 1 of 5 Rev. 30
AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS Parameter Value (1) tion bed demineralizer decontamination tors Noble Gases, N-16, Halogens, H-3 1.0 Cations Design: Cs, Y, Mo 10.0 Expected: Cs, Rb 10.0 All other nuclides including activation products Design 1.0 Expected 10.0 Expected halogens 10.0 tio of cation bed demineralizer flow to 0.1 rification bed demineralizer flow actor coolant letdown discharged via ron recovery system Design 500 lb/hr Expected 500 lb/hr am flow rate 1.589 x 107 lb/hr mary to secondary leak rate Design 1,370 lb/day Expected 100 lb/day am generator partition factor circulating U-tube)
Noble Gases: N-16, H-3 1.0 Halogens 0.01 Cations Design: Cs, Rb 0.0025 Expected: Cs, Rb 0.001 Others Design 0.0025 Page 2 of 5 Rev. 30
AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS Parameter Value (1)
Expected 0.001 ndensate polishing demineralizer contamination factors Cations Design: Cs, Y, Mo 2.0 Expected: Cs, Rb 2.0 All other nuclides including activations products Design 10.0 Expected 10.0 ndensate polishing flow rate 9.89 x 106 lb/hr ermal neutron flux 2.0 x 1013 n/cm2-sec erating time (650 EFPD) 15,600 hr olant cycle time 9.9 sec olant in core time 0.7 sec gasification factor 1.0 condary side equilibrium time 104 hr mber of condensate demineralizers 7 plus 1 spare lume of each condensate demineralizer 197 ft3 pe of condensate demineralizers Deep bed generation time for each demineralizer Design 3.5 days (estimate)
Expected 7.0 days mber of regenerations/yr for condensate ishing demineralizer Design 100 (estimate)
Expected 56 lume control tank volumes Page 3 of 5 Rev. 30
AND ACTIVATION PRODUCT ACTIVITY - ORIGINAL LICENSE BASIS Parameter Value (1)
Vapor 240 ft3 Liquid 160 ft3 tal secondary fluid per steam generator Expected Design Liquid 99,253 lb 103,000 lb Steam 8,437 lb 8,000 lb Total 107,690 lb 111,000 lb pothetical Design flowrates for the steam nerator blowdown system (see Section 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) pothetical Expected flowrates for the am generator blowdown system 1% MSR from each steam generator 148,000 lb/hr (37,000 lb/hr per steam generator) ction removed from steam generator wdown (purification factors for design d expected cases)
Noble Gases 0.0 Halogens 0.8505 Cs, Rb 0.4975 Others 0.8955 Tritium 0.0 tio of condensate demineralizer flow rate 0.6224 the total steam flow rate Page 4 of 5 Rev. 30
Rev. 30 TABLE 11.1-3A PARAMETERS USED IN THE CALCULATION OF DESIGN REACTOR FISSION AND ACTIVATION PRODUCT ACTIVITY Parameter Value re thermal power (includes 2% instrumentation error) 3,723 MWt rcent fuel defects 1.0 sion 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) actor coolant average density 44.39 lbs/ft3 actor coolant liquid volume (without pressurizer and surge line) 10,100 ft3 actor coolant full power average temperature 594.5°F rification flow rate, normal 82 gpm xed bed demineralizer decontamination factors Noble Gases, N 16, H 3 1.0 Cations Cs, Mo, Y 1.0 Halogens 10.0 All other nuclides including activation products 10.0 tion bed demineralizer decontamination factors Noble Gases, N 16, H 3 1.0 Cations Cs, Mo, Y 10.0 Halogens 1.0 All other nuclides including activation products 1.0 Page 1 of 2 Rev. 30
Parameter Value tio of cation bed demineralizer flow to purification bed 0.01 mineralizer flow actor coolant letdown discharged via boron recovery system 500 lb/hr ermal neutron flux - core / coolant 3.83 x 1013 n/cm2sec actor operating time (assumed 2 cycles) 28,800 hr olant cycle time 9.74 sec olant in core time 0.721 sec gasification factor 1.0 sign corrosion product concentrations in RCS Cr 51 2.0E-3 Ci/gm Mn 54 3.3E-4 Ci/gm Fe 55 1.7E-3 Ci/gm Fe 59 1.1E-3 Ci/gm Co 58 1.7E-2 Ci/gm Co -60 2.1E-3 Ci/gm Np 239 2.2E-3 Ci/gm sign Tritium concentration 3.5 Ci/gm Page 2 of 2 Rev. 30
Release Expected to Reactor Coolant Tritium Source Total Produced (Ci/yr) (Ci/yr) nary fissions Initial cycle 14,000 1,400 Equilibrium cycle 10,900 1,090 rnable poison rods Initial cycle 1,950 195 olant (soluble boron)
Initial cycle 388 388 Equilibrium cycle 285 285 olant lithium, deuterium Initial cycle 141 141 Equilibrium cycle 109 109 tal initial cycle 16,470 2,124 tal equilibrium cycle 11,294 1,484 TES:
ues presented are for a typical 3,565 MWt PWR. However, calculated tritium releases are ed on 102 percent of this value (3,636 MWt) in order to comply with Regulatory Guide 1.49 ction 1.8.1.49). Calculation of tritium releases are based on methodology presented in REG-0017.
ease fraction from fuel, 10 percent ease fraction from burnable poison rods, 10 percent ght of boron-10 in burnable poison rods, 6,160 gm ial cycle boron, 900 ppm ilibrium cycle boron, 1,100 ppm ium concentration (99.9 atom percent lithium-7), 2.2 ppm ial cycle operating time, 9,240 effective full-power hours ilibrium cycle operating time, 7,200 effective full-power hours duction in control rods is based on continuous daily load follow (12, 3, 6, 3 cycle). During e load full power operation, the production would be negligible.
Page 1 of 1 Rev. 30
Position in Loop Loop Transit Time (sec) N-16 Activity (Ci/g) aving core 0.0 189 aving reactor vessel 1.1 170 tering steam generator 1.4 164 aving steam generator 5.4 112 tering reactor coolant pump 6.0 106 tering reactor vessel 6.8 98 tering core 9.0 86 aving core 9.7 189 e:
These values are based on a typical 3,565 Mwt PWR.
Page 1 of 1 Rev. 30
ABLE 11.1-6 SECONDARY SIDE LIQUID EQUILIBRIUM CONCENTRATIONS ORIGINAL LICENSE BASIS (HISTORICAL)
Nuclide Design (1) (Ci/g) Expected (2) (Ci/g)
Br 83 1.5E-05 (3) 7.3E-08 Br 84 3.1E-06 1.9E-08 Br 85 5.2E-08 2.7E-10 I 130 1.7E-06 4.1E-08 I 131 6.9E-04 5.4E-06 I 132 2.0E-04 2.0E-06 I 133 1.0E-03 7.7E-06 I 134 6.3E-05 4.5E-07 I 135 4.9E-04 3.6E-06 Cr 51 7.3E-07 5.6E-08 Mn 54 1.2E-07 1.2E-08 Fe 55 6.2E-07 5.0E-08 Fe 59 4.0E-07 3.8E-08 Co 58 6.2E-06 5.0E-07 Co 60 7.6E-07 5.6E-08 Rb 86 1.8E-07 4.6E-09 Rb 88 1.8E-04 1.0E-06 Sr 89 1.6E-06 1.3E-08 Sr 90 6.2E-08 2.5E-10 Sr 91 6.1E-06 1.6E-08 Y 90 7.5E-08 5.3E-11 Y 91m 3.7E-07 1.2E-08 Y 91 2.5E-07 1.9E-09 Y 93 1.1E-07 7.8E-10 Zr 95 2.6E-07 2.5E-09 Nb 95 2.7E-07 2.5E-09 Mo 99 1.2E-03 2.6E-06 Tc 99m 7.9E-04 2.5E-06 Page 1 of 3 Rev. 30
Nuclide Design (1) (Ci/g) Expected (2) (Ci/g)
Ru 103 1.3E-07 1.3E-09 Ru 106 1.2E-08 2.5E-10 Ru 103m 1.2E-07 2.3E-09 Rh 106 1.2E-08 0.0 Te 125m 3.3E-08 6.3E-10 Te 127m 7.7E-07 6.3E-09 Te 127 4.5E-07 2.4E-08 Te 129m 1.4E-05 3.8E-08 Te 129 8.8E-06 6.7E-08 Te 131m 8.1E-06 6.9E-08 Te 131 2.2E-06 2.5E-08 Te 132 9.6E-05 6.5E-07 Cs 134 2.2E-04 1.4E-06 Cs 136 1.1E-04 7.3E-07 Cs 137 1.1E-03 1.0E-06 Ba 137m 5.7E-04 1.2E-06 Ba 140 1.6E-06 6.3E-09v La 140 5.9E-07 4.7E-09 Ce 141 2.5E-07 2.5E-09 Ce 143 1.8E-07 6.8E-10 Ce 144 1.8E-07 1.2E-09 Pr 143 2.5E-07 1.3E-09 Pr 144 1.8E-07 2.6E-09 Np 239 1.4E-06 4.0E-08 H3 1.3E-03 1.0E-03 tal (excluding H-3) 6.8E-03 3.1E-05 tal (including H-3) 8.0E-03 1.0E-03 Page 2 of 3 Rev. 30
Rev. 30 ABLE 11.1-7 SECONDARY SIDE STEAM EQUILIBRIUM CONCENTRATIONS -
ORIGINAL LICENSE BASIS (HISTORICAL)
Nuclide Design (1) (Ci/g) Expected (2) (Ci/g)
Kr-83m 1.6E-06 (3) 5.7E-09 Kr-85m 6.2E-06 2.4E-08 Kr-85 1.2E-07 5.6E-10 Kr-87 4.4E-06 1.7E-08 Kr-88 1.2E-05 4.9E-08 Kr-89 3.8E-07 1.6E-09 Xe-131m 4.0E-08 1.5E-09 Xe-133m 2.2E-06 1.1E-08 Xe-133 9.4E-05 4.4E-07 Xe-135m 4.1E-06 4.1E-09 Xe-135 1.8E-05 5.6E-08 Xe-137 6.0E-07 2.9E-09 Xe-138 2.1E-06 1.4E-08 Br-83 1.5E-07 7.3E-10 Br-84 3.1E-08 1.9E-10 Br-85 5.2E-10 2.7E-12 I-130 1.7E-08 4.1E-10 I-131 6.9E-06 5.4E-08 I-132 2.0E-06 2.0E-08 I-133 1.0E-05 7.7E-08 I-134 6.3E-07 4.5E-09 I-135 4.9E-06 3.6E-08 Cr-51 1.8E-09 5.6E-11 Mn-54 3.0E-10 1.2E-11 Fe-55 1.6E-09 5.0E-11 Fe-59 9.9E-10 3.8E-11 Co-58 1.5E-08 5.0E-10 Co-60 1.9E-09 5.6E-11 Page 1 of 3 Rev. 30
Nuclide Design (1) (Ci/g) Expected (2) (Ci/g)
Rb-86 4.4E-10 4.8E-12 Rb-88 4.6E-07 1.0E-09 Sr-89 3.9E-09 1.3E-12 Sr-90 2.5E-10 2.5E-13 Sr-91 1.5E-09 1.6E-11 Y-90 1.9E-10 5.6E-14 Y-91m 9.3E-10 1.2E-11 Y-91 6.3E-10 1.9E-12 Y-93 2.7E-10 7.8E-13 Zr-95 6.5E-10 2.5E-12 Nb-95 6.7E-10 2.5E-12 Mo-99 3.0E-06 2.6E-09 Tc-99m 2.0E-06 2.5E-09 Ru-103 1.1E-10 1.3E-12 Ru-106 3.0E-11 2.5E-13 Rh-103m 2.9E-10 2.3E-12 Rh-106 3.0E-11 0.0 Te-125m 8.2E-11 6.3E-13 Te-127m 1.9E-09 6.3E-12 Te-127 1.1E-09 2.4E-11 Te-129m 3.6E-08 3.8E-11 Te-129 2.2E-08 6.7E-11 Te-131m 2.0E-08 6.9E-11 Te-131 5.5E-09 2.5E-11 Te-12 2.4E-07 6.5E-10 Cs-134 5.4E-07 1.4E-09 Cs-136 2.8E-07 7.3E-10 Cs-137 2.7E-06 1.0E-09 Ba-137m 1.4E-06 1.2E-09 Page 2 of 3 Rev. 30
Nuclide Design (1) (Ci/g) Expected (2) (Ci/g)
Ba-140 4.0E-09 6.3E-12 La-140 4.5E-09 4.7E-12 Ce-141 6.4E-10 2.5E-12 Ce-143 4.5E-10 6.8E-13 Ce-144 4.5E-10 1.2E-12 P-143 6.2E-10 1.3E-12 P-144 4.5E-10 2.6E-12 Np-239 3.4E-09 4.0E-11 H-3 1.3E-03 1.0E-03 tal (Excluding H 3): 1.8E-04 8.3E-07 tal (Including H 3): 1.5E-03 1.0E-03 TES:
Based on 1,370 lb/day primary to secondary leak rate.
Based on 100 lb/day primary to secondary leak rate.
1.6E-06 = 1.6 x 10-6 torical, not subject to future updating. This table has been retained to preserve original design s.
Page 3 of 3 Rev. 30
ccordance with General Design Criterion 60, liquid waste management systems are provided ontrol, collect, process, store, recycle, and dispose of liquid radioactive waste generated as the lt of normal plant operation, including anticipated operational occurrences. The liquid waste agement systems include the radioactive liquid waste system (LWS) and condensate ineralizer liquid waste system (LWC) (which has been removed from service). Figures 11.2-1 11.2-2 are the piping and instrumentation diagrams of the radioactive liquid waste system and densate demineralizer liquid waste systems, respectively.
boron recovery system (Section 9.3.5) also processes radioactive fluid for ultimate discharge m the plant. The radioactive waste handling aspects of this system are described in this section.
occasion, the Unit will generate liquid radioactive waste that cannot practicably be processed he liquid radwaste system. The station may process this waste outside the Unit in compliance h state and federal regulations, and in accordance with the Radiological Effluent Control gram outlined in the Administrative Section of the Technical Specifications (e.g., Unit 1 porator or shipped off site for processing).
radioactivity values provided in this section are the design basis values used for the design of liquid waste system. As such, they are considered historical and not subject to future updating.
information is retained to avoid loss of the original design bases. Actual liquid radioactive ase quantities can be found in the annual radioactive effluent release reports as submitted to NRC.
.1 DESIGN BASES
- 1. The design objectives of the liquid waste management systems are:
- a. To control the releases of radioactive materials within the limits set forth in 10 CFR 20 and to meet the numerical design objectives of Appendix I to 10 CFR 50.
- 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. Table 11.2-1 gives the daily input, in terms of average and peak flows, to the waste management systems. These values are 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 Tables 11.2-6 and 11.2-9.
11.2-1 Rev. 30
Table 11.2-2.
- 5. The radioactive liquid waste (LWS) and condensate demineralizer liquid waste systems (LWC) (removed from service) are designated nonnuclear safety (NNS).
Equipment is designed and fabricated in conformance with codes and standards identified in Regulatory Guide 1.143 (Section 1.8).
- 6. The foundation and walls of the radioactive liquid waste building and the condensate polishing facility building are seismically designed in conformance with Regulatory Guide 1.143.
- 7. A cost-benefit analysis for reducing cumulative dose to the population by using available technology has been performed in accordance with Regulatory Guide 1.110 and is included in Appendix 11A.
- 8. General Design Criterion 61 applies with regard to provisions for suitable shielding for radiation protection of personnel under normal and postulated accident conditions. Radiation protection criteria for the radioactive liquid waste system are given in Section 12.2.
- 9. Releases to the environment are monitored prior to discharge. Process and effluent radiological monitoring systems are described in Section 11.5.
- 10. The following design features are incorporated to reduce maintenance, equipment downtime, liquid leakage, and gaseous releases to the building atmosphere, or otherwise improve radwaste operations:
- a. Dished, sloped, and conical bottoms are used in vessels and tanks with a potential for high activity and suspended 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 preferable. Screwed connections in which threads provide the only seals are not used except for instrumentation connections where welded 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 nominal 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 performed in accordance with ANSI B31.1.
11.2-2 Rev. 30
- 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 liquids 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 provided to allow system operation and monitoring from one location.
- h. Components handling highly radioactive liquids are separated by shield walls to minimize exposure to operators and maintenance personnel.
- i. Plastic pipes are not used for radioactive service.
- 11. The design provisions to control radioactive releases due to overflows from all liquid tanks containing potentially radioactive materials are shown in Table 11.2-3.
.2 SYSTEM DESCRIPTION
.2.1 Radioactive Liquid Waste System (LWS) radioactive liquid waste system consists of two separate, but interconnected, portions: The h-level waste portion and the low-level waste portion.
h-Level Waste Portion o 26,000-gallon high-level waste drain tanks accept and store high-level radioactive liquid te from the sources identified in Table 11.2-1. Tank capacity allows time for recirculating and pling of one tank while the other tank is being filled from any of the above sources. Two te evaporator feed pumps service either tank. The tanks are cross-connected to the low-level te drain tanks (described below) at the discharge of the pumps.
o filters, located downstream of each waste evaporator feed pump, are available to pre-filter h-level waste drain tank contents. These filters may be operated in parallel, in series, or assed when recirculating or processing tank contents of a high-level waste drain tank.
sequent to this flowpath, the effluent from the high-level waste drain tank is processed in one 11.2-3 Rev. 30
waste evaporator, which was designed as an alternate path for processing high level waste has n demonstrated by analysis to not be required to operate in order to meet 10 CFR 50 endix I Release Criteria.
waste evaporator is designed with an external reboiler, a large liquid disengaging space, a or-liquid separator, and a tray section. These features combine to form a system with emely high separation factors for nonvolatile nuclides. A decontamination factor of greater 104 for nonvolatile nuclides is expected.
te evaporator bottoms are allowed to concentrate until either approximately 15 percent total olved and undissolved solids by weight have concentrated or an activity level to be rmined by the characteristics of the container used to ship the evaporator bottoms offsite is umulated. Evaporator bottoms are pumped to the radioactive solid waste system (Section 11.4) the waste bottoms holding tank.
uent from the high level radioactive waste demineralizer or distillate from the waste porator is collected in the waste test tanks. Samples of the liquid are analyzed for radioactivity chemistry parameters. Depending on analysis results, the liquid is discharged to the ulating water discharge tunnel (Section 10.4.5), or is capable of being recycled to the primary de water system.
amples indicate that the distillate is unacceptable for reuse or discharge, it is either passed ugh a second waste demineralizer and resampled, or sent back to the high-level waste drain s for reprocessing. The second waste demineralizer is a mixed bed of ion exchange resins in H and OH form. The resin is replaced when analysis of influent and effluent samples indicate the decontamination factor becomes unacceptable or when the radiation level exceeds a determined limit.
expected that liquid from the waste test tanks is totally discharged. For the purpose of luating the radiological impact on the environment, one hundred percent of the input flow is med to be discharged. Assurance that waste above activity limits is not inadvertently harged to the environment is provided through sampling of the waste test tank effluent and by radiation monitor in the discharge line. This monitor provides audible and visual alarms if vity levels in the effluent exceed limits. An air-operated valve in the discharge line is actuated erminate the release.
h batch is isotopically analyzed prior to release and the total activity discharged is recorded.
mposite samples are retained in accordance with the procedures outlined in Regulatory de 1.21. Detailed administrative records of all radioactive liquid releases are maintained.
imum expected decontamination factors are shown on Figure 11.2-3.
11.2-4 Rev. 30
o 4,000-gallon low-level waste drain tanks accept and store low level radioactive liquid waste m the reactor plant aerated drains system (Section 9.3.3), as identified in Table 11.2-1. These s and their respective pumps are arranged in the same manner as the high-level waste drain s.
bine building floor and equipment drains are normally discharged directly to the environment.
en there is high radioactivity in the discharge line, flow is diverted to the low-level waste em for further processing.
tem design provides for conveying the contents of the low-level waste drain tanks to the high-l waste drain tanks if the activity level of the liquid in the low-level tanks is greater than a determined level. Normally, the low-level waste is sampled, analyzed and discharged to the ulating water discharge tunnel. If the particulate concentration is above permit discharge ls, the contents of the low-level waste drain tanks will be recirculated through filter mblies using the low-level waste drain pump(s) until the particulate concentration levels are eptable. These filters are cartridge-type filters provided to remove particulate matter from the uent. Filter elements are changed when the radioactivity level at the filter surface or the sure drop across the filter exceeds a predetermined value.
urance that high-level radioactive waste is not inadvertently discharged to the environment is vided through the analysis of low-level waste drain tank liquid samples and the radiation nitor in the discharge line.
imum expected decontamination factors are shown on Figure 11.2-3.
.2.2 Condensate Demineralizer Liquid Waste System (LWC)
LWC has been removed from service and is no longer used.
s system has been isolated from the plant via locked closed valves, where possible. The system ng remains intact, such that if any leakage occurs across the locked, closed valves, it will be tained within the existing system boundary. The system is interconnected to the Unit 2 LWC at regenerant evaporator feed tanks, regenerant evaporator feed pump suction and discharge s, and at the system discharge downstream of the regenerant demineralizer (originally gned to provide added system availability, capacity, and redundancy). These interconnections isolated also, with locked closed valves, where possible. Electrical power to the system ponents is administratively controlled by plant procedures.
LWC was originally designed and installed as an evaporator system to receive and process ntially radioactive liquid waste from the condensate demineralizer-mixed bed system ction 10.4.6). Evaporator bottoms were designed to be pumped to Unit 2 for processing.
11.2-5 Rev. 30
culation Manual (ref. REMODCM CR# 95-7).
lysis has shown that LWC is not required to be operated to meet 10 CFR 50 Appendix I ase criteria.
the purpose of evaluating the radiological impact on the environment, it is assumed that 100 ent of the regenerant chemical waste is discharged to the environment.
.2.3 Other Systems Discharging Radioactive Liquid Waste boron recovery system (Section 9.3.5) receives degasified reactor coolant letdown and tor plant gaseous drains flow, as identified on Figure 11.2-3. For the purpose of evaluating the ological impact on the environment, one hundred percent of the input flow is assumed to be harged. Provisions made to protect against inadvertent discharge are similar to those for high-l radioactive liquid waste, as the boron recovery system distillate is discharged to the liquid te system upstream of the radiation monitor in the liquid waste system. Minimum expected ontamination factors are shown on Figure 11.2-3.
ing open-cycle blowdown, the steam generator blowdown (Section 10.4.8) intermittently harges directly to the circulating water discharge tunnel as part of the secondary water mistry control program. For the purpose of evaluating the radiological impact on the ironment, the following hypothetical cases have been developed.
design base case:
- 100 percent discharge to the circulating water discharge tunnel.
- Four steam generators blow down at 1 percent maximum steaming rate (MSR) for 14 days per year during hot standby.
- One steam generator blows down intermittently at 4 percent MSR, and the remaining three steam generators blow down at 1 percent MSR for 5 minutes per week per steam generator.
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.
actual radioactive releases will be reported annually.
11.2-6 Rev. 30
dels and assumptions contained in NUREG-0017 are used to calculate the expected oactivity concentrations in the liquid discharge. The concentration from each of the parent id waste streams following treatment are presented in Tables 11.2-4 and 11.2-7 for the ected and design conditions, respectively. Liquid releases to the environment are listed in les 11.2-5 and 11.2-6 for the expected nuclide concentrations prior and subsequent to dilution h the circulating water discharge system. A similar evaluation shown in Tables 11.2-8 and
-9 are for design nuclide concentrations prior and subsequent to dilution with the circulating er discharge system. The diluted release for the expected nuclide concentrations are further lyzed for the environmental impact as described in Appendix I of 10 CFR 50 and in endix 11A of this FSAR. Table 11.2-10 presents the design nuclide concentration releases to unrestricted area in terms of fraction of maximum permissible concentration (MPC) limits cribed in 10 CFR 20, Appendix B, Table II, Column 2. The results indicate that the sum of the tions of MPC values does not exceed the limits in 10 CFR 20.
.3.1 Radioactive Liquid Waste System Leak or Failure (Atmospheric Release) accident is defined as an unexpected and uncontrolled atmospheric release from the tulated rupture of a boron recovery tank. This tank is the highest potential atmospheric release rce term because of its large volume, the relatively high potential for activity in streams ing the tank, and its location in the yard area.
atmospheric release due to the postulated rupture of this tank is minimized by prior asification and demineralization of letdown, which removes essentially all xenon and krypton most of the iodine from the tank feed. The liquid released from the tank is held in a dike until ned or pumped, under administrative control, to the liquid radioactive waste system for her processing.
radiological consequences of a postulated radioactive liquid waste system failure resulting ospheric release is reported in Table 15.0-8 based on design release assumptions in le 11.2-11, boron recovery tank concentrations in Table 11.2-12 and the X/Q values in le 15.0-11. The resulting releases are listed in Table 11.2-13. The dose methodology is ussed in Appendix 15A.
rder to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric ertainty), each of the iodine and noble gas isotopes released in Table 11.2-13 were scaled ed on factors determined from the ratio of the primary activity concentrations (RCS centration from Table 15.0-10 and Design concentration from Table 11.1-2) to determine the ated doses for the SPU operating condition.
radiological consequences are consistent with the guidelines of the pre-1991 version of CFR 20, i.e., the whole body dose does not exceed 500 mRem to an individual at the nearest lusion area boundary, and is substantially below the guidelines of 10 CFR 100.
11.2-7 Rev. 30
s accident is defined as an unexpected and uncontrolled postulated rupture of the boron very tank. The boron recovery tanks are located in the yard area northeast of the containment cture (Figure 1.2-2). This area is provided with dikes to retain any liquid released from a tank ure. This analysis assumes a combined rupture of a boron recovery tank and leakage into the undwater.
cription of the analysis of this event is provided in Section 2.4.13.3.
concentration of radionuclides in Niantic Bay resulting from a liquid containing tank failure given in Table 11.2-14 based upon assumptions in Table 11.2-11.
rder to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric ertainty), each of the isotopes in Table 11.2-14 were scaled based on factors determined from ratio of the primary activity concentrations (RCS concentration from Table 15.0-10 and ign concentration from Table 11.1-2) to determine the updated concentrations for the SPU rating condition.
centrations in Niantic Bay are within the concentration of 10 CFR 20, Appendix B, Table II as existed prior to the 1991 revision to 10 CFR 20 (see Section 2.4.13).
.4 REFERENCE FOR SECTION 11.2 11.2-8 Rev. 30
NUREG-0017 Peak Flow Rate Average Flow Ra Liquid Waste Stream Category Source (gal/day) ( gal/day)
Misc. Liquid Waste System Clean Wastes High Level Waste Drain Tanks Containment Building Sump 300 40 Auxiliary Building Sump 1,500 200 Sample Fluids 200 35 Laboratory Waste 500 400 Misc. High Level Waste 800 660 Dirty Wastes Low Level Waste Drain Tanks Misc. Low Level Waste 650 40 Turbine Plant Floor Drains Turbine Plant Leakage 7,200 Boron Recovery System Shim Bleed Boron Recovery Tank 1,440 Equipment Drains Boron Recovery Tank 300 Secondary Waste System Steam Generator Blowdown Steam Generators 426,411 Regenerant Chemicals Regenerant Line 3,400 Detergent Wastes Laundry Facility 450 Page 1 of 1 Rev
TABLE 11.2-2 LIQUID WASTE MANAGEMENT SYSTEM DESIGN DATA dioactive Liquid Waste System ste Evaporator Feed Pumps Number 2 Capacity (gpm) 35 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction 316 SS ste Evaporator Reboiler Pump Number 1 Capacity (gpm) 4,000 Design pressure (psig) 600 Design temperature (°F) 350 Material of construction Alloy 20 ste Evaporator Bottoms Pump Number 1 Capacity (gpm) 15 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction Incoloy ste Distillate Pump Number 1 Capacity (gpm) 50 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction 316 SS ste Test Tank Pumps Number 2 Capacity (gpm) 150 Design pressure (psig) 350 Design temperature (°F) 250 Page 1 of 9 Rev. 30
Material of construction 316 SS w-Level Waste Drain Pumps Number 2 Capacity (gpm) 50 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction 316 SS ste Bottoms Coolant Pump Number 1 Capacity (gpm) 120 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction 316 SS ste Test Heating Pumps Number 2 Capacity (gpm) 60 Design pressure (psig) 350 Design temperature (°F) 250 Material of construction 316 SS ste Bottom Holding Tank Pump Number 1 Capacity (gpm) 50 Design pressure (psig) 250 Design temperature (°F) 170 Material of construction Alloy 20 ste Demineralizer Number 1 Capacity (ft3) 35 Design pressure (psig) 150 Design temperature (°F) 140 Material of construction 304 SS Page 2 of 9 Rev. 30
gh-Level Waste Demineralizer Number 1 Capacity (ft3) 35 Design pressure (psig) 140 Design temperature (°F) 200 Material of construction 304 SS ste Evaporator Reboiler Number 1 Total duty (Btu/hr) 26,200,000 Shell Side Tube Side Total fluid entering 29,900 1,930,000 Design pressure (psig) 180 100 Operating pressure inlet (psig) 100 25 Temperature in/out (°F) 338 / 338 251.2 / 265.1 Material of construction Carbon steel Incoloy 825 ste Evaporator Bottoms Cooler (3LWS-E2)
Total duty (Btu/hr) 637,500 Shell Side Tube Side Total fluid entering (lb/hr) 63,750 7,500 Temperature in/out (°F) 140 / 150 255 / 170 Design pressure (psig) 210 170 Operating pressure inlet (psig) 125 50 Material of construction Carbon steel Carpenter 20 ste Bottoms Coolant Preheater (3LWS-E3)
Total duty (kW) 90 Flow rate (gpm) 150 Design pressure (psig) 210 Temperature in/out (°F) 140 / 185 Material of construction Carbon steel ste Evaporator Condenser Total duty (Btu/hr) 22,717,947 Page 3 of 9 Rev. 30
Shell Side Tube Side Total liquid entering (lb/hr) 24,040 1,135,898 Temperature in/out (°F) 250 / 250 95 / 115 Design pressure (psig) 100 150 Material of construction 304 SS 304 SS ste Distillate Cooler Total duty (Btu/hr) 2,285,500 Shell Side Tube Side Total liquid entering (lb/hr) 17,500 114,275 Temperature in/out (°F 250 / 120 95 / 115 Design pressure (psig) 150 200 Material of construction 304 SS 304 SS ste Test Tank Heaters Number 2 Total duty (Btu/hr) 10,236 Operating flow (gpm) 60 Design pressure (psig) 50 Material of construction 304 SS ste Bottoms Holding Tank Heaters Number 2 Design temperature (°F) 200 Design pressure (psig) 10 Total load (kW) 18 Material of construction Incoloy 825 ste Evaporator Capacity (gpm) 35 Design pressure (psig) 100 and full vacuum Design temperature (°F) 350 Material of construction Top, 316 SS Bottom, Incoloy 825 ste Demineralizer Filter Page 4 of 9 Rev. 30
Design flow rate (gpm) 160 Design pressure (psig) 165 Design temperature (°F) 200 Material of construction Internals, 304 SS luent Filters Number 2 Design Flow rate (gpm) N/A Design pressure (psig) 150 Design temperature (°F) N/A Design pressure drop, clean (psi) 1 Material of construction Internals, 304 SS gh-Level Waste Recirc Filters Number 4 Design Flow rate (gpm) 75 Design pressure (psig) 150 Design temperature (°F) 200 Material of construction 304 SS gh Waste Demineralizer Filter Flow rate (gpm) 75 Design pressure (psig) 110 Design temperature (°F) 200 Material of construction Internal, 304 SS w-Level Waste Recirc Filters Number 2 Design flow rate (gpm) 60 Design pressure (psig) 150 Design Temperature (°F) 150 Design Pressure drop, clean (psi) 1 Material of construction Housing 304 L gh-Level Waste Drain Tanks Number 2 Page 5 of 9 Rev. 30
Capacity (gal) 26,000 Design pressure (psig) 25 Design temperature (°F) 200 Material of construction Shell 304 SS ste Distillate Tank Capacity (gal) 500 Design pressure (psig) 100 Design temperature (°F) 340 Material of construction Shell, 304 SS aste Test Tanks Number 2 Capacity (gal) 24,000 Design pressure (psig) atm and full liquid Design temperature (°F) 200 Material of construction Shell, 304 SS w-Level Waste Drain Tanks Number 2 Capacity (gal) 4,000 Design pressure (psig) 25 Design temperature (°F) 212 Material of construction Shell, 304 SS ste Bottoms Holding Tank Number 1 Capacity (gal) 3,000 Design pressure (psig) atm Design temperature (°F) 212 Material of construction Incoloy 825 ndensate Demineralizer Liquid Waste System (removed from service) generant Evaporator Feed Tanks (removed from service)
Number 2 Capacity (gal) 13,000 Page 6 of 9 Rev. 30
Design pressure (psig) atm Design temperature (°F) 212 Material of construction Fiberglass generant Distillate Tank (removed from service)
Number 1 Capacity (gal) 500 Design pressure (psig) 100 Design temperature (°F) 340 Material of construction 304 SS generant Evaporator Feed Pump (removed from service)
Capacity (gal) 50 Design pressure (psig) 150 Design temperature (°F) 150 Material of construction 316 SS generant Evaporator Bottoms Pump (removed from service)
Capacity (gal) 15 Design pressure (psig) 150 Design temperature (°F) 274 Material of construction A-20 SS generant Bottoms Coolant Pump (removed from service)
Capacity (gal) 120 Design pressure (psig) 150 Design temperature (°F) 185 Material of construction 316 SS generant Distillate Pump (removed from service)
Capacity (gal) 50 Design pressure (psig) 150 Design temperature (°F) 274 Material of construction 316 SS generant Evaporator Reboiler Pump (removed from service)
Capacity (gal) 4,000 Page 7 of 9 Rev. 30
Design pressure (psig) 600 Design temperature (°F) 350 Material of construction Alloy 20 generant Distillate Cooler (removed from service)
Shell Side Tube Side Total fluid entering (lb/hr) 17,500 114,275 Design pressure (psig) 150 150 Design temperature (°F) 274 274 Total duty (Btu/hr) 2,285,500 2,285,500 Material of construction 304SS 304SS generant Evaporator Reboiler (removed from service)
Shell Side Tube Side Total fluid entering (lb/hr) 23,700 1,930,000 Design pressure (psig) 180 100 Design temperature (°F) 380 350 Total duty (Btu/hr) 21,600,000 21,600,000 Material of construction Carbon steel Incoloy 825 generant Evaporator Bottoms Cooler (removed from service)
Shell Side Tube Side Total fluid entering (lb/hr) 63,750 7,500 Design pressure (psig) 150 150 Design temperature (°F) 274 274 Total duty (Btu/hr) 637,500 637,500 Material of construction Carbon steel Carpenter 20 generant Evaporator Condenser (removed from service)
Shell Side Tube Side Total fluid entering (lb/hr) 24,040 1,135,898 Design pressure (psig) 100 150 Design temperature (°F) 350 350 Total duty (Btu/hr) 22,717,947 22,717,947 Material of construction 304 SS 304 SS Page 8 of 9 Rev. 30
generant Bottoms Coolant Reheater (removed from service)
Design flow (gpm) 150 Duty (kW) 90 Design pressure (psig) 175 Design temperature (°F) 185 Material of construction SA 106 generant Demineralizer (removed from service)
Design flow rate (gpm) 50 Design pressure (psig) 150 Design temperature (°F) 250 Material of construction 304 SS generant Evaporator (removed from service)
Capacity (gpm) 50 Design pressure (psig) 100 and full vacuum Design temperature (°F) 300 Material of construction Top, 316 SS Bottom, Incoloy 825 generant Demineralizer Filter (removed from service)
Flow rate (gpm) 50 Design pressure (psig) 150 Design temperature (°F) 120 Material of construction 304 SS Page 9 of 9 Rev. 30
Level Monitoring and Monitoring or Tanks Mark No. Alarms Alarm Location (1) Overflow Provisions Processing of Overflow Containment Drains Transfer Tank 3DGS-TK1 Indicator MCB Overflows to containment sump In RLWS Primary Drain Transfer Tank 3DGS-TK2 Indicator MCB Overflows to auxiliary building sump In RLWS High Boron Recovery Tanks 3BRS-TK1A/1B Indicator BRP Overflows to second tank then to diked In RLWS if it overflows to dike High area Boron Test Tanks 3BRS-TK2A/2B Indicator BRP Overflows to the waste disposal building In RLWS if it overflows to dike High sump Boron Distillate Tank 3BRS-TK3 Indicator BRP Closed tank None High Cation Regeneration Tank 3CND-TK Indicator Local Located in diked area. Overflows to waste Contents of sumps are discharge neutralization sump to the recirculating water dischar tunnel or, if radioactive, processe in the condensate demineralizer liquid waste system Anion Regeneration Tank 3CND-TK2 Indicator Local Located in diked area. Overflows to waste Contents of sumps are discharge neutralization sump to the recirculating water dischar tunnel or, if radioactive, processe in the condensate demineralizer liquid waste system Resin Mix and Storage Tank 3CND-TK3 Indicator Local Located in diked area. Overflows to waste Contents of sumps are discharge neutralization sump to the recirculating water dischar tunnel or, if radioactive, processe in the condensate demineralizer liquid waste system Page 1 of 3 Rev
Level Monitoring and Monitoring or Tanks Mark No. Alarms Alarm Location (1) Overflow Provisions Processing of Overflow Recovered Caustic Tank 3CND-TK8 Indicator CD Located in diked area. Overflows to waste Contents of sumps are discharge High neutralization sump to the recirculating water dischar tunnel or, if radioactive, processe in the condensate demineralizer liquid waste system Recovered Water Tank 3CND-TK9 Indicator CD Located in diked area. Overflows to waste Contents of sumps are discharge High neutralization sump to the recirculating water dischar tunnel or, if radioactive, processe in the condensate demineralizer liquid waste system Regenerant Evaporator Feed Tanks 3LWC-TK1A/B Indicator LWC Located in diked area. Overflows to (removed from service) High contaminated floor drains High-Level Waste Drain Tanks 3LWS-TK1A/1B Indicator LWP Overflows to the waste disposal building In RLWS High sump Low-Level Waste Drain Tanks 3LWS-TK4A/4B Indicator LWP Overflows to the waste disposal building In RLWS High sump Waste Distillate Tank 3LWS-TK2 Indicator LWP Closed tank None High Regenerant Distillate Tank 3LWC-TK3 Indicator LWC Closed tank None (removed from service) High Waste Test Tanks 3LWS-TK3A/3B Indicator LWP Overflows to the waste disposal building In RLWS High sump Spent Resin Dewatering Tank 3WSS-TK1 Indicator SWP To solid waste building sump In RLWS High Page 2 of 3 Rev
Level Monitoring and Monitoring or Tanks Mark No. Alarms Alarm Location (1) Overflow Provisions Processing of Overflow Spent Resin Hold Tank 3WSS-TK3 Indicator SWP Overflows to spent resin dewatering tank In RLWS Waste Bottoms Holding Tank 3LWS-TK5 Indicator SWP Overflows to the waste disposal building In RLWS High sump Condensate Surge Tank 3CNS-T2 Indicator MCB Overflows to turbine building floor drains Contents of sump are routed to th High sump radioactive liquid waste system NOTE (1) Location Symbols BRP = Boron Recovery Panel MCB = Main Control Board LWP = Radioactive Waste Panel Page 3 of 3 Rev
(CI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)
SECONDARY (2) TURB BORON MISC. SYSTEM BLDG TOTAL ADJUSTED (3)
RECOVERY WASTES WASTES DRAINS LWS TOTAL LWS DETERGENT (4)
ISOTOPE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) WASTES (Ci/ml)
CORROSION AND ACTIVATION PRODUCTS Cr-51 1.66E-08 (5) 7.05E-07 6.68E-09 0.00E+00 8.77E-09 9.17E-09 0.00E+00 Mn-54 4.16E-09 1.21E-07 1.75E-09 0.00E+00 2.10E-09 2.19E-09 1.61E-06 Fe-55 2.91E-08 6.15E-07 7.08E-09 0.00E+00 8.97E-09 9.36E-09 0.00E+00 F-59 1.25E-08 3.73E-07 4.71E-09 0.00E+00 5.82E-09 6.08E-09 0.00E+00 Co-58 2.16E-07 6.06E-06 6.54E-08 0.00E+00 8.37E-08 8.74E-08 6.43E-06 Co-60 3.74E-08 7.68E-07 8.01E-09 0.00E+00 1.04E-08 1.08E-08 1.40E-05 Zr-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.25E-06 Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.21E-06 Np-239 0.00E+00 3.68E-07 3.68E-09 0.00E+00 4.75E-09 4.96E-09 0.00E+00 FISSION PRODUCTS Br-83 0.00E+00 3.89E-07 7.00E-09 0.00E+00 8.05E-09 8.41E-09 0.00E+00 Br-84 0.00E+00 1.58E-08 1.83E-09 0.00E+00 1.84E-09 1.94E-09 0.00E+00 Br-85 0.00E+00 0.00E+00 3.36E-11 0.00E+00 3.29E-11 3.29E-11 0.00E+00 Rb-86 4.16E-09 2.79E-07 2.69E-10 0.00E+00 1.17E-09 1.22E-09 0.00E+00 Rb-88 0.00E+00 2.47E-07 9.58E-08 0.00E+00 9.43E-08 9.85E-08 0.00E+00 Sr-89 4.16E-09 1.31E-07 1.60E-09 0.00E+00 1.99E-09 2.07E-09 0.00E+00 Page 1 of 4 Rev
(CI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)
SECONDARY (2) TURB BORON MISC. SYSTEM BLDG TOTAL ADJUSTED (3)
RECOVERY WASTES WASTES DRAINS LWS TOTAL LWS DETERGENT (4)
ISOTOPE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) WASTES (Ci/ml)
S-90 0.00E+00 5.26E-09 3.36E-11 0.00E+00 4.93E-11 4.93E-11 0.00E+00 S-91 0.00E+00 1.42E-07 1.43E-09 0.00E+00 1.84E-09 1.94E-09 0.00E+00 Y-90 0.00E+00 0.00E+00 1.68E-11 0.00E+00 1.64E-11 1.64E-11 0.00E+00 Y-91m 0.00E+00 9.47E-08 1.16E-09 0.00E+00 1.43E-09 1.50E-09 0.00E+00 Y-91 0.00E+00 2.63E-08 2.52E-10 0.00E+00 3.29E-10 3.29E-10 0.00E+00 Y-93 0.00E+00 5.26E-09 6.73E-11 0.00E+00 9.86E-11 9.86E-11 0.00E+00 Zr-95 0.00E+00 2.10E-08 3.20E-10 0.00E+00 3.94E-10 4.11E-10 0.00E+00 Nb-95 0.00E+00 2.10E-08 3.53E-10 0.00E+00 4.11E-10 4.27E-10 0.00E+00 Mo-99 4.16E-08 2.62E-05 2.45E-07 2.01E-09 3.21E-07 3.35E-07 0.00E+00 Tc-99m 4.16E-08 2.06E-05 2.44E-07 2.01E-09 3.02E-07 3.16E-07 0.00E+00 Ru-103 0.00E+00 1.58E-08 1.51E-10 0.00E+00 2.14E-10 2.14E-10 2.25E-07 Ru-106 0.00E+00 5.26E-09 3.36E-11 0.00E+00 4.93E-11 4.93E-11 3.86E-06 Rh-103m 0.00E+00 1.58E-08 2.69E-10 0.00E+00 3.12E-10 3.29E-10 0.00E+00 Rh-106 0.00E+00 5.26E-09 6.73E-11 0.00E+00 8.21E-11 8.21E-11 0.00E+00 Ag-110m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7.07E-07 Te-125m 0.00E+00 1.05E-08 8.41E-11 0.00E+00 1.15E-10 1.15E-10 0.00E+00 Te-127m 4.16E-09 1.05E-07 8.41E-10 0.00E+00 1.17E-09 1.22E-09 0.00E+00 Te-127 4.16E-09 2.37E-07 2.44E-09 0.00E+00 3.14E-09 3.29E-09 0.00E+00 Page 2 of 4 Rev
(CI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)
SECONDARY (2) TURB BORON MISC. SYSTEM BLDG TOTAL ADJUSTED (3)
RECOVERY WASTES WASTES DRAINS LWS TOTAL LWS DETERGENT (4)
ISOTOPE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) WASTES (Ci/ml)
Te-129m 1.25E-08 5.21E-07 4.56E-09 0.00E+00 6.14E-09 6.41E-09 0.00E+00 Te-129 8.31E-09 3.58E-07 7.34E-09 0.00E+00 8.31E-09 8.67E-09 0.00E+00 Te-131m 0.00E+00 7.10E-07 6.31E-09 0.00E+00 8.38E-09 8.76E-09 0.00E+00 Te-131 0.00E+00 1.31E-07 2.52E-09 0.00E+00 2.89E-09 3.02E-09 0.00E+00 Te-132 1.66E-08 8.55E-06 6.12E-08 1.00E-09 8.66E-08 9.04E-08 0.00E+00 I-130 0.00E+00 5.00E-07 4.07E-09 0.00E+00 5.54E-09 5.78E-09 0.00E+00 I-131 6.54E-06 9.30E-05 1.11E-06 4.92E-08 1.40E-06 1.46E-06 9.64E-08 I-132 1.66E-08 1.46E-05 1.97E-07 4.02E-09 2.38E-07 2.49E-07 0.00E+00 I-133 9.97E-08 1.02E-04 7.93E-07 5.83E-08 1.10E-06 1.14E-06 0.00E+00 I-134 0.00E+00 8.78E-07 4.40E-08 0.00E+00 4.57E-08 4.77E-08 0.00E+00 I 135 0.00E+00 3.53E-05 3.46E-07 1.81E-08 4.48E-07 4.69E-07 0.00E+00 Cs-134 4.82E-06 1.05E-04 9.40E-08 1.00E-09 4.40E-07 4.60E-07 2.09E-05 Cs-136 5.82E-07 3.83E-05 4.10E-08 0.00E+00 1.62E-07 1.69E-07 0.00E+00 Cs-137 3.59E-06 7.63E-05 6.84E-08 1.00E-09 3.19E-07 3.34E-07 3.86E-05 Ba-137m 3.35E-06 7.14E-05 1.41E-07 0.00E+00 3.74E-07 3.90E-07 0.00E+00 Ba-140 0.00E+00 7.89E-08 6.73E-10 0.00E+00 9.04E-10 9.36E-10 0.00E+00 La-140 0.00E+00 6.31E-08 5.38E-10 0.00E+00 7.23E-10 7.56E-10 0.00E+00 Ce-141 0.00E+00 2.63E-08 3.03E-10 0.00E+00 3.78E-10 3.94E-10 0.00E+00 Page 3 of 4 Rev
(CI/ML) (1) FOLLOWING TREATMENT (HISTORICAL)
SECONDARY (2) TURB BORON MISC. SYSTEM BLDG TOTAL ADJUSTED (3)
RECOVERY WASTES WASTES DRAINS LWS TOTAL LWS DETERGENT (4)
ISOTOPE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) WASTES (Ci/ml)
Ce-143 0.00E+00 1.05E-08 6.73E-11 0.00E+00 9.86E-11 9.86E-11 0.00E+00 C-144 0.00E+00 1.05E-08 1.68E-08 0.00E+00 2.14E-10 2.14E-10 8.36E-06 Pr-143 0.00E+00 1.58E-08 1.35E-10 0.00E+00 1.97E-10 1.97E-10 0.00E+00 Pr-144 0.00E+00 1.05E-08 3.20E-10 0.00E+00 3.61E-10 3.78E-10 0.00E+00 NOTES:
(1) Refer to Figure 11.2-3 for respective steam processing.
(2) Includes steam generator blowdown during open cylce 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 included 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 has been retained to preserve original license basis.
Page 4 of 4 Rev
BLE 11.2-5 EXPECTED ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO LUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM AND PRIOR TO CLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Concentration Radioactivity Released Isotope (Ci/gm) (Ci/yr)
RROSION AND ACTIVATION PRODUCTS 51 6.85E-08 (2) 5.34E-03
-54 1.64E-08 1.28E-03
-55 7.01E-08 5.46E-03
-59 4.54E-08 3.54E-03
-58 6.53E-07 5.09E-02
-60 8.10E-08 6.31E-03 95 0.00E+00 0.00E+00
-95 0.00E+00 0.00E+00
-239 3.71E-08 2.89E-03 SSION PRODUCTS 83 6.29E-08 4.90E-03 84 1.44E-08 1.12E-03 85 2.57E-10 2.00E-05
-86 9.11E-09 7.10E-04
-88 7.36E-07 5.74E-02 89 1.55E-08 1.21E-03 90 3.85E-10 3.00E-05 91 1.44E-08 1.12E-03 90 1.28E-10 1.00E-05 91m 1.12E-08 8.70E-04 91 2.57E-09 2.00E-04 93 7.70E-10 6.00E-05 95 3.08E-09 2.40E-04
-95 3.21E-09 2.50E-04
-99 2.51E-06 1.95E-01 Page 1 of 3 Rev. 30
CLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Concentration Radioactivity Released Isotope (Ci/gm) (Ci/yr) 99m 2.36E-06 1.84E-01
-103 1.67E-09 1.30E-04
-106 3.85E-10 3.00E-05
-103m 2.44E-09 1.90E-04
-106 6.42E-10 5.00E-05
-110m 0.00E+00 0.00E+00 125m 8.98E-10 7.00E-05 127m 9.11E-09 7.10E-04 127 2.45E-08 1.91E-03 129m 4.80E-08 3.74E-03 129 6.49E-08 5.06E-03 131m 6.54E-08 5.10E-03 131 2.26E-08 1.76E-03 132 6.76E-07 5.27E-02 30 4.32E-08 3.37E-03 31 1.09E-05 8.50E-01 32 1.86E-06 1.45E-01 33 8.56E-06 6.67E-01 34 3.57E-07 2.78E-02 35 3.50E-06 2.73E-01
-134 3.44E-06 2.68E-01
-136 1.26E-06 9.86E-02
-137 2.49E-06 1.94E-01
-137m 2.92E-06 2.27E-01
-140 7.06E-09 5.50E-04
-140 5.65E-09 4.40E-04
-141 2.95E-09 2.30E-04
-143 7.70E-10 6.00E-05 Page 2 of 3 Rev. 30
CLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Concentration Radioactivity Released Isotope (Ci/gm) (Ci/yr)
-144 1.67E-09 1.30E-04 143 1.54E-09 1.20E-04 144 2.82E-09 2.20E-04 tal (excluding H-3) 4.30E-05 3.35E+00 3 9.24E-03 7.20E+02 tal (including H-3) 9.28E-03 7.23E+02 TES:
Detergent wastes not included in this table 6.85E-08 = 6.85 x 10-8 otal annual liquid waste release volume (excluding detergent wastes) is 7.79E+10 ml torical, not subject to future updating. This table has been retained to preserve original license s.
Page 3 of 3 Rev. 30
ABLE 11.2-6 EXPECTED ANNUAL RADIOACTIVE LIQUID RELEASES AFTER LUTION IN THE CIRCULATING WATER DISHARGE SYSTEM AND INCLUSION OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Released Radioactivity Released Isotope (Ci/gm) (Ci/yr)
RROSION AND ACTIVATION PRODUCTS 51 3.41E-12 (2) 5.60E-03
-54 1.40E-12 2.30E-03
-55 3.47E-12 5.70E-03
-59 2.25E-12 3.70E-03
-58 3.47E-11 5.70E-02
-60 9.12E-12 1.50E-02 95 8.51E-13 1.40E-03
-95 1.22E-12 2.00E-03
-239 1.82E-12 3.00E-03 SSION PRODUCTS 83 3.10E-12 5.10E-03 84 7.30E-13 1.20E-03 85 1.22E-14 2.00E-05
-86 4.50E-13 7.40E-04
-88 3.65E-11 6.00E-02 89 7.91E-13 1.30E-03 90 1.82E-14 3.00E-05 91 7.30E-13 1.20E-03 90 6.08E-15 1.00E-05 91m 5.53E-13 9.10E-04 91 1.22E-13 2.00E-04 93 3.65E-14 6.00E-05 95 1.52E-13 2.50E-04
-95 1.58E-13 2.60E-04
-99 1.22E-10 2.00E-01 Page 1 of 3 Rev. 30
OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Released Radioactivity Released Isotope (Ci/gm) (Ci/yr) 99m 1.16E-10 1.90E-01
-103 1.64E-13 2.70E-04
-106 1.46E-12 2.40E-03
-103m 1.22E-13 2.00E-04
-106 3.04E-14 5.00E-05
-110m 2.68E-13 4.40E-04 125m 4.26E-14 7.00E-05 127m 4.50E-13 7.40E-04 127 1.22E-12 2.00E-03 129m 2.37E-12 3.90E-03 129 3.22E-12 5.30E-03 131m 3.22E-12 5.30E-03 131 1.09E-12 1.80E-03 132 3.34E-11 5.50E-02 30 2.13E-12 3.50E-03 31 5.41E-10 8.90E-01 32 9.12E-11 1.50E-01 33 4.26E-10 7.00E-01 34 1.76E-11 2.90E-02 35 1.76E-10 2.90E-01
-134 1.76E-10 2.90E-01
-136 6.08E-11 1.00E-01
-137 1.40E-10 2.30E-01
-137m 1.46E-10 2.40E-01
-140 3.47E-13 5.70E-04
-140 2.80E-13 4.60E-04
-141 1.46E-13 2.40E-04
-143 3.65E-14 6.00E-05 Page 2 of 3 Rev. 30
OF ANTICIPATED OPERATIONAL OCCURRENCES (1) (HISTORICAL)
Radioactivity Released Radioactivity Released Isotope (Ci/gm) (Ci/yr)
-144 3.22E-12 5.30E-03 143 7.30E-14 1.20E-04 144 1.40E-13 2.30E-04 tal (excluding H-3) 2.19E-09 3.60E+00 3 4.38E-07 7.20E+02 tal (including H-3) 4.40E-07 7.24E+02
- 1. Detergent wastes are included in this table 3.41E-12 = 3.41 x 1012 Includes source term adjustment for anticipated operational occurrences 1.50E-01 Ci/yr Total annual liquid waste release volume is 7.86E+10 ml/yr MP3 average dilution flow of 1840 ft3/sec or 1.64E+15 ml/yr torical, not subject to future updating. This table has been retained to preserve original license s.
Page 3 of 3 Rev. 30
Reactor Misc. Reactor Reactor Plant Steam Cont.Bldg. Aux. Bldg. Laboratory Plant Low-Level Misc. High- Coolant Gaseous Regenerant Turbine Bldg. Generator Isotope Sump Sump Waste Samples Waste Level Waste Bleed Drains Chemicals Drains Blowdown (
Cr-51 1.7E-07 4 1.7E-08 3.3E-10 1.7E-07 8.7E-09 1.7E-09 8.8E-09 8.8E-08 6.9E-11 1.8E-09 7.2E-07 Mn-54 3.2E-08 3.2E-09 6.5E-11 3.2E-08 3.0E-06 3.2E-10 3.0E-09 3.0E-08 1.2E-11 3.0E-10 1.2E-07 Fe-55 1.7E-07 1.7E-08 3.4E-10 1.7E-07 1.7E-05 1.7E-09 1.7E-08 1.7E-07 6.2E-11 1.5E-09 6.2E-07 Fe-59 9.8E-08 9.8E-09 2.0E-10 9.8E-08 6.3E-06 9.8E-10 6.4E-09 6.4E-08 3.9E-11 1.0E-09 4.0E-07 Co-58 1.6E-06 1.6E-07 3.2E-09 1.6E-06 1.2E-04 1.6E-08 1.2E-07 1.2E-06 6.1E-10 1.5E-08 6.2E-06 Co-60 2.1E-07 2.1E-08 4.2E-10 2.1E-07 2.1E-05 2.1E-09 2.1E-08 2.1E-07 7.6E-11 1.9E-09 7.6E-07 Sr-89 3.9E-07 3.9E-08 7.8E-10 3.9E-07 2.6E-05 3.9E-09 2.7E-08 2.7E-07 1.5E-10 3.9E-09 1.6E-06 Sr-90 1.7E-08 1.7E-09 3.4E-11 1.7E-08 1.7E-06 1.7E-10 1.7E-09 1.7E-08 6.2E-12 1.6E-10 6.2E-08 Sr-91 3.8E-09 3.8E-10 7.6E-12 3.8E-09 1.4E-07 3.8E-11 4.8E-13 4.8E-12 7.9E-12 1.5E-09 6.1E-07 Y-90 1.8E-08 1.8E-09 3.6E-11 1.8E-08 1.7E-06 1.8E-10 1.7E-09 1.7E-08 7.1E-12 1.9E-10 7.5E-08 Y-91M 2.6E-09 2.6E-10 5.2E-12 2.6E-09 9.5E-08 2.6E-11 3.2E-13 3.2E-12 5.4E-12 9.3E-10 3.7E-07 Y-91 6.4E-08 6.4E-09 1.3E-10 6.4E-08 4.5E-06 6.4E-10 4.6E-09 4.6E-08 2.5E-11 6.3E-10 2.5E-07 Y-93 7.4E-10 7.4E-11 1.5E-12 7.4E-10 2.6E-08 7.4E-12 1.3E-13 1.3E-12 1.5E-12 2.6E-10 1.1E-07 Zr-95 6.6E-08 6.6E-09 1.3E-10 6.6E-08 4.8E-06 6.6E-10 4.9E-09 4.9E-08 2.5E-11 6.5E-10 2.6E-07 Nb-95 7.3E-08 7.3E-09 1.5E-10 7.3E-08 6.3E-06 7.3E-10 6.4E-09 6.4E-08 2.7E-11 6.7E-10 2.7E-07 Mo-99 8.0E-05 8.0E-06 1.6E-07 8.0E-05 1.7E-03 8.0E-07 8.5E-07 8.5E-06 8.0E-08 3.0E-06 1.2E-03 Tc-99M 7.6E-05 7.6E-06 1.5E-07 7.6E-05 1.6E-03 7.6E-07 8.2E-07 8.2E-06 7.5E-08 2.0E-06 7.8E-04 Ru-103 3.0E-08 3.0E-09 6.0E-11 3.0E-08 1.8E-06 3.0E-10 1.9E-09 1.9E-08 1.2E-11 3.1E-10 1.2E-07 Page 1 of 4 Rev
Reactor Misc. Reactor Reactor Plant Steam Cont.Bldg. Aux. Bldg. Laboratory Plant Low-Level Misc. High- Coolant Gaseous Regenerant Turbine Bldg. Generator Isotope Sump Sump Waste Samples Waste Level Waste Bleed Drains Chemicals Drains Blowdown (
Ru-106 3.2E-09 3.2E-10 6.4E-12 3.2E-09 3.0E-07 3.2E-11 3.0E-10 3.0E-09 1.2E-12 3.0E-11 1.2E-08 Rh-103M 3.0E-08 3.0E-09 6.0E-11 3.0E-08 1.8E-06 3.0E-10 1.9E-09 1.9E-08 1.2E-11 2.9E-10 1.2E-07 Rh-106 3.2E-09 3.2E-10 6.4E-12 3.2E-09 3.0E-07 3.2E-11 3.0E-10 3.0E-09 1.2E-12 2.7E-11 1.1E-08 Te-125M 8.2E-09 8.2E-10 1.6E-11 8.2E-09 5.8E-07 8.3E-11 5.9E-10 5.9E-09 3.2E-12 8.2E-11 3.3E-08 Te-127M 2.0E-07 2.0E-08 4.0E-10 2.0E-07 1.7E-05 2.0E-09 1.7E-08 1.7E-07 7.6E-11 1.9E-09 7.7E-07 Te-127 1.9E-07 1.9E-08 3.9E-10 1.9E-07 1.6E-05 1.9E-09 1.6E-08 1.6E-07 7.0E-11 1.1E-09 4.5E-07 Te-129M 3.3E-06 3.3E-07 6.7E-09 3.3E-06 1.5E-04 3.3E-08 1.9E-07 1.9E-06 1.4E-09 3.5E-08 1.4E-05 Te-129 2.1E-06 2.1E-07 4.3E-09 2.1E-06 1.2E-04 2.1E-08 1.2E-07 1.2E-06 8.7E-10 2.2E-08 8.7E-06 Te-131M 2.3E-07 2.3E-08 4.5E-10 2.3E-07 5.3E-06 2.3E-09 9.7E-10 9.7E-09 3.5E-10 2.0E-08 8.0E-06 Te-131 4.1E-08 4.1E-09 8.3E-11 4.1E-08 9.9E-07 4.1E-10 1.8E-10 1.8E-09 6.3E-11 5.5E-09 2.2E-06 Te-132 7.5E-06 7.5E-07 1.5E-08 7.5E-06 1.6E-04 7.5E-08 9.0E-08 9.0E-07 6.8E-09 2.4E-07 9.6E-05 Ba-137M 1.6E-04 1.6E-05 3.1E-07 1.6E-04 1.6E-02 1.6E-06 3.9E-04 7.8E-04 1.0E-07 1.4E-06 5.7E-04 Ba-140 3.0E-07 3.0E-08 5.9E-10 3.0E-07 1.0E-05 3.0E-09 9.6E-09 9.6E-08 1.4E-10 4.0E-09 1.6E-06 La-140 2.9E-07 2.9E-08 5.8E-10 2.9E-07 1.0E-05 2.9E-09 1.1E-08 1.1E-07 1.0E-10 1.5E-09 5.9E-07 Ce-141 5.9E-08 5.9E-09 1.2E-10 5.9E-08 3.4E-06 5.9E-10 3.4E-09 3.4E-08 2.4E-11 6.3E-10 2.5E-07 Ce-143 5.6E-09 5.6E-10 1.1E-11 5.6E-09 1.3E-07 5.6E-11 2.8E-11 2.8E-10 8.3E-12 4.5E-10 1.8E-07 Ce-144 4.9E-08 4.9E-09 9.8E-11 4.9E-08 4.5E-06 4.9E-10 4.5E-09 4.5E-08 1.8E-11 4.5E-10 1.8E-07 Pr-143 5.0E-08 5.0E-09 1.0E-10 5.0E-08 1.8E-06 5.0E-10 1.7E-09 1.7E-08 2.3E-11 6.1E-10 2.5E-07 Pr-144 4.9E-08 4.9E-09 9.7E-11 4.9E-08 4.5E-06 4.9E-10 4.5E-09 4.5E-08 1.8E-11 4.5E-10 1.8E-07 Page 2 of 4 Rev
Reactor Misc. Reactor Reactor Plant Steam Cont.Bldg. Aux. Bldg. Laboratory Plant Low-Level Misc. High- Coolant Gaseous Regenerant Turbine Bldg. Generator Isotope Sump Sump Waste Samples Waste Level Waste Bleed Drains Chemicals Drains Blowdown (
Np-239 7.8E-08 7.8E-09 1.6E-10 7.8E-08 1.7E-06 7.8E-10 7.2E-10 7.2E-09 8.5E-11 3.4E-09 1.4E-06 Br-83 5.5E-08 5.5E-09 1.1E-10 5.5E-08 1.6E-06 5.5E-10 0.0 0.0 1.7E-10 1.5E-07 1.5E-05 Br-84 3.1E-13 3.1E-14 0.0 3.1E-13 1.6E-07 3.1E-15 0.0 0.0 4.7E-14 3.1E-08 3.1E-06 Br-85 0.0 0.0 0.0 0.0 2.0E-09 0.0 0.0 0.0 0.0 5.1E-10 5.2E-08 I-130 2.1E-07 2.1E-08 4.2E-10 2.1E-07 6.7E-07 2.1E-09 9.0E-11 9.0E-10 3.1E-10 1.7E-08 1.7E-06 I-131 1.4E-03 1.4E-04 2.9E-06 1.4E-03 3.8E-03 1.4E-05 3.3E-05 3.3E-04 6.0E-07 6.9E-06 6.9E-04 I-132 1.2E-05 1.2E-06 2.4E-08 1.2E-05 1.7E-04 1.2E-07 9.3E-08 9.3E-07 2.3E-08 2.0E-06 2.0E-04 I-133 2.5E-04 2.5E-05 5.0E-07 2.5E-04 6.5E-04 2.5E-06 5.2E-07 5.2E-06 3.3E-07 1.0E-05 1.0E-03 I-134 1.1E-09 1.1E-10 2.2E-12 1.1E-09 3.8E-06 1.1E-11 0.0 0.0 2.2E-11 6.3E-07 6.3E-05 I-135 2.2E-05 2.2E-06 4.3E-08 2.2E-05 1.1E-04 2.2E-07 2.9E-10 2.9E-09 3.8E-08 4.9E-06 4.9E-04 Rb-86 2.0E-08 2.0E-09 4.1E-11 2.0E-08 8.6E-07 2.0E-10 2.1E-08 4.3E-08 1.6E-11 4.4E-10 1.7E-07 Rb-88 0.0 0.0 0.0 0.0 7.5E-06 0.0 0.0 0.0 0.0 4.6E-07 1.8E-04 Cs-134 3.3E-05 3.3E-06 6.6E-08 3.3E-05 3.2E-03 3.3E-07 8.0E-05 1.6E-04 2.2E-08 5.4E-07 2.2E-04 Cs-136 1.1E-05 1.1E-06 2.3E-08 1.1E-05 3.9E-04 1.1E-07 9.4E-06 1.9E-05 1.0E-08 2.7E-07 1.1E-04 Cs-137 1.7E-04 1.7E-05 3.3E-07 1.7E-04 1.7E-02 1.7E-06 4.1E-04 8.3E-04 1.1E-07 2.7E-06 1.1E-03 Page 3 of 4 Rev
NOTES:
(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 10 CFR 20 Appendix B, Table II, Column 2 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 Page 4 of 4 Rev
BLE 11.2-8 DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES PRIOR TO DDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM Isotope Activity Released (Ci/gm) Activity Released (Ci/yr) 51 5.5E-07 (2) 1.9E-02
-54 9.3E-08 3.2E-03
-55 4.8E-07 1.6E-02
-59 3.0E-07 1.0E-02
-58 4.7E-06 1.6E-01
-60 5.9E-07 2.0E-02 89 1.2E-06 4.1E-02 90 4.8E-08 1.6E-03 91 4.5E-07 1.5E-02 90 5.8E-08 2.0E-03 91M 2.8E-07 9.4E-03 91 1.9E-07 6.5E-03 93 7.9E-08 2.7E-03 95 2.0E-07 6.7E-03
-95 2.1E-07 7.0E-03
-99 9.0E-04 3.0E+01 99M 5.9E-04 2.0E+01
-103 9.5E-08 3.2E-03
-106 9.2E-09 3.1E-04
-103M 8.9E-08 3.0E-03
-106 8.7E-09 3.0E-04 125M 2.5E-08 8.5E-04 127M 5.9E-07 2.0E-02 127 3.6E-07 1.2E-02 129M 1.1E-05 3.7E-01 129 6.7E-06 2.3E-01 131M 6.0E-06 2.0E-01 Page 1 of 3 Rev. 30
DDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM Isotope Activity Released (Ci/gm) Activity Released (Ci/yr) 131 1.6E-06 5.5E-02 132 7.2E-05 2.4E+00
-137M 4.4E-04 1.5E+01
-140 1.2E-06 4.0E-02
-140 4.6E-07 1.5E-02
-141 1.9E-07 6.5E-03
-143 1.3E-07 4.5E-03
-144 1.4E-07 4.7E-03 143 1.9E-07 6.3E-03 144 1.4E-07 4.7E-03
-239 1.0E-06 3.4E-02 83 1.1E-05 3.8E-01 84 2.3E-06 7.9E-02 85 3.9E-08 1.3E-03 30 1.3E-06 4.3E-02 31 5.2E-04 1.8E+01 32 1.5E-04 5.0E+00 33 7.7E-04 2.6E+01 34 4.7E-05 1.6E+00 35 3.6E-04 1.2E+01
-86 1.3E-07 4.4E-03
-88 1.4E-04 4.6E+00
-134 1.7E-04 5.6E+00
-136 8.2E-05 2.8E+00
-137 8.3E-04 2.8E+01 3 7.5E-02 2.5E+03 Page 2 of 3 Rev. 30
TES:
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 10 CFR 20 Appendix B, Table II, Column 2 effluent limits.
5.5E-07 = 5.5 x 10-7.
al liquid waste release is 3.4E+10 gm/yr.
al release (excluding tritium) is 1.7+02 Ci/yr.
al release concentration (excluding tritium) is 5.1E-03 mCi/gm.
Page 3 of 3 Rev. 30
TABLE 11.2-9 DESIGN (1) ANNUAL RADIOACTIVE LIQUID RELEASES LLOWING ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM Isotope Activity Released (Ci/gm) Activity Released (Ci/yr) 51 6.3E-12 (2) 1.9E-02
-54 1.1E-12 3.2E-03
-55 5.5E-12 1.6E-02
-59 3.5E-12 1.0E-02
-58 5.4E-11 1.6E-01
-60 6.8E-12 2.0E-02 89 1.4E-11 4.1E-02 90 5.5E-13 1.6E-03 91 5.2E-12 1.5E-02 90 6.6E-13 2.0E-03 91M 3.2E-12 9.4E-03 91 2.2E-12 6.5E-03 93 9.0E-13 2.7E-03 95 2.3E-12 6.7E-03
-95 2.4E-12 7.0E-03
-99 1.0E-08 3.0E+01 99M 6.7E-09 2.0E+01
-103 1.1E-12 3.2E-03
-106 1.1E-13 3.1E-04
-103M 1.0E-12 3.0E-03
-106 1.0E-13 3.0E-04 125M 2.9E-13 8.5E-04 127M 6.7E-12 2.0E-02 127 4.1E-12 1.2E-02 129M 1.2E-10 3.7E-01 129 7.6E-11 2.3E-01 131M 6.8E-11 2.0E-01 Page 1 of 3 Rev. 30
LLOWING ADDITION OF ANTICIPATED OPERATIONAL OCCURANCES AND DILUTION IN THE CIRCULATING WATER DISCHARGE SYSTEM Isotope Activity Released (Ci/gm) Activity Released (Ci/yr) 131 1.9E-11 5.5E-02 132 8.2E-10 2.4E+00
-137M 5.0E-09 1.5E+01
-140 1.4E-11 4.0E-02
-140 5.2E-12 1.5E-02
-141 2.2E-12 6.5E-03
-143 1.5E-12 4.5E-03
-144 1.6E-12 4.8E-03 143 2.1E-12 6.3E-03 144 1.6E-12 4.7E-03
-239 1.2E-11 3.5E-02 83 1.3E-10 3.8E-01 84 2.6E-11 7.9E-02 85 4.4E-13 1.3E-03 30 1.5E-11 4.3E-02 31 5.9E-09 1.8E+01 32 1.7E-09 5.0E+00 33 8.7E-09 2.6E+01 34 5.3E-10 1.6E+00 35 4.1E-09 1.2E+01
-86 1.5E-12 4.4E-03
-88 1.6E-09 4.6E+00
-134 1.9E-09 5.6E+00
-136 9.4E-10 2.8E+00
-137 9.4E-09 2.8E+01 3 8.6E-07 2.5E+03 Page 2 of 3 Rev. 30
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.
6.3E-12 = 6.3 x 10-12 icipated operational occurrences = 1.5E-01 Ci/yr.
ution release rate is 2.97E+15 gm/yr.
al release (excluding tritium) is 1.7E+02 Ci/yr.
al release concentration (excluding tritium) is 5.8E-08 Ci/gm.
Page 3 of 3 Rev. 30
BLE 11.2-10 FRACTION OF MPC RELEASED - DESIGN CASE (1) (HISTORICAL)
Fraction MPC Isotope Conc (Ci/cc) MPC (Ci/cc) Released Cr-51 6.3E-12 (1) 2.0E-03 3.1E-09 Mn-54 1.1E-12 1.0E-04 1.1E-08 Fe-55 5.5E-12 8.0E-04 6.9E-09 Fe-59 3.5E-12 5.0E-05 7.0E-08 Co-58 5.4E-11 9.0E-05 6.0E-07 Co-60 6.8E-12 3.0E-05 2.3E-07 Sr-89 1.4E-11 3.0E-06 4.7E-06 Sr-90 5.5E-13 3.0E-07 1.8E-06 Sr-91 5.2E-12 5.0E-05 1.0E-07 Y-90 6.6E-13 2.0E-05 3.3E-08 Y-91M 3.2E-12 3.0E-03 1.1E-09 Y-91 2.2E-12 3.0E-05 7.3E-08 Y-93 9.0E-13 3.0E-05 3.0E-08 Zr-95 2.3E-12 6.0E-05 3.8E-08 Nb-95 2.4E-12 1.0E-04 2.4E-08 Mo-99 1.0E-08 4.0E-05 2.5E-04 Tc-99M 6.7E-09 3.0E-03 2.2E-06 Ru-103 1.1E-12 8.0E-05 1.4E-08 Ru-106 1.1E-13 1.0E-05 1.1E-08 Rh-103M 1.0E-12 1.0E-02 1.0E-10 Te-125M 2.9E-13 1.0E-04 2.9E-09 Te-127M 6.7E-12 5.0E-05 1.3E-07 Te-127 4.1E-12 2.0E-04 2.0E-08 Te-129M 1.2E-10 2.0E-05 6.0E-06 Te-129 7.6E-11 8.0E-04 9.5E-08 Te-131M 6.8E-11 4.0E-05 1.7E-06 Te-132 8.2E-10 2.0E-05 4.1E-05 Ba-140 1.4E-11 2.0E-05 7.0E-07 Page 1 of 2 Rev. 30
Fraction MPC Isotope Conc (Ci/cc) MPC (Ci/cc) Released La-140 5.2E-12 2.0E-05 2.6E-07 Ce-141 2.2E-12 9.0E-05 2.4E-08 Ce-143 1.5E-12 4.0E-05 3.7E-08 Ce-144 1.6E-12 1.0E-05 1.6E-07 Pr-143 2.1E-12 5.0E-05 4.2E-08 Np-239 1.2E-11 1.0E-04 1.2E-07 Br-83 1.3E-10 3.0E-06 4.3E-05 I-130 1.5E-11 3.0E-06 5.0E-06 I-131 5.9E-09 3.0E-07 2.0E-02 I-132 1.7E-09 8.0E-06 2.1E-04 I-133 8.7E-09 1.0E-06 8.7E-03 I-134 5.3E-10 2.0E-05 2.6E-05 I-135 4.1E-09 4.0E-06 1.0E-03 Rb-86 1.5E-12 2.0E-05 7.5E-08 Cs-134 1.9E-09 9.0E-06 2.1E-04 Cs-136 9.4E-10 6.0E-05 1.6E-05 Cs-137 9.4E-09 2.0E-05 4.7E-04 H-3 8.6E-07 3.0E-03 2.9E-04 Totals 9.1E-07 2.4E-02 3.1E-02 TE
- 1. 6.3E-12 = 6.3 x 10-12 al concentration (excluding tritium) is 5.1E-08 Ci/gm.
al fraction of MPC released (excluding tritium) is 3.07E-02.
torical, not subject to future updating. This table has been retained to preserve original license s.
Page 2 of 2 Rev. 30
ABLE 11.2-11 ASSUMPTIONS USED FOR THE RADIOACTIVE LIQUID WASTE SYSTEM FAILURE (RELEASE TO ATMOSPHERE) AND FOR THE LIQUID CONTAINING TANK FAILURE Design nk with Assumed Highest Radionuclide Inventory Boron Recovery Tank nk Volume (gal) 150,000 l Fraction of Tank 0.8 actor Coolant Feed Rate to Tank (gpm) 75 gasification Fraction Upstream to Tank 1.0 urce Stream Feeding Tank Reactor Coolant (Table 11.1-2) ction of Volatile Nuclides Released to Atmosphere ble Gases 1.0 logens 0.1 ution Factor of Ground Water to Niantic Bay 73 nsition Time to Niantic Bay (years) (1) 6.64 ution Factor at Point of Entry to Niantic Bay 13,052 ution Factor at 1,000 Feet from Point of Entry to Bay 32,151 TE:
Transit time to Niantic Bay is the travel time of the ground water. Credit for sorption in soil was not assumed to this analysis.
Page 1 of 1 Rev. 30
TABLE 11.2-12 BORON RECOVERY TANK CONCENTRATIONS (CI/CC) otope Design (Ci/cc) Isotope Design (Ci/cc) Isotope Design (Ci/cc)
-83m 0.13E-03 (1) Y-94 0.39E-08 Te-127 0.19E-04
-131m 0.90E-03 Y-95 0.12E-08 Te-129m 0.43E-03
-133m 0.13E-03 Zr-95 0.79E-05 Te-129 0.42E-03
-133 0.24E-02 Zr-97 0.28E-05 Te-131m 0.20E-03
-135m 0.25E-02 Nb-95m 0.16E-06 Te-131 0.41E-04
-135 0.61E-02 Nb-95 0.83E-05 Te-132 0.27E-02 83 0.13E-03 Nb-97m 0.27E-05 Te-133m 0.14E-04 84 0.13E-04 Nb-97 0.31E-05 Te-133 0.15E-04 85 0.17E-06 Mo-99 0.33E-01 Te-134 0.14E-04 87 0.32E-07 Mo-101 0.24E-05 Cs-134 0.20E+00 29 0.53E-09 Mo-102 0.12E-05 Cs-136 0.10E+00 31 0.28E-01 Mo-105 0.95E-09 Cs-137 0.10E+01 32 0.37E-02 Tc-99m 0.27E-01 Cs-138 0.16E-01 33 0.31E-01 Tc-101 0.46E-05 Cs-139 0.45E-03 34 0.32E-03 Tc-102 0.16E-05 Cs-140 0.54E-05 35 0.85E-02 Tc-105 0.12E-07 Cs-142 0.22E-08 36 0.92E-07 Ru-103 0.38E-05 Ba-137m 0.93E+00
-81 0.11E-09 Ru-105 0.81E-07 Ba-139 0.52E-03
-83 0.24E-09 Ru-106 0.37E-06 Ba-140 0.48E-04
-84 0.17E-10 Ru-107 0.61E-11 Ba-141 0.23E-07
-88 0.33E-01 Rh-103m 0.38E-05 Ba-142 0.22E-07
-89 0.85E-03 Rh-105m 0.81E-07 La-140 0.23E-04
-90 0.40E-03 Rh-105 0.76E-06 La-141 0.67E-06
-91 0.64E-05 Rh-106 0.37E-06 La-142 0.79E-09
-92 0.28E-07 Rh-107 0.21E-09 La-143 0.21E-08 89 0.61E-04 Sn-127 0.21E-08 Ce-141 0.77E-05 90 0.19E-05 Sn-128 0.21E-08 Ce-143 0.45E-05 91 0.10E-04 Sn-130 0.14E-10 Ce-144 0.56E-05 92 0.13E-05 Sb-127 0.15E-06 Ce-145 0.59E-10 Page 1 of 2 Rev. 30
otope Design (Ci/cc) Isotope Design (Ci/cc) Isotope Design (Ci/cc) 93 0.46E-08 Sb-128 0.97E-11 Ce-146 0.10E-08 94 0.87E-10 Sb-129 0.87E-07 Pr-143 0.75E-05 90 0.23E-05 Sb-130 0.29E-09 Pr-144 0.56E-05 91m 0.65E-05 Sb-131 0.26E-08 Pr-145 0.56E-06 91 0.78E-05 Sb-132 0.27E-10 Pr-146 0.58E-08 92 0.28E-05 Sb-133 0.45E-10 Nd-147 0.26E-05 93 0.18E-05 Te-127m 0.24E-04 Nd-149 0.18E-07
-151 0.12E-09 Sm-151 0.44E-09 Mn-56 0.47E-04
-147 0.93E-06 Sm-153 0.13E-06 Fe-55 0.57E-04
-149 0.98E-06 Cr-51 0.11E-04 Fe-59 0.12E-04
-151 0.29E-06 Mn-54 0.88E-05 Co-58 0.30E-03 Co-60 0.86E-05 TE:
(1) 0.13E-03 = 0.13 x 10-3 Page 2 of 2 Rev. 30
BLE 11.2-13 ACTIVITY RELEASED TO ATMOSPHERE FROM A RADIOACTIVE LIQUID CONTAINING TANK FAILURE (BORON RECOVERY TANK)
Isotope Radioactivity Released (Ci)
Kr-83m 5.90E-02 (1)
Xe-131m 4.08E-01 Xe-133m 5.81E-02 Xe-133 1.08E+00 Xe-135m 1.15E+00 Xe-135 2.77E+00 I-131 1.28E+00 I-132 1.67E-01 I-133 1.40E+00 I-134 1.45E-02 I-135 3.84E-01 TE:
(1) 5.90E-02 = 5.90 x 10-2 Page 1 of 1 Rev. 30
TABLE 11.2-14 RADIOACTIVE CONCENTRATIONS IN GROUNDWATER ENTERING NIANTIC BAY FOLLOWING A RUPTURE OF BORON RECOVERY TANK Concentration at Entry Point Concentration 1000 Feet from Entry Isotope (Ci/cc) Point (Ci/cc) 29 *
- 90 1.72E-12**
- 90 1.72E-12 *
-106 3.99E-15 *
-106 3.99E-15
- 127m *
- 127 * *
-134 2.27E-08 7.07E-13
-137 9.22E-07 2.86E-11
-137m 8.48E-07 2.64E-11
-144 1.60E-14
- 144 1.60E-14 *
-147 1.75E-13 *
-151 * *
-54 4.34E-14 *
-55 1.09E-11 *
-58 * *
-60 3.77E-12
- 3 2.54E-06 7.90E-11 TES:
- Nuclide concentration less than 1.0E-15, Ci/cc
- 1.72E-12 = 1.72 x 10-12 Page 1 of 1 Rev. 30
figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
Rev. 30
figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
Rev. 30
s section describes the capabilities of Millstone 3 to control, collect, process, store, and ose of gaseous radioactive waste generated from normal operation and anticipated operational urrences. Section 11.5 describes the process and effluent radiation monitoring systems. The tor plant gaseous and aerated vents systems are described in Section 9.3.3. The gaseous waste agement systems include the radioactive gaseous waste system and ventilation systems.
radioactive gaseous waste system (Figure 11.3-1, Sheets 1 and 2) consists of three systems: the degasification subsystem, the process gas (hydrogenated) subsystem, and low vity process vent (aerated) subsystem.
he degasification subsystem, the fluid from the reactor coolant letdown stream (CHS), or rnately, from the reactor plant gaseous drains system (DGS), is sent to a degasifier, where condensable fission product gases are removed. The remaining liquid may be transferred to volume control tank (CHS) or to the boron recovery system (BRS). The normal flowpath is m the reactor coolant letdown to the volume control tank. The gases are forwarded to the cess gas portion of GWS.
he process gas (hydrogenated) subsystem, the noncondensable fission product gas stream is dehydrated. Then, radioactive iodine is removed and the activity of the radioactive xenon and pton is reduced. Finally, the gas is released into the process vent portion of GWS.
he low activity process vent (aerated) subsystem, aerated and hydrogenated gas streams from ous plant inputs (including the process gas portion of GWS) are collected, dehydrated, and harged to the reactor plant ventilation system (HVR) for release to the environment via the lstone stack. The gas streams are monitored for radioactivity prior to release.
separate flow path, relief effluents from the degasifier, the waste evaporator, and the boron porator are collected and discharged to the reactor plant ventilation system (HVR) for release he environment via the turbine building stack.
radioactivity values provided in this section are the design basis values used for the design of Gaseous Waste System. As such, they are considered historical and not subject to future ating. The information is retained to avoid loss of the original design bases. Actual airborne oactivity release quantities can be found in the annual radioactive effluent release reports as mitted to the NRC.
.1 DESIGN BASES
.1.1 Design Objective objective of the gaseous waste management system is to process and control the release of oactive gaseous effluents to the site environs so as to maintain the exposure to radioactive eous effluents of persons in unrestricted areas to as low a level as is reasonably achievable pendix I to 10 CFR 50, May 5, 1975). This is to be accomplished while also maintaining 11.3-1 Rev. 30
.1.2 Design Criteria design of the radioactive gaseous waste system and the ventilation systems meet the owing criteria.
- 1. Section 11.1 discusses the design basis source terms. Expected radioactive gaseous effluents from all sources (Table 11.3-1) have been calculated using the data shown in Table 11.3-2. These values are consistent with NUREG-0017, April 1976.
- 2. The systems have the capability to meet the requirements of 10 CFR 20 and the dose design objectives specified in Appendix I to 10 CFR 50, including provisions to treat gaseous radioactive wastes such that:
- a. The calculated annual total quantity of all radioactive material released from Millstone 3 to the atmosphere does not result in an estimated annual external dose from gaseous effluents to any individual in unrestricted areas in excess of 5 millirems to the total body or 15 millirems to the skin.
- b. The calculated annual total quantity of all radioactive iodine and radioactive material in particulate form released from Millstone 3 to the atmosphere does not result in an estimated annual dose or dose commitment from such radioactive iodine and radioactive material in particulate form for any individual in an unrestricted area from all pathways of exposure in excess of 15 mRem to any organ.
- c. The concentrations of radioactive materials in gaseous effluents released to an unrestricted area do not exceed the limits in 10 CFR 20, Appendix B, Table 2, Column 1.
- 3. The radioactive gaseous waste system is designed to meet the anticipated processing requirements of the plant. Adequate 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 accordance with the guidelines of Regulatory Guide 1.143 (formerly Branch Technical Position ETSB 11-1, Rev. 1)
(Section 1.8).
- 5. The radioactive gaseous waste system meets General Design Criteria 60 and 64 of Appendix A to 10 CFR 50 as discussed in Sections 3.1.2.60 and 3.1.2.64.
11.3-2 Rev. 30
endix 11A shows that the systems contain all items of reasonably demonstrated technology affect a reduction in dose to the population reasonably expected to be within a 50-mile radius he plant with a favorable cost benefit.
.1.4 Equipment Design Criteria le 11.3-3 lists the radioactive gaseous waste system major equipment items. This list includes erials, rates process conditions, and number of units supplied. Equipment and piping are gned and constructed in accordance with the requirements of the applicable codes ble 11.3-4).
le 3.2-1 shows the safety classes of the various systems. Seismic category, safety class, quality rance requirements, and principal construction codes information is contained in Section 3.2.
system is designed to Safety Classification NNS.
design of the system precludes an explosive mixture from accumulating. Since the system rates above atmospheric pressure, in-leakage cannot occur. Instrumentation with automatic m functions monitors the concentrations of hydrogen and oxygen in portions of the system ing the potential for containing explosive mixtures.
radioactive gaseous waste system processes letdown from the chemical and volume control em (CHS) or reactor plant gaseous drain through the degasifier at a maximum rate of 150
. Maximum letdown from the CHS is 120 gpm and the maximum flow rate from the gaseous ns system is no more than 150 gpm.
following design features are incorporated to minimize maintenance, equipment downtime, age, and radioactive gaseous releases. These features facilitate radwaste operation, and assist aintaining occupational exposures as low as is reasonably achievable.
- 1. Redundant degasifier recirculation 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.
servative analyses of the radioactive gaseous waste system, presented in Section 15.7, onstrate that equipment failure results in doses well within the guidelines of 10 CFR 100.
11.3-3 Rev. 30
ure 11.3-2 gives a composite diagram of ventilation systems which may release radioactivity ng normal operations. These systems are:
- 1. Fuel building ventilation (Section 9.4.2)
- 2. Auxiliary building ventilation (Section 9.4.3)
- 3. Turbine building ventilation (Section 9.4.4)
- 4. Containment structure ventilation (Section 9.4.7)
- 5. Engineered safety features building ventilation (Section 9.4.5)
- 6. Waste disposal building ventilation (Section 9.4.9)
- 7. Service building ventilation (Section 9.4.12)
.2 SYSTEM DESCRIPTIONS ures 11.3-1, Sheets 1 & 2 show the piping and instrumentation drawings (P&IDs) of the oactive gaseous waste system and Figure 9.3-5 shows reactor plant gaseous drains.
appropriate subsections of Section 9.4 provide specific component data with P&IDs for tilation systems subject to radioactive release.
.2.1 Radioactivity Inputs and Release Points ioactivity in process streams is processed by the radioactive gaseous waste system.
tilation releases for building housing systems which could potentially be radioactive during mal operations are:
- 1. Ventilation vent
- a. Containment
- b. Auxiliary building*
- c. Fuel building*
r calculation purposes, releases from the fuel building, waste disposal building, service ding, and engineered safety features building are combined with auxiliary building releases, ccordance with NUREG-0017, April 1976.
11.3-4 Rev. 30
- e. Service building*
- 2. Engineered safety features building*
- 3. Millstone stack
- a. Radioactive gaseous waste system
- b. Main condenser air ejector
- 4. Turbine Building
- a. Roof exhausters
- b. Steam generator blowdown flash tank vent
- 5. Condensate polishing building
- a. Turbine gland sealing system exhaust ioactivity releases are provided in Tables 11.3-5 through 11.3-10.
lding volumes and expected flow rates are provided in Section 9.4.
.2.2 Degasifier Subsystem of Radioactive Gaseous Waste System ctor coolant letdown, containing dissolved hydrogen and fission gases, is normally directed to degasifier from the letdown line upstream of the volume control tank in the chemical and ume control system. Alternately, liquid collected by the reactor plant gaseous drains system ction 9.3.3) may be also directed to either the degasifier or the boron recovery system.
solved gases are separated from the liquid in the degasifier.
degasifier processes reactor coolant letdown continuously except as needed to process eous (hydrogenated) drains. However, reactor coolant letdown may bypass the degasifier, if red. The degasifier design flow of 150 gpm exceeds the maximum expected throughput for liquid portion of the process gas subsystem. Separation of dissolved gases at all reactor lant letdown flow rates is thus ensured. The degasifier operates at approximately 2 psig. If the asifier is not operating, it may be placed in either the standby mode or in a shutdown dition.
11.3-5 Rev. 30
process gas (hydrogenated) portion of the radioactive gaseous waste system is designed to t gases stripped from in the reactor coolant letdown and the reactor plant gaseous drogenated) drains (Figure 11.3-1).
uent gases from the degasifier contain primarily hydrogen and water vapor. A small amount of ogen and traces of xenon, krypton, argon, carbon, and iodine are also present in the effluent es. These gases and any hydrogenated gas stream from the reactor plant gaseous vent header dehumidified (dew point 35°F) in one of the two redundant process gas refrigerant dryers.
densation effluent from the dryers is returned to the suction of the degasifier recirculation
- p. The dry stream is passed through and filtered by the ambient temperature process gas rcoal bed adsorbers and one of two redundant HEPA filters. The heat due to radioactive decay mall and does not affect the adsorption of noble gases on the charcoal. The charcoal bed orbers are designed to provide holdup of most krypton and xenon isotopes long enough in parison with their half-lives so that passage through the beds will result in the effective oval of these isotopes. In addition, decontamination of iodine to negligible levels is obtained ng passage through the charcoal beds. The charcoal is divided evenly between two vertical s in series. The tanks are piped so that either one may be bypassed, if necessary, with a esponding decrease in decay time; however, bypass is not anticipated. The only radioisotope ent in any quantity in the predominantly hydrogen stream after the decay period is pton-85. This processed hydrogenated stream is monitored by the supplementary leak ection monitor and released to the environment through the Millstone stack in accordance h technical specifications. The normal flow path for the processed stream is to the Millstone
- k. When a test of the gas at the vent dampers at the discharge of the process vent fans is uired, then the process gas flow is monitored and released through the reactor plant ventilation t.
uid seals are provided in the drain lines from the process gas precoolers and the process gas er traps to prevent backflow of vapor from the degasifier. The process gas precooler drains tinuously at approximately 3 pounds per hour of water at 120°F. The process gas water trap ns at approximately 0.3 pound per hour at 35°F. In the event of a sudden vacuum condition at suction of the degasifier recirculation pump, the seals are temporarily lost. However, the cess gas precooler drain seal refills in approximately 10 minutes and the process gas water trap refills in approximately 100 minutes.
rocess gas monitor is provided for the gaseous releases from the process gas portion to nitor radioactivity release to the environment and to automatically isolate the flow from the cess gas receiver when a predetermined level is exceeded (Section 11.5).
.2.4 Process Vent Portion of Radioactive Gaseous Waste System process vent (aerated) portion of the radioactive gaseous waste system is designed to collect low activity aerated gas stream from the following sources:
- 1. Reactor plant aerated vents system 11.3-6 Rev. 30
- 3. Condenser air removal system
- 4. Containment vacuum system
- 5. Low activity effluent from the process gas (hydrogenated) portion of the radioactive gaseous waste system
- 6. Radioactive gaseous waste system component purges
- 7. Boron recovery system relief valve discharge
- 8. Liquid waste system relief valve discharges
- 9. Degasifier relief valve discharge process vents, with the exception of the relief valve vent header, are monitored by the plementary leak collection monitor and discharged to the Millstone stack. The process vent ion is operated continuously, unless required to be shutdown for maintenance. The relief valve t header is discharged to the ventilation vent where it is monitored and released without ation.
.2.5 Steam and Power Conversion System main condenser is evacuated by steam jet air ejectors (Section 10.4.2). Air ejector exhaust is nitored and discharged to the Millstone stack.
turbine gland seal steam condenser (Section 10.4.3) exhausts directly to the atmosphere.
se releases are given in Tables 11.3-1 and 11.3-11 for expected and design cases, respectively.
ing the intermittent and hot standby blowdown (open cycle), the steam generator flash tank is ted directly to the atmosphere through a vent on the turbine building roof. Releases are given ables 11.3-7 and 11.3-10.
quantity of steam released during steam dumps to the atmosphere is provided in Section 10.3.
ual unit trips are expected to be less than those assumed in NUREG-0017 (2 turbine trips per r).
.2.6 System Instrumentation Requirements
.2.6.1 Radioactive Gaseous Waste System process gas radiation monitor is located downstream of the process gas charcoal bed orbers prior to discharge into the process vent portion. This monitor automatically isolates from the process gas receiver. Radioactive releases from the process vent portion of the 11.3-7 Rev. 30
rogen and oxygen analyzers are provided to measure the process gas composition and to m before a dangerous concentration exists.
perature and moisture detectors and alarms are provided downstream of the process gas igerant dryer to assure proper moisture control in the process gas charcoal bed adsorbers.
.2.6.2 Ventilation Systems iation monitoring of ventilation system effluents is discussed in Section 11.5. Instrumentation uirements are discussed in Section 9.4.
.2.7 Seismic Design Provisions of the Radioactive Gaseous Waste System radioactive gaseous waste system has not been provided with special seismic design.
wever, the entire portion of the system that provides for treatment of gases stripped from the tor coolant is located in the auxiliary building, a Seismic Category I building (Section 3.8).
ventilation in the auxiliary building system (Section 9.4.3) has the capability of detecting oactive gas leakage and filtering prior to release. This is an alternate method of meeting the gn guidance given in Regulatory Guide 1.143.
.2.8 Quality Control rogram is established to ensure that the design, construction, and testing requirements are met.
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 therefrom 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 instructions, procedures, and drawings for accomplishing the activity.
- 3. Handling, Storage, and Shipping - Procedures 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 - Procedures are established to provide for the identifications of items which have satisfactorily passed required inspections and tests.
11.3-8 Rev. 30
and equipment, and nonconformances, are promptly identified and corrected.
.2.9 Welding welding constituting the pressure boundary of pressure retaining components is performed by lified welders employing qualified welding procedures according to Table 11.3-4.
consumable weld inserts are prohibited in process lines unless they are ground out after the d is completed.
.2.10 Materials erials for pressure retaining components of process systems are selected from those covered he material specifications listed in Section II, Part A of the ASME Boiler and Pressure Vessel e, except that malleable, wrought, cast iron materials, or plastic pipe is not used. The ponents meet all of the mandatory requirements of the material specifications with regard to ufacture, examination, repair, testing, identification, and certification.
escription of the major process equipment, including the design temperature and pressure and materials of construction, is given in Table 11.3-3.
.2.11 Construction of Process Systems ssure retaining components of process systems use welded construction to the maximum ticable extent. Process piping systems include the first root valve on sample and instrument
- s. Process lines are not less than 3/4-inch nominal pipe size. Sample and instrument lines are considered as portions of the process systems. Flanged joints or suitable rapid disconnect ngs are not used except where maintenance requirements clearly indicate that such struction is preferable. Screwed connections in which threads provide the only seal are not
- d. Screwed connections backed up by seal welding or mechanical joints are used only on lines
/4-inch nominal pipe size.
nes 3/4-inch or greater, but less than 2 1/2-inch nominal pipe size, socket type welds are used.
nes 2 1/2-inch nominal pipe size and larger, pipe welds are of the butt joint type.
.2.12 System Integrity Testing mpleted process systems are pressure tested to the maximum practicable extent. Piping ems are hydrostatically tested in their entirety, using available valves for pressure test ndaries (lines vented to atmosphere do not require a pressure test). Hydrostatic testing of ng systems is performed at a pressure 1.2 times the design pressure and held for a minimum of minutes with no leakage indicated. Pneumatic testing may be substituted for hydrostatic ing in accordance with the applicable codes.
11.3-9 Rev. 30
les 11.3-5 thru 11.3-10 give the calculated sources of radioactive nuclide inventory released gaseous effluents. Table 11.3-1 gives the expected radioactive gaseous isotope releases from h release point assumed in terms of curies per year per nuclide. Table 11.3-11 provides the gn releases for these release points.
le 11.3-2 lists the parameters in these calculations.
ummary of the estimated expected annual radioactivity doses is presented in Appendix 11A. A mary of design release concentrations at the site boundary, maximum permissible centration (MPC) and Fraction of MPCs is presented in Tables 11.3-8, 11.3-9 and 11.3-10.
design releases are within the limits of 10 CFR 20. The doses from expected releases are hin the numerical design objectives of Appendix I of 10 CFR 50. Atmospheric diffusion and und deposition factors used in the dose calculations are discussed in Section 2.3.5.
eous effluents are discharged to the environment through the following release points:
- 1. The reactor plant ventilation vent
- 2. The turbine building roof exhausters
- 3. The Millstone stack
- 4. The engineered safety features building exhaust
- 5. The steam generator blowdown flash tank vent
- 6. The turbine gland seal steam condenser exhaust on the condensate polishing building roof ease points are identified on Figure 11.3-2.
austs from the auxiliary building, the fuel building, the waste disposal building, the tainment purge, the service building, and the gaseous waste process vent are released from the tor plant ventilation vent. This vent is located on the turbine building 133 feet above grade 157 feet-0 inches above sea level. The base elevation is 75 feet above grade. The square vent s-sectional dimensions are 10 feet by 10 feet and the discharge velocity is 3,000 feet per ute. The maximum discharge temperature is 104°F. Containment purge is considered an rmittent release. Others are assumed as continuous.
turbine building ventilation system exhausts through the turbine building roof exhausters, ted on the turbine building roof. The exhausters are approximately 114 feet above grade, feet above sea level, and have a base elevation of 107 feet above grade. The system exhausts ugh roof-mounted, mushroom-type hoods with dimensions of 14 feet-6 inches by 11.3-10 Rev. 30
ing the open cycle blowdown, the steam generator blowdown flash tank is vented directly to atmosphere through a vent on the turbine building roof. The vent is a 10-inch diameter pipe ched to a roof-mounted silencer. The discharge elevation is approximately 113 feet above sea
- l. Exhaust flow velocity is 8,450 feet per minute.
turbine gland sealing system exhaust is vented to the atmosphere at a point above the densate polishing building. The vent is a 10-inch diameter pipe at an elevation of 72 feet ve sea level.
reactor plant aerated vents, the reactor plant gaseous vents, condenser air removal effluent, oactive gaseous waste system discharges, steam generator blowdown tank condenser vent, containment vacuum pump discharge are released through the Millstone stack. The Millstone k is 375 feet above grade, 389 feet above sea level, and has a circular orifice with a 7 foot de diameter. The stack discharges at a maximum velocity of 332 feet per minute for accident dition and 40 feet per minute for normal condition (MP2 & MP3) at a maximum temperature 04°F. These discharges are assumed as continuous releases. The portion of the discharge vent ween the ESF building and the stack is underground.
engineered safety features building ventilation system exhausts through the ESF building t located on the east wall. The vent is 12 feet-9 inches above grade, 36 feet-9 inches above sea
- l. The vent cross sectional dimensions are 3 feet-0 inches by 10 feet-0 inches and the harge velocity is 350 feet per minute. The maximum discharge temperature is 104°F. The ase from the ESF building is considered continuous and is included as part of the continuous ase from the auxiliary building in Table 11.3-5.
.3.1 Radioactive Gaseous Waste System Failure aseous waste system failure is postulated to produce a unique unplanned release by a pathway normally used for planned releases and requiring a reasonable time to detect and take remedial on to terminate the release. An inadvertent bypass of the process gas charcoal bed adsorbers h continued normal operation of the gaseous waste system for one hour is assumed. This rce is then assumed to be continually released to the auxiliary building and then directly to the ironment without holdup.
gaseous waste system design, described in Section 11.3, precludes the single failure of an ve component from causing this event, but it is postulated as the design basis accident as cribed by Branch Technical Position ETSB 11-5. For this event to occur the two manual ass valves must fail or be improperly aligned, and the piping system downstream of the orbers must experience a passive failure to result in a release directly into the auxiliary ding.
11.3-11 Rev. 30
rnal pressure is approximately 1 psig, a gross rupture of the tanks is not considered credible.
iation monitors are provided for the ventilation system in order to detect the release of noble es to the environs from the building ventilation system.
wever, no credit is taken for normal operating plant systems, instrumentation or controls, nor ration of any engineered safety feature systems to mitigate the consequences of this event.
mally, all noncondensable gases are removed from the reactor coolant letdown stream in the asifier. For this analysis it is assumed that the activity released following the bypass of process charcoal bed adsorbers consists of noble gas activities discharged from the degasifier for 60 utes and released without benefit of delay in charcoal beds. It is also assumed that a fraction of noble gases adsorbed in the charcoal beds is released. The fractions released were calculated g the model developed for the Fast Flux Test Facility (Underhill). Table 11.3-13 lists the owing:
aximum radioisotope inventory in the process gas charcoal bed adsorbers and associated ng.
raction of the charcoal bed radioisotope inventory released.
ixty-minute discharge from the degasifier.
otal radioisotope release.
le 11.3-12 gives the assumptions used for analyzing the postulated bypass of the process gas rcoal bed adsorbers.
radiological consequences of bypassing the charcoal bed adsorbers are listed in Table 15.0-8 ed on design release assumptions in Table 11.3-12, the releases in Table 11.3-13 and the X/Q es in Table 15.0-11. The dose methodology of Appendix 15A is used here.
rder to bound the Stretch Power Uprate (SPU) to 3723 MWt (including 2% calorimetric ertainty), each of the noble gas isotopes released in Table 11.3-13 were scaled based on factors rmined from the ratio of the primary activity concentrations (RCS concentration from Table
-10 and Design concentration from Table 11.1-2) to determine the updated doses for the SPU rating condition.
s event will not cause a Condition IV event as defined in Section 15.0.1. The radiological sequences are well within the guidelines of 10 CFR 100.
.4 REFERENCE FOR SECTION 11.3 11.3-12 Rev. 30
ATMOSPHERE FROM MILLSTONE 3 (HISTORICAL)
Condensate Millstone Ventilation Turbine Polishing Stack Vent Building Building (1) Total uclide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr)
-83m 2.9E-01 (2) 4.7E-01 3.1E-05 1.5E-04 7.6E-01 85m 1.2E+00 2.0E+00 1.3E-04 6.5E-04 3.2E+00
-85 2.4E+02 1.5E+00 3.0E-06 1.5E-05 2.4E+02
-87 8.7E-01 1.4E+00 9.2E-05 4.6E-04 2.3E+00
-88 2.5E+00 4.0E+00 2.7E-04 1.3E-03 6.5E+00
-89 8.2E-02 1.3E-01 8.7E-06 4.3E-05 2.1E-01
-131m 2.4E-01 9.6E-01 8.1E-06 4.0E-05 1.2E+00
-133m 5.4E-01 1.9E+00 5.9E-05 3.0E-04 2.4E+00
-133 2.3E+01 1.5E+02 2.4E-03 1.2E-02 1.7E+02
-135m 2.1E-01 3.4E-01 2.2E-05 1.1E-04 5.5E-01
-135 2.9E+00 5.0E+00 3.0E-04 1.5E-03 7.9+00
-137 1.5E-01 2.3E-01 1.6E-05 7.8E-05 3.8-01
-138 7.2E-01 1.1E+00 7.6E-05 3.8E-04 1.8E+00 31 1.9E-02 4.6E-02 7.3E-03 2.2E-04 7.3E-02 33 2.8E-02 6.8E-02 9.9E-03 3.1E-04 1.1E-01
-58 0.0 6.4E-02 0.0 0.0 6.4E-02
-60 0.0 2.9E-02 0.0 0.0 2.9E-02
-54 0.0 1.9E-02 0.0 0.0 1.9E-02
-59 0.0 6.4E-03 0.0 0.0 6.4E-03 89 0.0 1.4E-03 0.0 0.0 1.4E-03 90 0.0 2.1E-04 0.0 0.0 2.1E-04
-134 0.0 1.9E-02 0.0 0.0 1.9E-02
-137 0.0 3.2E-02 0.0 0.0 3.2E-02 14 7.0E+00 1.0E+00 0.0 0.0 8.0E+00
-41 0.0 2.5E+01 0.0 0.0 2.5E+01 3 0.0 7.3E+02 0.0 0.0 7.3E+02 Page 1 of 2 Rev. 30
Rev. 30 TABLE 11.3-2 RADIOACTIVE GASEOUS SOURCE TERM PARAMETERS (HISTORICAL) nt Capacity Factor 0.8 el Defects (%) 0.12 (Expected) 1 (Design) ntainment Building:
Noble gas release to containment building (fraction/day of primary 0.01 coolant activity)
Iodine release to containment building (fraction/day of primary coolant 10-5 activity)
Purge exhaust ventilation rate (cfm) 35,000 Purge exhaust ventilation time per purge (hr) 8 Containment air filtration subsystem recirculation rate during purge 12,000 (1)
(cfm)
Charcoal iodine adsorber depth (in) 4 Iodine exhaust filter efficiency (%) 95 Particulate exhaust filter efficiency (%) 95 Number of cold purges/year 4 Continuous ventilation exhaust rate (cfm) 0 Free containment volume (cu ft) 2.32 x 106 ntainment Internal Cleanup System:
Operates prior to purging cold shutdown (hr) 16 Operates prior to purging hot shutdown (hr) 16 Mixing efficiency (%) 70 Containment air filtration subsystem recirculation rate prior to purge 12,000 (1)
(cfm)
Charcoal iodine adsorber depth (in) 4 Iodine filter efficiency (%) 90 Particulate filter efficiency (%) 90 xiliary Building:
Iodine exhaust filter efficiency (%) (2) 0 Particulate exhaust filter efficiency (%) (2) 0 Page 1 of 3 Rev. 30
Primary coolant leakage rate into building (lb/day) 160 Iodine partition factor 0.0075 rbine Building:
No special design to collect valve leakage)
Steam leakage (lb/hr) 1,700 in Condenser/Air Ejector:
Volatile iodine/total iodine in primary system 0.05 Volatile iodine is treated as noble gas in steam generator Primary to secondary leak rate (lb/day) 1370 (Design) 100 (Expected)
MC/AE volatile iodine partition factor 0.15 (Design)
Charcoal iodine adsorber depth (in) 0.0 Iodine exhaust filter efficiency (%) 0.0 Particulate exhaust filter efficiency (%) 0.0 Volatile iodine condenser bypass fraction 0.35 am Generator Blowdown:
Flash tank vented to the atmosphere during open cycle blowdown Flash tank iodine partition factor 0.05 Hypothetical flow assumptions (see Section 11.2.2.3 for time periods for each release)
Hot Standby: 150,520 lb/hr 1% MSR from four steam generators (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) rbine Gland Sealing System Exhaust:
Total steam flow rate 8,460 lb/hr Iodine partition factor 0.15 dioactive Gaseous Waste System (Process Gas System):
Letdown flow to degasifier (lb/hr) 35,900 Page 2 of 3 Rev. 30
Holdup time prior to charcoal beds (minutes) 7.41 Krypton dynamic adsorption coefficient (cm3/gm) 6.3 Xenon dynamic adsorption coefficient (cm3/gm) 146 System flow rate (scfm) expected/maximum 0.3/3 Total mass of charcoal in beds (lb x 1,000) 27 (No iodine or particulates are released from system)
Krypton holdup time in delay bed (hr) 147 Xenon holdup time in delay bed (hr) 3,410 (Complete degasification is handled by the same equipment as normal operation)
TES:
This is the flow rate for one fan unit. The containment air filtration subsystem includes two fan units, which operate simultaneously.
Filters installed but not used during normal operation. Capability exists to filter exhaust upon high activity in building.
Data and Assumptions from NUREG-0017, April 1976.
torical, not subject to future updating. This table has been retained to preserve original license s.
Page 3 of 3 Rev. 30
TABLE 11.3-3 RADIOACTIVE GASEOUS WASTE SYSTEM cess Gas Charcoal Bed Absorbers Parameters Number 2 Vessel pressure Operating (psig) 1 Design (psig) 335 Vessel temperature Operating (psig) 104 Design (psig) 150 Total weight - one unit Empty 16,650 Operating 30,150 Construction material Stainless steel gasifier Number 1 Capacity (gpm) 150 Pressure Operating (psig) 2 Design (psig) 150 + full vac.
Temperature Operating (°F) 219 Design (°F) 366 and 100 Construction material Stainless steel cess Gas Receiver Number 1 Pressure Page 1 of 8 Rev. 30
Operating (°F) Approximately Atmospheric Design (°F) 1,800 Temperature Operating (°F) 120 Design (°F) 450 Total weight Empty (lb) 1,755 Full (lb) 1,762 Construction material Stainless steel gasifier Recovery Exchangers Shell Side Tube Side Number 2 Total duty (Btu/hr) 6,300,000 Total liquid entering (lb/hr) 75,000 75,000 Pressure Operating - inlet (psig) 100 150 Design (psig) 300 300 Temperature In (°F 219 115 Out (°F) 135 199 nstruction material Stainless Stainless steel steel Page 2 of 8 Rev. 30
gasifier Feed Preheater Shell Side Tube Side Number 1 Total duty (Btu/hr) 4,600,000 Total fluid entering (lb/hr) 5,400 75,000 Pressure Operating - inlet (psig) 145 50 Design (psig) 300 + full 300 + full vac. vac.
Temperature In (°F 363 199 Out (°F) 363 260 Construction material Carbon steel Stainless steel gasifier Condenser Shell Side Tube Side Number 1 Total duty (Btu/hr) 3,219,000 Pressure Operating - inlet (psig) 125 2 Design (psig) 185 50 + full vac.
Temperature In (°F 95 219 Out (°F) 116 190 Construction material Carbon steel Stainless steel cess Gas Compressor (Abandoned in Place, Currently Bypassed)
Number 2 Capacity (scfm) 3 Page 3 of 8 Rev. 30
Pressure - suction Minimum (psig) 14.7 Maximum (psig) 240 Pressure Discharge (psig) 75 Design (psia) 1,800 Discharge temperature Leaving aftercooler, max 110
)
cess Gas Compressor Aftercooler Number 2 Gas flow (cfm) 2.4 Gas temperature Inlet (°F 408 Outlet (°F) 105 Cooling water flow (gpm) 2 Cooling water temperature Inlet (°F 95 Outlet (°F) 100 Pressure Operating (psig) 75 Design (psia) 1,800 cess Gas Prefilter Max Design Refueling Purge Number 2 Pressure Operating (psig) 1 Design (psia) 335 Temperature Operating (psig) 104 Page 4 of 8 Rev. 30
Design (psia) 150 tration efficiency (%) (based on 99.97 99.97 99.97 99.97 P test) cess Gas Prefilter Max Design Refueling Purge w rate (scfm) 8 9 3 20 ssure drop (in H20) 0.5 0.5 4 1 cess Gas H2/02 Analyzers H2 Analyzer 02 Analyzer ssure Operating - inlet (psig) 12 12 Minimum operating (psig) 2 2 Maximum operating (psig) 40 40 Design (psig) 40 40 Relief valves set (psig) 40 40 sign Temperature In (°F) 104 104 Out (°F) 104 104 alyzers Required Flow (cc/min) normal 100-125 100-125 maximum 150 150 gasifier Trim Cooler Shell Side Tube Side Number 1 Total duty (Btu/hr) 1,500,000 Temperature In (°F) 95 135 Out (°F) 110 115 Pressure Operating inlet (psig) 125 85 Page 5 of 8 Rev. 30
Design (psig) 175 300 Construction material Carbon steel Stainless steel cess Gas Precooler Shell Side Tube Side Number 2 Total duty (Btu/hr) 12,900 Total fluid entering (lb/hr) 27.1 2,600 Pressure Operating inlet (psig) 1 125 Design (psig) 335 200 Temperature In (°F) 190 95 Out (°F) 120 100 Design (°F) 190 120 Number of passes 1 2 Construction material Stainless Stainless steel steel cess Gas Glycol Chiller Refrigerant Process Gas Tube Side Tube Side Number 2 Total fluid entering (lb/hr) 50.7 Pressure Inlet (psig) 20 1 Design (psig) 150 335 Temperature In (°F) 34 120 Out (°F) 30 35 Construction material Stainless Stainless steel steel Page 6 of 8 Rev. 30
cess Gas Water Trap Number 2 Pressure Operating (psig) 1 Design (psig) 335 Temperature Operating (°F) 35 Design (°F) 150 Construction material Stainless steel gasifier Recirculation Pump Number 2 Capacity (gpm) @ 386 ft head 150 Pressure Operating-discharge (psig) 186 Design (psig) 200 Temperature Operating (°F) 218 Design (°F) 250 Construction material Stainless steel cess Vent Fans Number 2 Capacity (cfm) @ 20 in water 180 Operations pressure - discharge (psig) 0.4 Operating temperature (°F) 50 Outlet air velocity (fpm) 1,939 cess Vent Cooler Shell Side Tube Side Number 1 Page 7 of 8 Rev. 30
Total duty (Btu/hr) 148,828 Total fluid entering (lb/hr) 14,828 752.5 Pressure Operating - inlet (psig) 125 14.6 Design (psig) 150 75 Temperature In (°F) 45 180 Out (°F) 55 50 Construction material Carbon steel Stainless steel Page 8 of 8 Rev. 30
radioactive gaseous waste, the reactor plant aerated vents, and the reactor plant gaseous vents em shall be designed and constructed in accordance with the requirements of the following es and standards:
System Component Safety Class Codes and Standards (1) (2) ing, fittings NNS ASME III, Class 3 ANSI B31.1 lves NNS ASME III, Class 3 ANSI B16.5 ANSI B16.34 sorbers, degasifier, heat exchangers, NNS ASME VIII, Division 1 ers, water removal equipment mps and compressors NNS ASME III, Class 3 ns NNS Manufacturers standards ter assemblies NNS ASME VIII, Division 1 trumentation and controls N/A tors NEMA MG.1 TES:
Portions of the system were procured to ASME Section III requirements and are defined in the procurement specification.
The overall system classification is NNS.
Page 1 of 1 Rev. 30
VIA VENTILATION VENT (HISTORICAL)
Containment Auxiliary Building Nuclide Building (Ci/yr) (Ci/yr) (1) Total (Ci/yr)
Kr-83m 5.7E-05 4.7E-01 4.7E-01 Kr-85m 2.0E-02 2.0E+00 2.0E+00 Kr-85 1.4E+00 4.5E-02 1.5E+00 Kr-87 7.4E-06 1.4E+00 1.4E+00 Kr-88 5.7E-03 4.0E+00 4.0E+00 Kr-89 0.0 1.3E-01 1.3E-01 Xe-131m 8.5E-01 1.2E-01 9.6E-01 Xe-133m 1.0E+00 8.7E-01 1.9E+00 Xe-133 1.1E+02 3.6E+01 1.5E+02 X-135m 0.0 3.4E-01 3.4E-01 Xe-135 3.4E-01 4.7E+00 5.0E+00 Xe-137 0.0 2.3E-01 2.3E-01 Xe-138 0.0 1.1E+00 1.1E+00 I-131 2.8E-06 4.6E-02 4.6E-02 I-133 2.8E-07 6.8E-02 6.8E-02 Co-58 3.8E-03 6.0E-02 6.4E-02 Co-60 1.7E-03 2.7E-02 2.9E-02 Mn-54 1.1E-03 1.8E-02 1.9E-02 Fe-59 3.8E-04 6.0E-03 6.4E-03 Sr-89 8.5E-05 1.3E-03 1.4E-03 Sr-90 1.5E-05 2.0E-04 2.1E-04 Cs-134 1.1E-03 1.8E-02 1.9E-02 Cs-137 1.9E-03 3.0E-02 3.2E-02 C-14 1.0E+00 0.0 1.0E+00 Ar-41 2.5E+01 0.0 2.5E+01 H-3 1.3E+02 6.0E+02 7.3E+02 Page 1 of 2 Rev. 30
Rev. 30 FROM MILLSTONE 3 VIA MILLSTONE STACK (HISTORICAL)
Main Condenser/ Air Radioactive Gaseous Nuclide Ejector (Ci/yr Waste System (Ci/yr) Total (Ci/yr)
Kr-83m 2.9E-01 0.0 2.9E-01 Kr-85m 1.2E+00 0.0 1.2E+00 Kr-85 2.8E-02 2.4E+02 2.4E+02 Kr-87 8.7E-01 0.0 8.7E-01 Kr-88 2.5E+00 0.0 2.5E+00 Kr-89 8.2E-02 0.0 8.2E-02 Xe-131m 7.3E-02 1.7E-01 2.4E-01 Xe-133m 5.4E-01 0.0 5.4E-01 Xe-133 2.3E+01 0.0 2.3E+01 Xe-135m 2.1E-01 0.0 2.1E-01 Xe-135 2.9E+00 0.0 2.9E+00 Xe-137 1.5E-01 0.0 1.5E-01 Xe-138 7.2E-01 0.0 7.2E-01 I-131 1.9E-02 0.0 1.9E-02 I-133 2.8E-02 0.0 2.8E-02 Co-58 0.0 0.0 0.0 Co-60 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0 Fe-59 0.0 0.0 0.0 Sr-89 0.0 0.0 0.0 Sr-90 0.0 0.0 0.0 Cs-134 0.0 0.0 0.0 Cs-137 0.0 0.0 0.0 C-14 0.0 7.0E+00 7.0E+00 Ar-41 0.0 0.0 0.0 H-3 0.0 0.0 0.0 torical, not subject to future updating. This table has been retained to preserve original license s.
Page 1 of 1 Rev. 30
ATMOSPHERE VIA TURBINE BUILDING (HISTORICAL)
Steam Generator Blowdown Nuclide Roof Exhausters (Ci/yr) Flash Tank Vent (2) (Ci/yr) Total (Ci/yr)
Kr-83m 3.1E-05 (3) 0.0 3.1E-05 Kr-85m 1.3E-0 0.0 1.3E-04 Kr-85 3.0E-06 0.0 3.0E-06 Kr-87 9.2E-05 0.0 9.2E-05 Kr-88 2.7E-04 0.0 2.7E-04 Kr-89 8.7E-06 0.0 8.7E-06 Xe-131m 8.1E-06 0.0 8.1E-06 Xe-133m 5.9E-05 0.0 5.9E-05 Xe-133 2.4E-03 0.0 2.4E-03 Xe-135m 2.2E-05 0.0 2.2E-05 Xe-135 3.0E-04 0.0 3.0E-04 Xe-137 1.6E-05 0.0 1.6E-05 Xe-138 7.6E-05 0.0 7.6E-05 I-131 2.9E-04 7.0E-03 7.3E-03 I-133 4.2E-04 9.5E-03 9.9E-03 Co-58 0.0 Co-60 0.0 Mn-54 0.0 Fe-59 0.0 Sr-89 0.0 Sr-90 0.0 Cs-134 0.0 Cs-137 0.0 C-14 0.0 Ar-41 0.0 H-3 0.0 Page 1 of 2 Rev. 30
Rev. 30 A. CURIES PER YEAR RELEASED Containment Building Auxiliary Building Nuclide (Ci/yr) (Ci/yr) (1) Total (Ci/yr)
Kr-83m 1.2E-03 9.5E+00 9.5E+00 Kr-85m 3.6E-01 3.6E+01 3.6E+01 Kr-85 2.3E+01 7.3E-01 2.4E+01 Kr-87 1.4E-04 2.6E+01 2.6E+01 Kr-88 1.0E-01 7.2E+01 7.2E+01 Kr-89 0.0 2.2E+00 2.2E+00 Xe-131m 1.7E+00 2.4E-01 1.9E+00 Xe-133m 1.5E+01 1.3E+01 2.8E+01 Xe-133 1.7E+03 5.6E+02 2.3E+03 Xe-135m 0.0 2.4E+01 2.4E+01 Xe-135 7.9E+00 1.1E+02 1.2E+02 Xe-137 0.0 3.5E+00 3.5E+00 Xe-138 0.0 1.3E+01 1.3E+01 I-131 2.6E-05 4.2E-01 4.2E-01 I-133 2.7E-06 6.6E-01 6.6E-01 Co-58 3.8E-03 6.0E-02 6.4E-02 Co-60 1.7E-03 2.7E-02 2.9E-02 Mn-54 1.1E-03 1.8E-02 1.9E-02 Fe-59 3.8E-04 6.0E-03 6.4E-03 Sr-89 9.9E-04 1.5E-02 1.6E-02 Sr-90 2.6E-04 3.4E-03 3.7E-03 Cs-34 1.3E-02 2.2E-01 2.3E-01 Cs-137 1.7E-01 2.6+00 2.8E+00 C-14 1.0E+00 0.0 1.0E+00 Ar-41 2.5E+01 0.0 2.5E+01 H-3 4.6E+02 2.1E+03 2.6E+03 Page 1 of 4 Rev. 30
CONCENTRATION AND FRACTION OF MPC (2) AT SITE BOUNDARY DUE TO RELEASES FROM CONTAINMENT PURGE (3) VIA VENTILATION VENT:
Concentration MPC Value Nuclide (Ci/cc) (Ci/cc) Fraction of MPC Kr-83m 1.02E-15 (4) 3.0E-08 3.40E-08 Kr-85m 3.06E-13 1.0E-07 3.06E-06 Kr-85 1.95E-11 3.0E-07 6.52E-05 Kr-87 1.19E-16 2.0E-08 5.95E-09 Kr-88 8.50E-14 2.0E-08 4.25E-06 Kr-89 0.0 3.0E-08 0.0 Xe-131m 1.44E-12 4.0E-07 3.61E-06 Xe-133m 1.27E-11 3.0E-07 4.25E-05 Xe-133 1.44E-9 3.0E-07 4.82E-03 Xe-135m 0.0 3.0E-08 0.0 Xe-135 6.71E-12 1.0E-07 6.71E-05 Xe-137 0.0 3.0E-08 0.0 Xe-138 0.0 3.0E-08 0.0 I-131 2.21E-17 1.0E-10 2.21E-07 I-133 2.29E-18 4.0E-10 5.74E-09 Co-58 3.23E-15 2.0E-09 1.61E-06 Co-60 1.44E-15 3.0E-10 4.82E-06 Mn-54 9.35E-16 1.0E-09 9.35E-07 Fe-59 3.23E-16 2.0E-09 1.61E-07 Sr-89 8.41E-16 3.0E-10 2.80E-06 Sr-90 2.21E-16 3.0E-11 7.37E-06 Cs-134 1.10E-14 4.0E-10 2.76E-05 Cs-137 1.44E-13 5.0E-10 2.89E-04 C-14 8.50E-13 1.0E-07 8.50E-06 Ar-41 2.12E-11 4.0E-08 5.31E-04 H-3 3.91E-10 2.0E-07 1.95E-03 Page 2 of 4 Rev. 30
C. CONCENTRATION AND FRACTION OF MPC AT SITE BOUNDARY DUE TO RELEASES FROM AUXILAIRY BUILDING (1) , (5) VIA VENTILATION VENT:
Concentration MPC Value Nuclide (Ci/cc) (Ci/cc) Fraction of MPC Kr-83m 1.20E-12 3.0E-08 4.01E-05 Kr-85m 4.55E-12 1.0E-07 4.55E-05 Kr-85 9.24E-14 3.0E-07 3.08E-07 Kr-87 3.29E-12 2.0E-08 1.64E-04 Kr-88 9.11E-12 2.0E-08 4.55E-04 Kr-89 2.78E-13 3.0E-08 9.28E-06 Xe-131m 3.04E-14 4.0E-07 7.59E-08 Xe-133m 1.64E-12 3.0E-07 5.48E-06 Xe-133 7.09E-11 3.0E-07 2.36E-04 Xe-135m 3.04E-12 3.0E-08 1.01E-04 Xe-135 1.39E-11 1.0E-07 1.39E-04 Xe-137 4.43E-13 3.0E-08 1.48E-05 Xe-138 1.64E-12 3.0E-08 5.48E-05 I-131 5.31E-14 1.0E-10 5.31E-04 I-133 8.35E-14 4.0E-10 2.09E-04 Co-58 7.59E-15 2.0E-09 3.80E-06 Co-60 3.42E-15 3.0E-10 1.14E-05 Mn-54 2.28E-15 1.0E-09 2.28E-06 Fe-59 7.59E-16 2.0E-09 3.80E-07 Sr-89 1.90E-15 3.0E-10 6.33E-06 Sr-90 4.30E-16 3.0E-11 1.43E-05 Cs-134 2.78E-14 4.0E-10 6.96E-05 Cs-137 3.29E-13 5.0E-10 6.58E-04 C-14 0.0 1.0E-07 0.0 Ar-41 0.0 4.0E-08 0.0 H-3 2.66E-10 2.0E-07 1.33E-03 Page 3 of 4 Rev. 30
Rev. 30 A. CURIES PER YEAR RELEASED Radioactive Main Condenser / Air Gaseous Waste Nuclide Ejector (Ci/yr) System (Ci/yr) Total (Ci/yr)
Kr-83m 8.2E+01 0.0 8.2E+01 Kr-85m 3.2E+02 0.0 3.2E+02 Kr-85 6.3E+00 4.0E+03 4.0E+03 Kr-87 2.2E+02 0.0 2.2E+02 Kr-88 6.2E+02 0.0 6.2E+02 Kr-89 1.9E+01 0.0 1.9E+01 Xe-131m 2.1E+00 3.5E-01 2.5E+00 Xe-133m 1.1E+02 0.0 1.1E+02 Xe-133 4.8E+03 2.4E-02 4.8E+03 Xe-135m 2.1E+02 0.0 2.1E+02 Xe-135 9.3E+02 0.0 9.3E+02 Xe-137 3.0E+01 0.0 3.0E+01 Xe-138 1.1E+02 0.0 1.1E+02 I-131 2.3E+00 0.0 2.3E+00 I-133 3.7E+00 0.0 3.7E+00 Co-58 0.0 0.0 0.0 Co-60 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0 Fe-59 0.0 0.0 0.0 Sr-89 0.0 0.0 0.0 Sr-90 0.0 0.0 0.0 Cs-134 0.0 0.0 0.0 Cs-137 0.0 0.0 0.0 C-14 0.0 7.0E+00 7.0E+00 Ar-41 0.0 0.0 0.0 Page 1 of 3 Rev. 30
Radioactive Main Condenser / Air Gaseous Waste Nuclide Ejector (Ci/yr) System (Ci/yr) Total (Ci/yr)
H-3 0.0 0.0 0.0 CONCENTRATIONS AND FRACTION OF MPC (1) AT SITE BOUNDARY DUE TO RELEASES VIA MILLSTONE STACK (2):
MPC Value Nuclide Concentration (Ci/cc) (Ci/cc) Fraction of MPC Kr-83m 3.06E-14 (4) 3.0E-08 1.02E-06 Kr-85m 1.20E-13 1.0E-07 1.20E-06 K-85 1.50E-12 3.0E-07 4.99E-06 Kr-87 8.23E-14 2.0E-08 4.12E-06 Kr-88 2.32E-13 2.0E-08 1.16E-05 Kr-89 7.11E-15 3.0E-08 2.37E-07 Xe-131m 9.35E-16 4.0E-07 2.34E-09 Xe-133m 4.12E-14 3.0E-07 1.37E-07 Xe-133 1.80E-12 3.0E-07 5.99E-06 Xe-135m 7.86E-14 3.0E-08 2.62E-06 Xe-135 3.48E-13 1.0E-07 3.48E-06 Xe-137 1.12E-14 3.0E-08 3.74E-07 Xe-138 4.12E-14 3.0E-08 1.37E-06 I-131 8.61E-16 1.0E-10 8.61E-06 I-133 1.38E-15 4.0E-10 3.46E-06 Co-58 0.0 2.0E-09 0.0 Co-60 0.0 3.0E-10 0.0 Mn-54 0.0 1.0E-09 0.0 Fe-59 0.0 2.0E-09 0.0 Sr-89 0.0 3.0E-10 0.0 Sr-90 0.0 3.0E-11 0.0 Cs-134 0.0 4.0E-10 0.0 Page 2 of 3 Rev. 30
tained to preserve original license Rev. 30
5.1E-01 2.2E-02 1.0E-01 3.2E-03 1.2E-02 1 9.0E-01 00 1.4E+00 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Rev. 30
7.15E-10 3.84E-07 1.04E-06 2.24E-08 1.76E-10 1.28E-08 5.44E-07 2.35E-07 3.20E-07 3.42E-08 1.28E-07 1.18E-04 4.48E-05 0.0 0.0 0.0 0.0 Rev. 30
ue (Ci/cc) Fraction of MPC 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1.33E-02 4.99E-03 0.0 Rev. 30
able II, Column I. When releasing imits are based on the values in ry 1, 1994 (Reference 11.3-1.)
1.01 x 10-5 sec/m3, which is the tive concentrations at the Site s 4.84 x 10-5 sec/m3, which is the tive concentrations at the Site tained to preserve original license Rev. 30
2.5E+00 7.1E+03 1.1E-01 2.3E+02 5.0E-01 1.1E+03 1.6E-02 3.4E+01 5.8E-02 1.2E+02 2.8E-02 3.6E+00 4.2E-02 5.8E+00 0.0 6.4E-02 0.0 2.9E-02 0.0 1.9E-02 0.0 6.4E-03 0.0 1.6E-02 0.0 3.7E-03 0.0 2.3E-01 0.0 2.8E+00 0.0 8.0E+00 0.0 2.5E+01 Rev. 30
Rev. 30 3.58E-06 3.82E-07 1.39E-06 2.01E-04 7.52E-05 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Rev. 30
2.63E-06 1.06E-04 3.58E-06 2.13E-04 3.82E-07 1.56E-05 1.39E-06 5.77E-05 2.01E-04 1.41E-02 7.52E-05 5.32E-03 0.0 5.41E-06 0.0 1.62E-05 0.0 3.22E-06 0.0 5.41E-07 0.0 9.13E-06 0.0 2.17E-05 0.0 9.72E-05 0.0 9.47E-04 0.0 8.53E-06 0.0 5.31E-04 0.0 3.28E-03 Rev. 30
Rev. 30 ABLE 11.3-12 ASSUMPTIONS USED FOR THE PROCESS GAS CHARCOAL BED ADSORBER BYPASS ANALYSIS (1)
Design Release tdown flow to degasifier (gpm) 75 actor coolant activity Table 11.1-2 arcoal bed adsorber holdup time Kr (days) 6.1 Xe (days) 142 ction of noble gas released from bed Table 11.3-13 ration of release (min.) 60 TE:
The bypass of the process gas charcoal bed adsorbers releases the gaseous activity discharged from the degasifier for 60 minutes and a fraction of the activity on the beds.
Page 1 of 1 Rev. 30
TABLE 11.3-13 RADIOISOTOPE RELEASES FROM THE PROCESS GAS CHARCOAL BED ADSORBER AND ASSOCIATED PIPING Design Discharge from Inventory in the Fraction the degasifier Charcoal Bed Released from Total Release Isotope (Ci) (Ci) Bed (Ci)
-83m 7.65E+00 (1) 2.06E+01 1.000 2.83E+01
-85m 2.93E+01 1.86E+01 0.998 2.15E+02
-85 5.88E-01 8.64E+01 0.109 1.00E+01
-87 2.11E+01 3.84E+01 1.000 5.95E+01
-88 5.76E+01 2.32E+02 1.000 2.90E+02
-89 1.80E+00 1.38E-01 1.000 1.94E+00
-131m 1.91E-01 7.91E+01 0.095 7.69E+00
-133m 1.04E+0.1 8.15E+02 0.410 3.44E+02
-133 4.48E+02 8.27E+04 0.202 1.72E+04
-135m 1.96E+01 7.39E+00 1.000 2.70E+01
-135 8.74E+01 1.16E+03 1.000 1.25E+03
-137 2.84E+00 2.62E-01 1.000 3.10E+00
-138 1.02E+01 3.48E+00 1.000 1.37E+01 TE:
7.65E+00 = 7.65 x 100 Page 1 of 1 Rev. 30
figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
Rev. 30
radioactive solid waste system is designed to collect, hold, process, dewater or solidify, kage, handle, and temporarily store radioactive materials prior to their shipment offsite and mate disposal.
radioactivity values provided in this section are the design basis values used for the design of Solid Waste System. As such, they are considered historical and not subject to future updating.
information is retained to avoid loss of the original design bases. Volumes and radioactivity tent, including specific nuclide percentages of actual shipments can be found in the annual oactive effluent release reports as submitted to the NRC.
.1 DESIGN BASES radioactive solid waste system is designed in accordance with the following criteria.
- 1. The system design parameters are based 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 required). 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. Dewatering equipment is portable 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 capacity 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 applicable NRC, e.g., 10 CFR Part 61, and Department of Transportation regulations, 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.
11.4-1 Rev. 30
11-3.
- 5. The permanent plant system contains provisions to reduce leakage and facilitate operations and maintenance in accordance with the provisions 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 personnel in accordance with the radiation protection design considerations in Section 12.1.
- 8. Radiation protection personnel conducting surveys use portable 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 nonnuclear safety (NNS).
- 10. The portion of the solid waste buildings foundation and adjacent walls up to a height sufficient to contain the liquid inventory 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 radioactive 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 volume of selected input streams to minimize packaged quantities for transport and disposal.
.2 SYSTEM DESCRIPTION radioactive solid waste system is shown on Figure 11.4-1.
layout of the solid waste building, including packaging, storage, and shipping areas, is shown igure 3.8-74.
11.4-2 Rev. 30
.2.1 System Inputs erials handled as solid wastes may include any of the following: concentrated waste solutions m the waste evaporator, concentrated boric acid discarded from the boron evaporator in the on recovery system (Section 9.3.5), spent resin from radioactive process demineralizers and hangers, spent filter cartridges, and miscellaneous sludges.
ure 11.4-2 is a flow chart of the expected quantities and activity levels of the radioactive solid te generated by the plant. Gross activities and weights or volumes for radioactive solid waste rces are also given on Figure 11.4-2. Figure 11.4-3 provides the design volumes and activity ls of radioactive solid wastes and waste sources.
.2.1.1 Spent Resins ure 11.4-2 provides the estimated volumes of spent demineralizer and ion exchanger resins per
- r. These resins come from a variety of different services, and the total volume estimated is ed on the individual resin bed volumes and the expected frequency of replacement (Table
-1).
.2.1.2 Waste Evaporator Bottoms porator bottoms from the waste evaporator in the radioactive liquid waste system may be sferred to the solid waste system. The estimated volume of bottoms solution to be shipped off is given on Figure 11.4-2. The calculated activity of these bottoms, also shown on Figure
-2, is based on the input of the radioactive liquid waste system (Section 11.2).
.2.1.3 Regenerant Chemical Evaporator Bottoms (Removed From Service)
.2.1.4 Boron Evaporator Bottoms boron evaporator in the boron recovery system processes reactor coolant letdown to the on recovery system for separation of boric acid and water. The estimated volume given on ure 11.4-2 for the boron evaporator bottoms is a 1 year average of the contents requiring cessing for eventual off site disposal. The activity shown for these bottoms, given on Figure
-2, is based on the expected performance of the boron recovery system (Section 9.3.5). Boric may be either recycled to the plant or transferred to the solid waste system depending upon rational considerations.
.2.1.5 Miscellaneous Radioactive Solid Wastes estimated that approximately 4,000 cubic feet per year of additional waste requiring disposal rocessed as radioactive solid waste. This volume of spent filters, contaminated cloths, and 11.4-3 Rev. 30
s category includes waste charcoal absorber media from filtration units. Charcoal is removed n external vacuum system outlined in CVI Topical Report No. CVI-TR-7301 (February 5).
.2.2 Equipment Description radioactive solid waste system equipment is operated on a batch basis. Individual ponents are designed to support the systems rated capacity.
permanently installed system consists of a control panel, spent resin dewatering and hold s, and the piping, pumps, and process equipment modules required for transfer of wastes to a ping container. The solid waste equipment requires a minimum manual action and, in junction with the building layout, is designed to minimize occupational radiation exposures.
portable portion of the system is designed to permit filling and dewatering of the resin slurry he shipping container. This portion consists of a fill/dewatering head, a dewatering pump, a trol panel, and other process equipment for drying the resin to suitable levels for shipment off pping containers are approved for the process for which they are to be used.
en slurries are being processed, the fill head is lifted by the overhead bridge crane and lowered o the shipping container having the integral dewatering line. Connected to the fill head is a te fill line, vent line, decontamination lines, and a dewatering line that extends from the fill
- d. These lines are flexible, with flanged fittings to facilitate periodic replacement or ntenance. When the fill head is lowered onto the shipping container, it is aligned with the rnal dewatering line and secured prior to filling.
ic acid concentrated in the boron evaporator in the boron recovery system is typically cled. However, the capability exists to transfer waste concentrated from the waste evaporator he radioactive liquid waste system to the solid waste system.
ins sluiced from demineralizers and ion exchangers are stored in the spent resin hold tank. The ns are then slurried to the shipping container, where they are allowed to settle. Excess water is oved by the spent resin dewatering pump within the portable dewatering unit and transferred he spent resin dewatering tank.
olidification were desired, it would be performed by an approved vendor whose Process trol Program was reviewed and SORC approved. The solidification will be performed with a bile solidification system provided by the approved vendor.
nt filters, contaminated tools, and other incompressible contaminated solid wastes can be rted into a shipping container prior to filling. If solidification is required, a vendor will be zed. Containers and shields are handled by overhead bridge crane.
11.4-4 Rev. 30
systems are designed to prevent external contamination of containers by use of reliable tainer sealing, appropriate system flushing into the container, and necessary mechanical gn or instrumentation interlock signals that prevent overfilling of containers.
sealed disposable container is transported to an interim on-site storage facility or a disposal
. In general, the disposal container may hold spent resin, spent filters, and other mpressible waste in addition to evaporator bottoms.
mpressible dry solid waste (e.g., contaminated clothing, wipeup toweling) are collected and warded for processing.
.2.2.1 Boron, Waste Evaporator Bottoms centrated liquids from the evaporators may be pumped to the waste bottoms tank for holdup l they are to be solidified. The tank is insulated, heat traced, and is recirculated to prevent tallization, stratification, settling, and accumulation of undissolved solids.
amount of waste liquid allowed per container is determined by prior analysis of the waste.
iation levels are also monitored while filling the container to ensure DOT shipping limits are exceeded. After the container is filled and solidified, it is sealed and stored. Prior to shipment, shielded, as necessary.
pments are made in accordance with NRC regulations 10 CFR 20, 10 CFR 50, and 10 CFR and Department of Transportation regulations 49 CFR 171 through 178. The shipping tainers are stored temporarily in the solidification area (Section 11.4.2.5) or an interim on-site age facility prior to off-site disposal.
.2.2.2 Spent Resin Handling in in a demineralizer or ion exchanger is considered spent when the decontamination factor s below a predetermined value, when the demineralizer or ion exchanger surface dose roaches a predetermined limit, or when the resin bed pressure drop becomes excessive. The ineralizer or ion exchanger is then isolated. Water from the spent resin dewatering tank is d to flush the spent resin into the spent resin hold tank utilizing the spent resin transfer pump.
flush water passes out of the spent resin hold tank, through a screened element to prevent any n carryover, and is recirculated. When the resin is to be packaged for disposal, it is pumped as sin-water slurry to the shipping container in the processing area. Excess water is dewatered m the shipping container, via the portable dewatering unit, and returned to the spent resin atering tank. Radiation levels are monitored to ensure that DOT shipping limits are not eeded. After dewatering, the container is sealed, labeled, and temporarily stored until shipped site. Prior to shipment the container is shielded as necessary.
11.4-5 Rev. 30
tridge filter elements are removed from service when the surface dose on the filter housing hes a predetermined level or when the element pressure drop becomes excessive. The ration is carried out using remote handling equipment and a filter removal shield, when uired. Radiation protection personnel conduct surveys during filter changeouts as required, g portable radiation detectors and take action as appropriate in accordance with ALARA delines.
.2.2.4 Incompressible Waste Handling taminated metallic materials and solid objects are placed in disposable shipping containers.
.2.2.5 Waste Compaction Operation taminated compressible materials are temporarily stored in suitably labeled containers in erent plant locations. These materials are then transported for processing or packaging.
.2.3 Expected Volumes le 11.4-1 presents a listing of the expected volumes of spent resins from various sources ring the radioactive solid waste system. (Note that resin from the condensate polishing ineralizers indicated on Table 11.4-1 is processed by Millstone Unit 2 - See Figure 10.4-5.)
ure 11.4-2 presents gross activities and weights or volumes for radioactive solid waste sources.
.2.4 Packaging ed on the gross activities supplied on Figure 11.4-2 and DOT Low Specific Activity Limits, tainer activity varies between negligible for most compressible or compacted wastes to less 300 Ci/cc for reactor water purification demineralizer resins. The specific radionuclide tent of the solid wastes is available, as discussed in Section 11.4.
filling of containers and the storage of radioactive solid wastes conform with 10 CFR 20 and CFR 50 requirements. Packages meet shipping regulations of 49 CFR 171-178 and 10 CFR 71 pplicable.
mplete solidification and absence of free liquid is ensured by the implementation of a process trol program and preoperational testing.
.2.5 Temporary On-site Storage Facilities processing area is located on the ground floor of the solid waste building. Wastes are kaged to allow for shipment after processing. Therefore, no decay time is assumed in the given mate of package contents and activity levels.
age capacity and storage time are based on anticipated operational factors.
11.4-6 Rev. 30
- 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 structures meet the criteria set forth in Section 3.5.1.4.
- 2. In addition to being used for the sorting, processing, loading and shipping of radioactive materials, the MRRF may be used to store dry activated waste. The capacity of the MRRF is dependent upon:
- a. the waste generated from unit/site activities; and
- b. waste volume reduction techniques employed.
- 3. The Unit 2/Unit 3 Condensate Polishing Facility (CPF) waste processing area may be used to store mixed waste (radioactive and hazardous combined) material.
.2.6 Shipment shipment of radioactive solid waste conforms with 10 CFR 20, 10 CFR 50, and 10 CFR 61 uirements and 10 CFR 71 and 49 CFR 171 through 178. Solid waste is transferred either ctly to a licensed disposal contractor or to a common carrier for delivery to a licensed burial
, or secondary processor as appropriate.
le 11.4-4 summarizes the annual number of shipments and shipped containers for the expected design cases.
.2.7 Protection Against Uncontrolled Releases tection against uncontrolled releases of radioactive material from the radioactive solid waste em is achieved through the use of alarms, interlocks, and a retaining structure.
spent resin dewatering tank, evaporator bottoms tank, spent resin transfer pump, and spent n recycle pump are located in curbed cubicles where any leakage is retained. The walls and rs are suitably finished to facilitate decontamination. The disposal waste shipping container the resin fill and dewater head are located on the 24 foot 6 inch elevation of the waste dification building. In the event of spillage of radioactive liquid in this area, the liquid is ected by a network of floor drains. The floor is pitched toward the floor drains. The drains are d to the waste disposal building sumps (FSAR Figure 3.8-74) which are collected by the ted drains system (FSAR Section 9.3.3) and forwarded to the radioactive liquid waste system AR Section 11.2) for processing.
11.4-7 Rev. 30
ducted.
lth physics personnel, equipped with portable radiation detectors, are present at the fill area to ure that the radiation levels are within the design levels.
table dewatering equipment receives solid waste from the plant and transfers excess water k to the plant.
single worst operator error or equipment failure would result in the spillage of spent resin
/or evaporator bottoms tank contents. The drainage system in the waste solidification building esigned to handle such an event.
11.4-8 Rev. 30
Frequency of Number of Replacement Demineralizers Beds Volume ft3/ bed (beds/yr) el pool demineralizer 1 15 1 sium removal ion exchangers 2 35 4 ron demineralizer 2 35 2 ste demineralizer 2 35 2 ndensate polishing demineralizers (1) 8 196 1 xed bed demineralizers 2 30 6 tion bed demineralizers 1 20 2 ermal regeneration demineralizers 5 74 2 TE:
The condensate polishing demineralizer resin will be processed by Millstone 2.
Page 1 of 1 Rev. 30
TABLE 11.4-2 RADIOACTIVE SOLID WASTE SYSTEM COMPONENT DATA ent Resin Hold Tank Parameters Number 1 Capacity (gal) 2,000 Operating pressure 50 Design temperature (°F) 200 Material of construction SS ent Resin Dewatering Tank Number 1 Capacity (gal) 500 Operating pressure Atmospheric Design temperature (°F) 200 Material of construction SS or FRP nder Storage Tank (not used)
Number 1 Capacity (gal) 6,000 Operating pressure Atmospheric Design temperature (°F) 100 Material of construction SS ent Resin Recycle Pump Number 1 Capacity (gpm) 125 Operating pressure (psig) 75 Design temperature (°F) 100 Material of construction SS Page 1 of 2 Rev. 30
ent Resin Transfer Pump Number 1 Capacity (gpm) 120 Operating pressure (psig) 146 Design temperature (°F) 100 Metal of construction SS nder Pump (not used)
Number 1 Capacity (gpm) 65 Design temperature (°F) 150 Material of construction Cast iron ent Resin Transfer Pump Filter Parameters Number 1 Capacity (gpm) 150 Design pressure (psig) 150 Design temperature (°F) 200 Material of construction SS TE:
Designed in accordance with ASME VIII, Division I.
Page 2 of 2 Rev. 30
Page 1 of 1 Rev. 30 Type of Waste and Packaging Expected Design lidified / Dewatered Wastes (3) ntainers / shipments 13 containers/13 shipments 33 containers / 33 shipments scellaneous ompressible Solids cu ft crates / shipments 4 crates / 1 shipment 8 crates / 1 shipment TES:
Condensate polishing spent resins are designed to be processed by MP2.
Miscellaneous compressible solids for Millstone 3 are processed at the MRRF.
If evaporators are used to process waste instead of resins, then expected shipments will be 56 containers/56 shipments.
Page 1 of 1 Rev. 30
figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
Rev. 30
.1 DESIGN BASES process, effluent, and airborne radiation monitoring system (RMS) is designed in accordance h NRC General Design Criterion (GDC) 64 (Section 3.1.2.64) and American National ndards Institute (ANSI) N13.1-1969.
mal and potential paths for release of radioactive materials, during both normal operation and cipated operational occurrences, are continuously monitored to ensure compliance with the uirements of 10 CFR 20, 10 CFR 50, and the guidelines of Regulatory Guide 1.21 ction 1.8). Sections 11.5.2 and 11.5.3 describe the design features provided to ensure ement with Regulatory Guide 1.21. Potential pathways for release of radioactive materials ng accident conditions are continuously monitored to ensure agreement with the guidelines of ulatory Guide 1.97 (Sections 1.8, 7.5).
tinuous monitoring means that the system operates essentially uninterrupted for extended ods during normal plant operation, but does not preclude periods when individual monitors be out of service for maintenance, repair, calibration, etc.
reactor coolant system (RCS) is monitored continuously for gross activity level. Maintaining RCS activity within acceptable levels ensures that the activity levels in the normally oactive auxiliary systems are at acceptable levels. Nonradioactive systems which may become taminated by leaks from radioactive systems are also monitored continuously. This monitoring ures that no conditions develop that are potentially hazardous to the operating personnel or to general public.
he event of an accident releasing radionuclides, the process, effluent, and airborne RMS and area RMS (Section 12.3.4), provide information on the concentration and dispersion of oactivity throughout the plant. This enables operating personnel to evaluate the severity and to gate the consequences of the accident.
following automatic actions are initiated by the process and effluent RMS to mitigate both the sequences of postulated accidents and excessive releases during normal operations.
- 1. Containment purge (Section 9.4.7) is automatically terminated in the event of a high radiation alarm.
- 2. Liquid effluent discharges from the turbine 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 release is terminated automatically upon a high radionuclide concentration alarm (Section 11.3.3).
11.5-1 Rev. 30
- 5. Control building ventilation intake is terminated 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 diverted to the regenerant evaporator feed tank (removed from service) upon a high radionuclide concentration alarm (Section 11.5.2.3.8).
- 7. Following a high radionuclide concentration alarm from the auxiliary condensate monitor, effluent from the auxiliary condensate flash tank is diverted from the auxiliary condensate feed tank to the auxiliary building sump.
- 8. A high radionuclide concentration alarm from either hydrogen recombiner ventilation monitor automatically shuts down ventilation from its cubicle (Section 9.4.11).
- 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 automatically terminated on high radionuclide concentration alarm on the steam generator blowdown monitor.
.2 SYSTEM DESCRIPTION
.2.1 Instrumentation process, effluent, and airborne RMS consists of separate and independent monitors, which rporate the following features:
- 1. Each liquid monitor has one detector. Each particulate and gas monitor has one gaseous detector and either a fixed paper 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 dedicated 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.
11.5-2 Rev. 30
applicable guidelines are those in Appendix B of 10 CFR 20.
dedicated microprocessor sends an alarm message to the RMS computer system in the event ocal power loss, loss of sample flow, filter failure, or other conditions specific to a given nitor.
ans are provided to purge each fixed volume sample chamber in the system with clean fluid to imize contamination of the chambers. Sample lines and all surfaces of each sampler exposed he sample are stainless steel and are run to minimize fixed contamination.
rability of many detectors can be checked using an installed check source remotely activated m the control room. Tests and calibrations of the radiation monitors are performed at specified rvals. For ease in calibration and maintenance, each detector is located in an easily accessible and is provided with a local readout device or connects to a plug-in portable indication and trol (PIC) module.
monitors and microprocessors are powered by 120 V AC buses. Monitor skids designed in ordance with Regulatory Guide 1.97 and safety-related monitors, such as the control building t and ventilation monitors, use the safety-related Class 1E buses (Section 8.3.1.1.2). All other nitors use regular 120 V AC buses (Section 8.3.1.1.1).
itionally, all monitors with pumps require 480 V AC power. Safety-related monitors with ps use safety-related Class 1E buses, and all others use regular 480 V AC buses.
6 and Fission Product Main Steam Line Monitor alarms are processed by the plant process puter. All other alarms are displayed visually on the RMS workstations in the control room.
se include equipment malfunctions, alarm/high radiation levels, conductivity, sample flow, sample temperature. Local and control room annunciation is provided for high radiation ms, as well as an interface to the main plant annunciator panel. All alarms can be nowledged locally or in the control room.
itionally, those variables designated Class 1E are also displayed on the Class 1E control room els, as required by Regulatory Guide 1.97. A digital display and control module is provided each Class 1E monitor, as well as a dedicated two-pen recorder. In order to record this data at RMS computer along with that from the non-Class 1E devices, these cabinets are connected to RMS computer via electronic isolators.
cesses Monitored le 11.5-1 gives numbers and location of gaseous process and effluent radiation monitors, and le 11.5-2 gives numbers and locations of liquid process and effluent radiation monitors.
11.5-3 Rev. 30
h detector has sufficient shielding to ensure that the required sensitivity is achieved for the imum expected background radiation level at the detector location.
monitors measure gross concentrations. The output is typically measured in microcuries per ic centimeter (Ci/cc), with a minimum range of five decades.
.2.2 Process and Effluent Monitors
.2.2.1 Ventilation Vent MonitorsNormal Range h skid has its own microprocessor. The two microprocessors are linked by a dedicated rface. When the normal range skid senses the upper level of its range, control is automatically sed to the high range skid. When the radionuclide concentration returns to normal, the reverse pens. Both skids use the same isokinetic nozzle. During high range operation, the normal ge gas sampler is automatically isolated.
sample line and all surfaces of the sampler exposed to the sample are stainless steel. A sample p provides isokinetic sample flow via a flow control valve which modulates proportional to flow signal from the process flow transmitter. Isokinetic sample flow is maintained through normal range skid even after the detectors are isolated during high range operation.
ventilation vent normal range monitor takes a continuous effluent sample from an isokinetic zle common to this normal and high range monitor and draws the gas sample through a iculate and charcoal filter and then to a gas sampling assembly where activity is measured by ta scintillation detector. Lead shielding is provided to reduce the background radiation to a l that minimizes interference with the detector sensitivity.
er leaving the gas sampler, the sample is returned to the duct downstream of the sample point.
urge system for flushing the sample volume is also provided.
ventilation vent normal range monitor detector output is transmitted via the dedicated roprocessor to the RMS computer system Class 1E panels located in the control room. Here, activity level is displayed and also recorded on a two-pen strip chart recorder. Activity levels also digitally displayed locally at the microprocessor location. Alarm conditions, such as high vity or monitor failure, are indicated by audible and visible alarms in the control room and by ble alarm locally.
.2.2.2 Ventilation Vent Monitor-High Range vent sample point is common to that of the ventilation vent normal range monitor and is ted downstream of the last point where radioactivity is introduced to the flow stream prior to elease via the ventilation vent stack. Sample flow through the high range monitor is ntained constant via a set hand control valve. Isokinetic flow through this monitor depends on tinued operation of the normal range monitor pump to maintain isokinetic flow conditions 11.5-4 Rev. 30
nitored by a Geiger-Mueller detector. An alarm will automatically direct flow to another filter mbly at a level which still allows analysis of the filter in a laboratory GeLi detector. An quate amount of lead shielding around the detector assembly reduces the background radiation level that minimizes interference with the detector sensitivity. After passing through the filter er, the sample passes through an inline easily removable charcoal filter cartridge arrangement then into a fixed and shielded volume where it is monitored by mid-range and high range nitors.
.2.2.3 Hydrogenated Vent Monitor hydrogenated vent monitor continuously monitors the effluent from the gaseous waste system nstream of the charcoal decay beds and prior to their release via the ventilation vent stack.
gas detector, a beta scintillator, is located in the well of an inline gas sampler. Four pi lead lding is provided in order to reduce the background radiation to a level that minimizes rference with the detector sensitivity.
s monitors output is transmitted, indicated, recorded, and alarmed in a manner similar to that ll non-Class 1E process monitors. When high activity is present, gas flow to the ventilation t stack is isolated. The sample line and all surfaces of the sampler exposed to the sample line stainless steel. Due to the presence of free hydrogen in this effluent, this monitor is purged h nitrogen.
.2.2.4 Containment Fuel Drop Monitors gross radionuclide concentration entering the containment purge air vent is monitored by two undant containment fuel drop monitors. These monitors, each consist of an ion chamber ctor measuring dose rates just above the surface of the refueling canal. These are safety-ted, Class 1E, monitors. Due to high radiation levels inside the containment, their roprocessors are located in the auxiliary building.
outputs of these monitors are transmitted, indicated, recorded, and alarmed in a manner ilar to that of the ventilation vent monitor. A high activity indication from either of these nitors automatically isolates containment purge (Section 9.4.7.2) based on the assumption that high dose rates are due to high airborne activity.
.2.2.5 Supplementary Leak Collection and Release System Monitor supplementary leak collection and release system (SLCRS) normal range and high range nitors are identical to the ventilation vent monitors. Using an isokinetic nozzle, the monitor hdraws a sample from the SLCRS exhaust vent prior to its discharge to the Millstone stack.
SLCRS collects, filters, and releases the leakage from the containment enclosure building contiguous areas to the atmosphere following a design basis accident (DBA) (Section 6.2.3).
ing normal operation, this detector monitors the discharge point for the containment vacuum 11.5-5 Rev. 30
rs and prior to the exhaust discharge point. The SLCRS monitor functions as a final effluent nitor.
s monitors output is transmitted, indicated, recorded, and alarmed in a manner similar to that he ventilation vent monitors. An alarm from this monitor warns the operator of a potential blem so that he may take appropriate action, which may include the manual transfer of flow to andby SLCRS filter bank.
.2.2.6 Condenser Air Ejector Monitor condenser air ejector monitor continuously analyzes the gaseous effluents from the condenser ejector discharge (Section 10.4.2). A gamma scintillation detector is inserted into an in-line air
- l. The detector is shielded with lead to reduce the background radiation to a level which imizes interference with the detector sensitivity.
monitors output is transmitted, indicated, recorded, and alarmed in a manner similar to that he hydrogenated vent monitor. Activity readings are indicative of primary-to-secondary age.
.2.2.7 Control Building Inlet Ventilation Monitors two control building inlet monitors are located in the upper level of the control building in the t plenum. Here, they continuously analyze the ventilation being supplied to the control ding by measuring gross activity. Each monitor consists of a beta scintillation detector. These Class 1E, safety-related monitors.
detector outputs are transmitted, indicated, and alarmed in a manner similar to that of the tilation vent monitor. A high activity condition initiates a control building isolation signal ch will isolate the control building atmosphere from the outside atmosphere (Section 6.4).
.2.2.8 Hydrogen Recombiner Ventilation Monitors owing a loss-of-coolant accident (LOCA), the hydrogen recombiners were originally installed e used to eliminate the free hydrogen in the containment atmosphere. However, the tainment atmosphere following a LOCA also contains large amounts of gaseous onuclides. Should a leak develop in a recombiner, these radionuclides are collected and harged to the atmosphere via the hydrogen recombiner ventilation system (Section 9.4.11).
refore, one safety-related, post-accident, beta scintillation radiation detector is placed within exhaust duct of each ventilation system, and its dedicated microprocessor is located in the rogen recombiner control room. Upon a high radiation alarm, a signal from the monitor matically shuts down the respective hydrogen recombiner package system and activates ure of the supply and exhaust duct dampers securing the ventilation. Recombiner operators then manually start the other recombiner unit.
11.5-6 Rev. 30
.2.2.9 Normal Range Particulate and Gas Monitors ddition to the ventilation vent and SLCRS monitors, the RMS also contains 11 normal range iculate and gas monitors. They are of a single, basic skid design. The particulate channels loy either a remotely controlled, variable speed moving paper filter with an integral beta ctor, or a fixed paper disc filter for specific radionuclide analysis. The gas channel is a beta ctor. 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 operating personnel to locate the source of radionuclide leakage into the air of the relevant spaces. The areas sampled are in auxiliary building, the fuel building, and the waste disposal building.
These monitors are all designated non-safety related, and their outputs are indicated, recorded, and annunciated similarly to that of the condenser air ejector monitor.
- 2. The Engineered Safety Features Building Heating and Ventilation System (Section 9.4.5)
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 the fixed paper filter disc design for a radionuclide analysis in compliance with 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 the 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 monitor provides an indication of radionuclide concentration within the control building. It provides greater sensitivity for detecting airborne activity in the control room than the control building inlet monitors.
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.
11.5-7 Rev. 30
This monitor continually withdraws a sample of the containment atmosphere, analyzes it, and returns it to the containment, using dedicated sample lines. These lines are heat traced to prevent condensation and slope backwards toward the containment structure. This prevents the sample lines from becoming sources of radiation. The sample lines contain valves to isolate this monitor upon a containment isolation signal. The removable charcoal filter is not used, Iodine samples are collected using a temporary sampler.
.2.2.10 Main Steam Relief Line Monitors se four ion chamber monitors, located in the main steam valve building, measure the gross onuclide concentration in the main steam lines in the event of a steam generator tube failure.
y are used to meet the intent of Regulatory Guide 1.97 Rev. 2. Lead shielding is provided to uce the background radiation to a level which does not interfere with the detector sensitivity.
main steam relief line monitors output is transmitted via their dedicated microprocessors to RMS workstations located in the control room where the activity level is digitally displayed.
igh activity level is indicated by audible and visible alarms in the main control room.
.2.2.11 Turbine Driven Auxiliary Feedwater Pump Discharge Monitor alternate effluent path for radionuclides entrained in the main steam is monitored by this ctor located in the ESF building. This is an ion chamber gross detector used to meet the intent egulatory Guide 1.97, Rev. 2. Lead shielding is provided to reduce the background radiation level which does not interfere with the detector sensitivity.
turbine driven auxiliary feedwater pump discharge monitors output is transmitted via its icated microprocessor to the RMS workstations located in the control room where the activity igitally displayed. A high activity level is indicated by audible and visible alarms in the main trol room.
.2.2.12 Main Steam Line Monitor; N-16 and Fission Product monitor has four unshielded gamma scintillation detectors, one mounted on each main steam at the containment penetration in the main steam valve building. Each detector has two nnels, one to measure high energy N-16 activity and one to measure lower energy fission duct activity. This monitor inputs all eight channels and an equipment failure signal to the t process computer. The plant process computer records, displays, and provides main control rd annunciation for this monitor. This monitoring system is Non-QA, designed to support lstones Primary-to-Secondary leak rate program.
11.5-8 Rev. 30
le 11.5-2 lists the locations of liquid process monitors and the streams being monitored.
.2.3.1 Containment Recirculation Cooler Service Water Outlet Monitors ing the recirculation phase of emergency core cooling, the containment recirculation cooler ice water outlet monitors continuously measure the radionuclide concentration in the service er effluent from each pair of containment recirculation coolers (one monitor per pair of lers). These are Class 1E, on-line, gamma scintillation, gross detectors, placed atop the harge pipes just outside the ESF building. As the pipes are underground, only the detectors are mounted, with their dedicated microprocessors located in the fuel building.
containment recirculation cooler service water outlet monitor output is transmitted via the icated microprocessor to the RMS computer system Class 1E cabinets located in the control
- m. Here, the concentration is digitally displayed and also recorded on a strip chart recorder.
activity level is also digitally displayed locally at the microprocessor location. A high activity l is indicated by audible and visible alarms locally and in the main control room. These nitors warn of a leak into the service water system, within the containment recirculation lers.
.2.3.2 Liquid Waste Monitor liquid waste monitor continuously analyzes the liquid waste effluent discharge pipe ction 11.2) downstream of the last possible point of radioactive liquid addition. The detector mbly consists of a gamma scintillation detector inserted into the well of a four-pi, lead lded liquid sampler. All surfaces in contact with the liquid sample are of stainless steel. Lead lding is provided in order to reduce the background radiation to a level which does not rfere with the detector sensitivity.
detector output is transmitted, indicated, recorded, and alarmed in a manner similar to that of condenser air ejector monitor. A high activity situation initiates closure of a discharge valve, eby preventing the discharge of effluent to the environment in excess of dose limits. The high vity alarm and isolation function are based on a time averaged activity to verify consistently h levels.
.2.3.3 Steam Generator Blowdown Sample Monitor steam generator blowdown sample monitor analyzes the steam generator blowdown effluent ction 10.4.8) for radioactivity which would be indicative of primary-to-secondary leakage.
ples from each of the four steam generator bottoms are mixed in a common header. This mon sample is continuously monitored by a gamma scintillation detector inserted into the l of a four-pi, lead-shielded liquid sampler. All surfaces in contact with the liquid sample are tainless steel. Lead shielding is provided in order to reduce the background radiation to a level ch does not interfere with the detector sensitivity.
11.5-9 Rev. 30
matically isolate the steam generator blowdown.
.2.3.4 Auxiliary Condensate Monitor auxiliary condensate monitor continuously analyzes samples drawn from the discharge of the iliary condensate flash tank (Section 10.4.10). The detector assembly consists of a gamma tillation detector inserted into the well of a four pi, lead-shielded liquid sampler. All surfaces ontact with the liquid sample are of stainless steel. Lead shielding is provided in order to uce the background radiation to a level which does not interfere with the detector sensitivity.
inline conductivity element provides sample specific conductivity measurement, indication, alarm. A sample cooler reduces the sample temperature to 140°F or less, and inlet and outlet noid valves isolate the detector assembly and stop the sample pump automatically on high ple temperatures.
detector output is transmitted, indicated, recorded, and alarmed in a manner similar to that of condenser air ejector monitor. During normal operation, activities significantly above kground are indicative of a leak into the auxiliary steam system (Section 10.4.10) from one of systems containing radioactive fluids which exchange heat with the auxiliary steam system. A h radiation alarm will automatically divert flow to auxiliary building sumps.
.2.3.5 Turbine Building Floor Drains Monitor turbine building floor drains monitor analyzes a sample from the turbine building floor drains harge line (Section 11.2), downstream of any possible fluid addition to the discharge line. The ctor assembly consists of a gamma scintillation detector inserted into the well of a four-pi, shielded liquid sampler. All surfaces in contact with the liquid sample are of stainless steel.
adequate amount of lead shielding is provided in order to reduce the background radiation to a l which does not interfere with the detector sensitivity.
s monitors output is transmitted, indicated, recorded, and alarmed in a manner similar to that he condenser air ejector monitor. A high activity level initiates valve action to divert the uent to the liquid waste system, thereby preventing the discharge of effluent to the ironment in excess of dose limits.
.2.3.6 Reactor Plant Component Cooling Water System Monitor reactor plant component cooling water system monitor continuously analyzes the component ling water for radioactivity (Section 9.2.2.1). A sample is continuously withdrawn from the tor plant component cooling water system downstream of the reactor plant component ling pumps discharge. The sample is monitored by a gamma scintillation detector inserted into well of a four-pi, lead shielded liquid sampler. All surfaces in contact with the liquid sample of stainless steel. Lead shielding is provided in order to reduce the background radiation to a l which does not interfere with the detector sensitivity.
11.5-10 Rev. 30
kground are indicative of a leak into the reactor plant component cooling water system from of the systems containing radioactive fluids which exchange heat with the reactor plant ponent cooling water subsystem.
.2.3.7 Deleted by FSARCR 05-MP3-015
.2.3.8 Regenerant Evaporator Monitor (Removed from Service) ated in the condensate demineralizer liquid waste system (removed from service) the nerant evaporator monitor was designed to measure gross radionuclide concentration in the illate discharged from the regenerant evaporator. Following a high concentration alarm, the uent would be automatically isolated.
evaluation has been performed which has determined that the LWC is not needed. This luation is documented by change to the Radiological Effluent Monitoring Offsite Dose culation Manual (ref. REMODCM CR# 95-7).
.2.3.9 Waste Neutralization Sump Monitor ated in the condensate polishing facility, the waste neutralization sump monitor measures the onuclide concentration in the effluent from a sump located below potentially contaminated densate polishing equipment.
sample is monitored by a gamma scintillation detector inserted into the well of a four-pi, lead lded liquid sampler.
surfaces in contact with the liquid sample are of stainless steel. Lead shielding is provided in er to reduce the background radiation to a level which does not interfere with the detector sitivity.
s monitors output is transmitted, indicated, recorded, and alarmed in a manner similar to that he condenser air ejector monitor.
.2.4 Inservice Inspection, Calibration, and Maintenance st channels of the process and effluent RMS are checked routinely with an installed check rce or manually at the monitor. Periodic tests are also used to verify the operability of alarms automatic actions.
ibration of monitors is conducted periodically in accordance with the technical specifications.
cess and effluent monitors were initially isotopically calibrated at the vendor and concurrently s calibrated to a secondary National Institute of Science and Technology (NIST) traceable rce in a fixed field geometry. Field calibration is accomplished using these secondary sources.
11.5-11 Rev. 30
iochemical analysis takes place in the chemistry laboratory counting room. The facility is ipped with a gamma spectrometer traceable to NIST for the purpose of isotopic analysis.
.2.5 Sampling tion 9.3.2 discusses the various process and effluent samples taken periodically for chemical radiochemical analysis.
le 11.5-3 lists liquid process and effluent samples to be taken periodically and monitored for oactivity. Those not covered in Section 9.3.2 are included in the individual system designs.
pling of these fluid systems is via local sampling connections, e.g., the fuel pool cooling and fication system (Section 9.1.3).
r to collecting a sample, liquid sample lines are purged of stagnant water and undissolved ds for a sufficient time to ensure that a representative sample is obtained.
ple taps suitable for connection to a sampling chamber are provided at all off-line process nitors to obtain a sample for laboratory gamma spectrum analysis.
samples are collected in a sample vessel with valves on each end. After adequate purging of sample vessel, the gas sample is collected by closing valves at both ends of the sample vessel.
isokinetic nozzles used for obtaining uniform samples (Section 11.5.2.2) are designed ording to the guidelines of ANSI N13.1.
ampling room (Section 9.3.2) is provided for remote sampling. Recirculation loops are arated from the sample taps by shielding. Sample sink drains are collected and sent to various ems, depending on the nature of the sample being taken, for reclaiming or processing as essary. Sample sinks are provided with ventilation exhaust hoods.
.3 REFERENCES FOR SECTION 11.5
-1 ANSI N13.1 - 1969. Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities, American National Standards Institute (1974).
-2 EPRI TR104788-R2 PWR Primary-To-Secondary Leak Guidelines - Revision 2.
11.5-12 Rev. 30
TABLE 11.5-1 GASEOUS MONITORS Number of Monitor Mark Number (1) Channels Medium Location Measurement Range Ventilation Vent (2) 3HVR*RE10 Air Aux Bldg 66'-6" Normal Range 1 Normal OPS High Range 1 Post Accident Hydrogenated Vent (2) 3GWS-RE48 1 Gas Aux Bldg 43'-6" Normal OPS Fuel Drop (3) 3RMS*RE41 2 (4) Air Containment 51'-4" Normal OPS 3RMS*RE42 Supplementary Leak 3HVR*RE19 Air Aux Bldg 66'-6" Collection (2)
Normal Range 1 Normal OPS High Range 1 Post Accident Condenser Air Ejector (5) 3ARC-RE21 1 Air Turbine Bldg 38'-6" Normal OPS -- Will Alar With Tube Failure Control Building 3HVC*RE16A 2 (4) Air Control Bldg 64'-6" Post Accident Inlet Ventilation (5) 3HVC*RE16B Containment Atmosphere (2) 3CMS*RE22 Air Aux Bldg 66'-6" Particulate 1 Normal OPS Gas 1 Normal OPS Auxiliary Building (2) 3HVR-RE11 Air Aux Bldg 66'-6" &
43'-6" Page 1 of 3 Rev
Number of (1)
Monitor Mark Number Channels Medium Location Measurement Range Particulate 3HVR-RE12 6 Normal OPS Gas 3HVR-RE13 6 Normal OPS 3HVR-RE14 3HVR-RE15 3HVR-RE16 Fuel Building (2) 3HVR-RE17 Air Aux Bldg 66'-6" Particulate 1 Normal OPS Gas 1 Normal OPS Waste Disposal (2) 3HVR-RE18 Air Aux Bldg 66'-6" Particulate 1 Normal OPS Gas 1 Normal OPS Control Building (2) 3HVC-RE91 Air Control Bldg 64'-6" Particulate 1 Normal OPS Gas 1 Normal OPS ESF Building (2) 3HVQ-RE49 Air ESF Bldg 36'-6" Particulate Note 6 Normal OPS Gas 1 Normal OPS Hydrogen (5) Recombiner 3HVZ*RE09A 2 (4) Air HR Bldg 37'-6" Post Accident Cubicle Vent 3HVZ*RE09B Page 2 of 3 Rev
Number of (1)
Monitor Mark Number Channels Medium Location Measurement Range Main Steam Relief Line (5) 3MSS-RE75 to 78 4 Steam MSVB 70'-6" Post Accident Turbine Driven 3MSS-RE79 1 Steam ESF 36'-6" Post Accident Auxiliary Feedwater Pump Discharge (5)
Main Steam Line 3MSS-RE80A N-16 3MSS-RE80B 4 Steam MSVB 706 Normal OPS Fission Product 3MSS-RE80C 4 Steam MSVB 706 Normal OPS 3MSS-RE80D NOTES:
(1) A and B used to indicate redundant monitors powered from separate safety trains.
(2) Offline monitors.
(3) The fuel drop monitors are configured as high range area monitors, having a minimum sensitivity of 0.1 R/hr and a range of 6 decades.
(4) Redundant monitors.
(5) Inline monitors.
(6) Offline laboratory radionuclide analysis.
Page 3 of 3 Rev
(1) A and B used to indicate redundant monitors powered from separate safet (2) Inline monitors.
(3) Offline monitors.
Page 1 of
BLE 11.5-3 RADIOLOGICAL SAMPLES TAKEN AT REACTOR PLANT SAMPLE SINK Sample Locations Number actor Coolant System (Chapter 5)
Loop No. 1 (Hot Leg) 1 Loop No. 3 (Hot Leg) 1 Pressurizer Vapor Space 1 sidual Heat Removal System (Chapter 5.4.7)
RHR Heat Exchanger Outlet 2 actor Plant Component Cooling Water System (Section 9.2.2.1)
Pump Discharge 2 mary Grade Water System (Section 9.2.8)
Primary Grade Water Tanks 2 emical and Volume Control System (Section 9.3.4)
Letdown Heat Exchanger Outlet 2 Reactor Coolant Filter Inlet 1 Volume Control Tank 1 Boron Thermal Regeneration Outlet 1 ron Recovery System (Section 9.3.5)
Boron Recovery Tanks 2 Boron Test Tanks 2 am Generator Blowdown System (Section 10.4.8)
Blowdown Sample 4 dioactive Liquid Waste System (Section 11.2)
Waste Test Tank 2 dioactive Gaseous Waste System (Section 11.3)
Degasifier Condenser Process Effluent 1 Page 1 of 1 Rev. 30
APPENDIX 11A T I -
SUMMARY
OF ANNUAL RADIATION DOSES (HISTORICAL)
T II - DOSE CALCULATION MODELS AND ASSUMPTIONS (HISTORICAL)
T III - COST-BENEFIT ANAlLYSIS (HISTORICAL)
Rev. 30
dose evaluation presented in this Appendix was developed in support of the original license portions were updated during the MPS-3 restart. These dose estimates are considered orical and not subject to future updating. This information is retained to avoid loss of original gn basis.
stated in Section 11.0, the Radiological Effluent Monitoring and Offsite Dose Calculation nual (REMODCM) provides guidance requirements for system operation, dose calculations, monitoring requirements to ensure MPS-3 compliance with effluent limits. Actual measured centrations of radioactivity released and real time dilution and dispersion estimates are zed to verify compliance with effluent limits. Therefore, MPS-3 operation within the uirements of the REMODCM ensures compliance within effluent limits, rather than operations hin the nominal assumptions utilized in the dose evaluation presented in this Chapter.
calculated annual radiation doses to the maximum individual from liquid and gaseous ways are presented in Tables 11A-1 through 11A-8 and 11A-13 through 11A-15.
le 11A-16 demonstrates that the calculated annual radiation doses are below the design ctives of 10 CFR 50, Appendix I.
maximum calculated organ dose per reactor for an individual from gaseous releases ticulates and radioiodines) is 4.4 mRem/yr to an infants thyroid. This represents a othetical infant living at the residence 2.4 km north-northeast of the site consuming milk from at at the same location.
calculated external exposure to the whole body and skin from immersion in noble gases is E-02 and 6.9E-02 mRem/yr, respectively. These represent the maximum values which occur at site boundary in the direction of the maximum overland /Q, 650 meters east-northeast of lstone 3. The maximum calculated beta and gamma air doses from noble gas releases are E-02 and 8.6E-02 mrad/yr, respectively. These were also calculated 650 meters east-northeast Millstone 3.
liquid releases, the maximum individual was assumed to consume aquatic foods whose cipal habitat is the edge of the initial mixing zone (EIMZ). This location was also servatively used in calculating doses from boating. Doses from swimming and shoreline eation were calculated at the nearest residents beach, located 1.1 km from the point of harge. The maximum calculated whole body dose for an individual from liquid pathways is E-02 mRem/yr in the adult age group. The maximum calculated organ dose for an individual m liquid pathways is 4.4E-01 mRem/yr to an adults GI-UI and 1.8E-01 mRem/yr to a childs oid. These doses were primary due to consumption of aquatic foods.
calculated annual gaseous and liquid doses from the population residing within an 80 km us of the site are presented in Table 11A-17. For liquid effluents, the calculated population e commitment within 80 km for whole body and thyroid are 1.7E+00 and 1.6E+01 man-Rem/
espectively.
11A.1-1 Rev. 30
ulation doses were calculated for a projected population of 3.3 million people residing within m of the site in the year 2010.
calculated annual gaseous and liquid doses to the continguous U.S. population are also ented in Table 11A-17. For liquid effluents, the calculated dose to the contiguous U.S.
ulation is 1.7E+00 man-Rem whole body and 1.6E+01 man-Rem thyroid. For gaseous uents, the calculated dose to the contiguous U.S. population is 2.2E+01 man-Rem whole body 2.5E+01 man-Rem thyroid.
11A.1-2 Rev. 30
EFFLUENTS (Residence 0.81 km ENE) Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-trac Contaminated ground 1.7E+00 (1) 2.0E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 Inhalation 1.2E-01 0.0 8.6E-03 1.2E-01 2.5E-01 1.1E-01 1.4E-01 1.1E-01 Fresh vegetation 9.7E-02 0.0 8.3E-02 1.2E-01 1.5E+00 6.2E-02 3.5E-02 6.0E-02 Stored vegetation 5.7E-01 0.0 4.7E-01 6.7E-01 2.2E-01 3.3E-01 2.2E-01 3.2E-01 Total dose 2.5E+00 2.0E+00 2.3E+00 2.6E+00 3.7E+00 2.2E+00 2.1E+00 2.2E+00 NOTE:
(1) 1.7E+00 = 1.7 x 100 Page 1 of 1 Rev
(Residence 0.81 km ENE) Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.7E+00 (1) 2.0E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 Inhalation 1.2E-01 0.0 1.1E-02 1.2E-01 2.9E-01 1.2E-01 1.6E-01 1.2E-01 Fresh vegetation 5.6E-02 0.0 7.1E-02 9.7E-02 1.2E+00 4.9E-02 2.7E-02 3.8E-02 Stored vegetation 5.9E-01 0.0 4.7E-01 1.0E+00 2.8E-01 4.7E-01 3.0E-01 3.7E-01 Total dose 2.5E+00 2.0E+00 2.5E+00 2.9E+00 3.5E+00 2.3E+00 2.2E+00 2.2E+00 NOTE:
(1) 1.7E+00 = 1.7 x 100 Page 1 of 1 Rev
EFFLUENTS (Residence 0.81 km ENE) Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.7E+00 (1) 2.0E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 Inhalation 1.0E-01 0.0 1.4E-02 1.1E-01 3.1E-01 1.0E-01 1.4E-01 1.0E-01 Fresh vegetation 5.0E-02 0.0 1.2E-01 1.2E-01 1.8E+00 5.9E-02 3.1E-02 3.1E-02 Stored vegetation 6.9E-01 0.0 1.6E+00 1.7E+00 4.8E-01 7.6E-01 4.7E-01 4.3E-01 Total dose 2.5E+00 2.0E+00 3.4E+00 3.6E+00 4.3E+00 2.6E+00 2.3E+00 2.3E+00 NOTE:
(1) 1.7E+00 = 1.7 x 100 Page 1 of 1 Rev
EFFLUENTS (Residence 0.81 km ENE) Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-trac Contaminated ground 1.7E+00 (1) 2.0E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.7E+00 Inhalation 5.9E-02 0.0 9.1E-03 6.2E-02 2.5E-01 6.0E-02 8.2E-02 5.8E-02 Total dose 1.8E+00 2.0E+00 1.7E+00 1.8E+00 1.9E+00 1.8E+00 1.8E+00 1.8E+00 NOTE:
(1) 1.7E+00 = 1.7 x 100 Page 1 of 1 Rev
EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)
Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.8E-01 (1) 2.2E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 Inhalation 3.6E-02 0.0 3.4E-03 3.7E-02 6.2E-02 3.6E-02 4.4E-02 3.6E-02 Fresh vegetation 2.0E-02 0.0 1.4E-02 2.2E-02 1.9E-01 1.4E-02 1.1E-02 1.4E-02 Stored vegetation 1.2E-01 0.0 8.1E-02 1.3E-01 6.4E-02 8.2E-02 6.6E-02 8.1E-02 Goat milk 1.8E-01 0.0 1.2E-01 2.4E-01 5.8E-01 1.1E-01 6.6E-02 5.0E-02 Total dose 5.4E-01 2.2E-01 4.0E-01 6.1E-01 1.1E+01 4.2E-01 3.7E-01 3.6E-01 NOTE:
(1) 1.8E-01 = 1.8 x 10-1 Page 1 of 1 Rev
(Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)
Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.8E-01 (1) 2.2E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 Inhalation 3.7E-02 0.0 4.6E-03 3.8E-02 6.9E-02 3.7E-02 4.8E-02 3.6E-02 Fresh vegetation 1.2E-02 0.0 1.2E-02 1.8E-02 1.5E-01 1.1E-02 7.7E-03 9.3E-03 Stored vegetation 1.3E-01 0.0 1.3E-01 1.9E-01 8.2E-02 1.1E-01 8.7E-02 9.8E-02 Goat milk 1.9E-01 0.0 2.2E-01 3.9E-01 9.0E-01 1.7E-01 1.0E-01 6.5E-02 Total dose 5.5E-01 2.2E-01 5.5E-01 8.2E-01 1.4E+00 5.1E-01 4.2E-01 3.9E-01 NOTE:
(1) 1.8E-01 = 1.8 x 10-1 Page 1 of 1 Rev
EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)
Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.8E-01 (1) 2.2E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 Inhalation 3.2E-02 0.0 6.1E-03 3.4E-02 7.0E-02 3.3E-02 4.2E-02 3.2E-02 Fresh vegetation 1.2E-02 0.0 2.1E-02 2.2E-02 2.2E-01 1.3E-02 9.3E-03 9.3E-03 Stored vegetation 1.7E-01 0.0 2.9E-01 3.2E-01 1.4E-01 1.8E-01 1.4E-01 1.4E-01 Goat milk 2.0E-01 0.0 5.2E-01 6.5E-01 1.8E+00 2.8E-01 1.6E-01 1.0E-01 Total dose 5.9E-01 2.2E-01 1.0E+00 1.2E+00 2.4E+00 6.9E-01 5.3E-01 4.6E-01 NOTE:
(1) 1.8E-01 = 1.8 x 10-1 Page 1 of 1 Rev
EFFLUENTS (Residence 2.4 km NNE; Goat Pasture 2.4 km NNE)
Annual Dose (mRem/yr)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI-tra Contaminated ground 1.8E-01 (1) 2.2E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 1.8E-01 Inhalation 1.9E-02 0.0 4.1E-03 2.0E-02 5.3E-02 1.9E-02 2.5E-02 1.9E-02 Goat milk 2.4E-01 0.0 8.5E-01 1.2E+00 4.2E+00 4.4E-01 2.6E-01 1.5E-01 Total dose 4.4E-01 2.2E-01 1.0E+00 1.4E+00 4.4E+00 6.4E-01 4.6E-01 3.5E-01 NOTE:
(1) 1.8E-01 = 1.8 x 10-1 Page 1 of 1 Rev
Page 1 of 1 Rev. 30 Page 1 of 1 Rev. 30 Page 1 of 1 Rev. 30 Page 1 of 1 Rev. 30 Annual Dose (mRem/yr)
Pathway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LL Fish 0.00E+00 7.29E-03 1.40E-02 1.04E-02 7.72E-02 5.15E-03 1.90E-03 1.02E-02 Invertebrate 0.00E+00 1.97E-02 1.34E-02 1.05E-02 1.02E-01 1.01E-01 6.33E-04 4.26E-01 Shoreline 3.18E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 Swimming 0.00E+00 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 Boating 0.00E+00 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 Total 3.18E-03 2.98E-02 3.02E-02 2.37E-02 1.82E-01 1.09E-01 5.31E-03 4.39E-01 Page 1 of 1 Rev
Annual Dose (mRem/yr)
Pathway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LL Fish 0.00E+00 7.63E-03 1.43E-02 6.28E-03 7.24E-02 5.16E-03 2.11E-03 7.29E-03 Invertebrate 0.00E+00 2.09E-02 1.38E-02 1.02E-02 9.63E-02 1.05E-01 7.12E-04 3.01E-01 Shoreline 3.18E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 2.73E-03 Swimming 0.00E+00 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 2.89E-05 Boating 0.00E+00 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 1.80E-05 Total 3.18E-03 3.13E-02 3.09E-02 1.93E-02 1.71E-01 1.12E-01 5.60E-03 3.11E-01 Page 1 of 1 Rev
Annual Dose (mRem/yr)
Pathway Skin Bone Liver Total Body Thyroid Kidney Lung GI-LL Fish 0.00E+00 9.43E-03 1.23E-02 2.95E-03 7.57E-02 4.34E-03 1.67E-03 2.74E-03 Invertebrate 0.00E+00 2.73E-02 1.27E-02 1.14E-02 1.06E-01 9.31E-02 5.94E-04 9.47E-02 Shoreline 1.78E-03 1.53E-03 1.53E-03 1.53E-03 1.53E-03 1.53E-03 1.53E-03 1.53E-03 Swimming 0.00E+00 1.62E-05 1.62E-05 1.62E-05 1.62E-05 1.62E-05 1.62E-05 1.62E-05 Boating 0.00E-00 1.00E-05 1.00E-05 1.00E-05 1.00E-05 1.00E-05 1.00E-05 1.00E-05 Total 1.78E-03 3.83E-02 2.66E-02 1.59E-02 1.84E-01 9.90E-02 3.82E-03 9.90E-02 Page 1 of 1 Rev
MILLSTONE 3 NUCLEAR PLANT WITH APPENDIX I DESIGN OBJECTIVES Appendix I Design Criterion Objective (1) Calculated Dose seous Effluents mma air dose 10 mrad/yr 8.6E-02 (2) mrad/yr (3) ta air dose 20 mrad/yr 6.6E-02 mrad/yr (3) ble gas - total body 5 mRem/yr 3.8E-02 mRem/yr (3) ble gas - skin 15 mRem/yr 6.9E-02 mRem/yr (3) ines and particulates, any 15 mRem/yr 4.4E+00 mRem/yr (4) an uid Effluents:
tal body 3 mRem/yr 2.4E-02 mRem/yr y organ 10 mRem/yr 4.4E-01 mRem/yr (5)
TES:
Per reactor.
8.6E-02 = 8.6 x 10-2.
Site boundary 650 meters ENE of Millstone 3.
Infant thyroid dose at residence with goat, 2.4 km NNE.
Adult GI-UI dose is calculated to be the highest organ dose.
Page 1 of 1 Rev. 30
80-KM POPULATION DOSE Annual Dose Per Reactor Unit Total Body (man-Rem) Thyroid (man-Rem) tural radiation background (1) 3.3E+05 (2) 3.3E+05 uid effluents 1.7E+00 1.6E+01 ble gas effluents 1.3E-01 1.3E-01 dioiodines and particulates (3) 4.7E+00 7.7E+00 CONTIGUOUS U.S. POPULATION DOSE Annual Dose Per Reactor Unit Total Body (man-Rem) Thyroid (man-Rem) uid effluents 1.7E+00 1.6E+0 ble gas effluents 1.5E-0 4.0E-0 dioiodines and particulates (3) 2.2E+0 2.5E+0 TES:
Natural Radiation Exposure in the United States, U.S. Environmental Protection Agency, ORP-SID-72-1 (June 1972), using the average state background dose (100 mRem/yr), and year 2010 projected population of 3.3 million.
3.3E+05 = 3.3 x 105 Carbon-14 and tritium have been added to this category.
Page 1 of 1 Rev. 30
es to Humans culation of dose rates to the maximum individual and to the population residing within an km radius of the site are based on the methodology and equations of U.S. Nuclear Regulatory de 1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents The Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October
- 7. Stone & Webster Engineering Corporation Computer Programs IND1109E, POP1109E, 1109E, DUCKMANE, and NEPA were used in the design case analysis of the maximum vidual and population doses. These computer codes have been verified by comparing the ne & Webster program outputs to results obtained from the U.S. NRC Computer Codes SPAR and LADTAP and to hand calculations, where appropriate. The U.S. NRC Computer e, LADTAP II was used for the expected case analysis. The U.S. NRC Computer Code BFIN was used to analyze the elevated release of noble gases. NRC default values have been d in lieu of site specific data, where site data was unavailable. The site specific data that was d for this analysis is listed in Tables 11A.2-1, 11A.2-2, and 11A.2-3.
following sections present the equations used in the analysis for each pathway considered.
es from Liquid Pathways generalized equation for calculating radiation doses to humans via liquid pathways is:
Raipj = (Cip)(Uap)(Daipj) (11A.2-1) re:
j = the annual dose to organ j of an individual of age group a from nuclide i via pathway p, in mRem/yr
= the concentration of nuclide i in the media of pathway p, in pCi/l, pCi/kg, or pCi/m2
= 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) j = 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 exposure to deposited activity in sediment or on the ground Aquatic Foods U ap M p R apj = 1100 ----------------- Q i B ip D aipj exp ( - i t p ) (11A.2-2)
F i
11A.2-1 Rev. 30
Bip = the equilibrium bioaccumulation factor for nuclide i in pathway p, expressed as the ratio of the concentration in biota (in pCi/kg) to the radionuclide concentration in water (in pCi/l), in l/kg Mp = the mixing ratio (reciprocal of the dilution factor) at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food), dimensionless F = the flow rate of the liquid effluent in cu ft/s Qi = the release rate of nuclide i, in Ci/yr Rapj = the total annual dose to organ j of individuals of age group a from all of the nuclides i in pathway p, in mRem/yr i = the radioactive decay constant of nuclide i, in hr tp = 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 the other symbols are as previously defined.
Doses from Shoreline Deposits Foods U ap M p W R apj = 110, 000 ----------------------- Q i T i D aipj [ exp ( - i t p ) ] [ 1 - exp ( - i t b ) ] (11A.2-3)
F i
re:
the shoreline width factor that describes the geometry of the exposure, dimensionless the radiological half-life of nuclide i, in days the period of time for which sediment or soil is exposed to the contaminated water, in hours
,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 other symbols are as previously defined.
Doses from Swimming and Boating 11A.2-2 Rev. 30
The equation for calculation of the external dose to skin and the total body dose from swimming (water immersion) or boating (water surface) is:
U ap M p R apj = 1100 ----------------- Q i D aipj exp ( - i t p ) (11A.2-4)
F Kp i
re:
= geometry correction factor equal to 1 for swimming and 2 for boating, dimensionless (no credit is taken for the shielding provided by the boat) other symbols are as previously defined.
es from Air Pathways Gamma and Beta Doses from Noble Gas Discharged to the Atmosphere
- a. Annual Gamma and Beta Air Doses from Noble Gas Releases (ground level) 4 D ( r, ) or D ( r, ) = 3.17 x 10 Q i [ Q ] ( r, ) ( DF i or DF i ) (11A.2-5) i re:
D(r,), D(r,) = the annual gamma and beta air doses at the distance r in the sector at angle from the discharge point in mrad/year Qi = the release rate of the radionuclide i, in Ci/year
[/Q](r,) = the annual average gaseous dispersion factor 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 104 = the number of pCi per Ci divided by the number of seconds per year
- b. Annual Total Body Dose from Noble Gas Releases (ground level)
D T ( r, ) = S F i ( r, ) DFB i (11A.2-6) i 11A.2-3 Rev. 30
DT (r,) = the total body dose due to immersion in a semi-infinite cloud at the distance r in sector , in mRem/year SF = the attention factor that accounts for dose reduction due to shielding provided by residential structures, dimensionless i (r,) = the annual average ground-level concentration of radionuclide i at the distance r in sector , in pCi/cu m DFBi = the total body dose factor for a semi-infinite cloud of the radionuclide i which includes the attenuation of 5 g/sq cm of tissue, in mRem-cu m/pCi-yr
- c. Annual Skin Dose from Noble Gas Releases (ground level)
D S ( r, ) = 1.11 S F i ( r, ) DF + i ( r, ) DFS i (11A.2-7) i i i re:
DS (r,) = the annual skin dose due to immersion in a semi-infinite cloud at the distance r in sector , in mRem/yr DFSi = the beta skin dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuation by the outer dead layer of the skin, in mRem-cu m/pCi-yr 1.11 = the average ratio of tissue to air energy absorption coefficients other parameters are as previously defined.
- d. Annual Gamma Air Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High 260 1 D D ( r, ) = --------------- ------- f ns a ( E k ) E k I ( H, u, s, z, E k ) Q ni A ki (11A.2-8) r ( ) un n s k i 11A.2-4 Rev. 30
Aki 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 fns is the joint frequency of occurrence of stability class s and wind speed class n for sector
, dimensionless I(H,u,s,z,Ek) is the dimensionless numerical integration constant 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 QDni is the release rate of radionuclide i, corrected for decay during transit to the distance r under wind speed un, in Ci/yr un is the mean wind speed of wind speed class n, in m/sec is the sector width over which atmospheric conditions are average, in radians a (Ek) 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-m3 - disintegration/sec-MeV-Ci
- e. Annual Total Body Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High T T D ( r, ) = 1.11S F D k ( r, ) exp [ - a ( E k )t d ] (11A.2-9) k 11A.2-5 Rev. 30
DT(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 SF is the attenuator factor that accounts for the dose reduction due to shielding provided by residential structures, dimensionless td is the product of tissue density and depth used to determine a total body dose, in g/cm2 Ta (Ek) is the tissue energy absorption coefficient, in cm2/g; and 1.11 is the average ratio of tissue to air energy absorption coefficients
- f. Annual Skin Dose from Noble Gas Releases from Free-Standing Stacks More Than 80 Meters High s 4 D D ( r, ) = 1.11 S F D ( r, ) + 3.17 x 10 Q i [ Q ] ( r, ) DFS i (11A.2-10) i re:
DFSi 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-m3/pCi-yr Ds(r,) is the annual skin dose at the distance r in sector , in mrem/yr other parameters are as defined in preceding paragraphs.
Doses from Radioiodines and Other Radionuclides (not including Noble Gases)
Released to the Atmosphere
- a. Annual Organ Dose from External Irradiation from Radionuclides Deposited onto the Ground Surface G G D j ( r, ) = 8760 S F C i ( r, ) DFG ij (11A.2-11) i re:
G j (r,) = the annual dose to the organ j at location (r,), in mRem/yr SF = a shielding factor that accounts for the dose reduction due to shielding provided by residential structures during occupancy, dimensionless 11A.2-6 Rev. 30
sq m DFGij = the open field ground plane dose conversion factor for organ j from radionuclide i, in mRem-sq m/pCi-hr 8,760 = the number of hours in a year
- b. Annual Organ Dose from Inhalation of Radionuclides in Air A
D ( r, ) = R a i ( r, ) DFA ija (11A.2-12) ja i
re:
DAja (r,) = the annual dose to organ j of an individual in the age group a at location (r,)
due to inhalation, in mRem/yr Ra = the annual air intake for individuals in the age group a, in cu m/yr i (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/pCi
- c. Annual Organ Dose from Ingestion of Atmospherically Released Radionuclides in Food D v v D ( r, ) = DFI ija [ U a f g C i ( r, ) (11A.2-13) ja i
m m
+ U a C i ( r, )
F F L L
+ U a C i ( r, ) + U a f 1 C i ( r, ) ]
re:
Cvi(r,), Cmi(r,) = the concentrations of radionuclide i in produce CLi(r,), CFi(r,) = (non-leafy vegetables, fruits, and grains), milk, leafy vegetables, and meat, respectively, at location (r,), in pCi/kg or pCi/l DDja(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 11A.2-7 Rev. 30
fg, f1 = the respective fractions of the ingestion rates of produce and leafy vegetables that are produced in the garden of interest Uva, Uma, UFa, ULa = 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 eral Expression for Population Doses general expression for calculating the annual population-integrated dose is:
P D j = 0.001 P d D jda f da (11A.2-14) d a re:
DPj = the annual population-integrated dose to organ j (total body or thyroid), in man-Rems or thyroid man-Rems Pd = the population associated with subregion d Djda = the annual dose to organ j (total body or thyroid) of an average individual of age group a in subregion d, in mRem/yr fda = the fraction of the population in subregion d that is in age group a 0.001 = the conversion factor from mRem to Rem above equation used in conjunction with the preceding equations and average usage factors each age group was used to calculate the population doses.
further refinements on the preceding equations used to calculate the doses to man, see ulatory Guide 1.109, Revision 1.
es to Biota Other Than Man culation of dose rates to biota other than man was performed by means of the computer grams ARRRG and CRITER (Soldat et al., 1974), developed at the Pacific Northwest oratory of Battelle Memorial Institute under contract to the Atomic Energy Commission C). Bioaccumulation factors used in ARRRG and CRITER have been updated to correspond he latest published values in Regulatory Guide 1.109, Revision 0 (plants) and Regulatory de 1.109, Revision 1 (all others). Site specific data used in this analysis are presented in Table
.2-4.
following sections provide a summary of the dose models used in the analysis for each way considered.
11A.2-8 Rev. 30
atic organisms were considered to receive an internal dose rate from uptake and concentration adiochemicals in the water and from exposure through the food chain. Dose rates to primary anisms were calculated directly from radioisotopic concentrations in discharge water and from ccumulation factors. The dose rate through the food chain was estimated for secondary anisms such as muskrats and raccoons feeding on primary organisms whose radionuclide tent was estimated in the first calculation.
ations used by the program CRITER for these calculations are as follows:
( DR ) i = AE i b i (11A.2-15) re:
(DR)i = dose rate for radionuclide i (mrad/yr)
Ei = effective absorbed energy in organ of interest (MeV/dis) bi = specific body burden of nuclide i (pCi/kg) dis - kg - mrad A = conversion factor = 0.0187 ---------------------------------------
pCi - yr - MeV bi = Ciw Bi re:
Ciw = concentration of nuclide i in water (pCi/l)
Bi = bioaccumulation factor for nuclide i (pCi/kg per pCi/l) concentration in water Ciw is calculated from:
Qi Ri Mp C iw = 1119 ------------------- exp ( - i t p ) (11A.2-16)
F re:
Qi = release rate of nuclide i (Ci/yr)
Ri = reconcentration factor to estimate recycling of effluent Mp = mixing ratio at point of exposure (1/dilution factor) 11A.2-9 Rev. 30
i = radiological decay constant of nuclide i (hr-1) tp = transit time for nuclides to reach point of exposure (hr) 1,119 = constant to convert Ci/yr per cu ft/s to pCi/l total body dose rate to secondary organisms was calculated as follows (Soldat et al., 1974):
DR'i = 0.365 bi P' D'i (11A.2-17) re:
i = total body dose rate to secondary organisms due to nuclide i (mrad/yr) 5 = kg-day/g-yr D i ( man ) e' i D' i = 70,000 ---------------------- -----
e i ( man ) m' Di (man) = total body dose conversion factor for man for radionuclide in mRem/pCi ei (man) = effective absorbed energy for man for radionuclide i (meV/disintegration) e'i = effective absorbed energy for secondary organism for radionuclide i (meV/disintegration) mi = mass of secondary organism (grams)
Pi = consumption rate of primary organisms by the secondary organism (grams/day) 70,000 = total body mass of adult (grams) actual equation used by CRITER was of the form:
n 7 M p P' DR' = 2.86 x 10 ------------ Q i R i B i e 'i exp ( - i t p ) [ D i e i ] ( man ) (11A.2-18)
Fm' i=1 re:
DR' = total body dose rate to secondary organisms (mrad/yr) n = 136, number of isotopes 2.86x107 = (0.365) (1119) (70,000) other parameters are as previously defined.
11A.2-10 Rev. 30
n Up Mp Wf
( DR ) pr = 111, 900 ------------------------ Q i R i T i exp ( - i t p ) ( 1 - exp ( i t )D ipr ) (11A.2-19)
F i=1 re:
(DR)pr = dose rate to organ r (total body or skin) from pathway P (mrad/yr)
Wf = shore width factor = 0.5 (ocean shoreline)
Ti = radiological half-life of isotope i (days) n = 136, number of isotopes 111,900 = constant to convert (Ci/yr)/(cu ft/sec) to pCi/liter e for Swimming and Water Surface Exposure n
Up Mp DR' pr = 1119 ---------------- Q i R i D ipr exp ( - i t p ) (11A.2-20)
F Kp i = 1 re:
Kp = hemispherical correction constant = 1 for swimming and 2 for boating n = 136, number of isotopes e from Immersion in Gaseous Effluents se doses were calculated in the same manner as doses to humans with appropriate changes in factors as shown in Table 11A.2-1.
erences for Appendix 11A Part II - Dose Calculation Models and Assumptions mic Energy Commission 1973. Final Environmental Statement Concerning Proposed Rule king Action; Numerical Guides for Design Objectives and Limiting Conditions for Operation eet the Criterion (as low as practicable) for Radioactive Material in Light Water Cooled lear Power Reactor Effluents. Washington, D.C.
dat, S.K.; Robinson, N.M.; and Baker, D.A. 1974. Models and Computer Codes for Evaluating ironmental Radiation Doses. Battelle Pacific Northwest Laboratories BNWL-1754, Richland, sh.
11A.2-11 Rev. 30
POPULATION SERVED Approximate Transit Time Distance from Dilution to Point of Population Location of Analysis Site (km) Factor Analysis (hr) Served ge of initial mixing zone (1) 0 3 0.0 (assumed) -
osest accessible shoreline (2) 1.1 7.2 0.0 (assumed) -
ge of initial mixing zone (3) 0 3 0.0 (assumed) 3.3E+06 (4)
TES:
Location used to calculate doses to maximum offsite individual from ingestion of aquatic foods and boating.
Location used to calculate doses to maximum offsite individual from shoreline recreation and swimming.
Location used to calculate doses to population 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 applied to the entire 80 km radius. It is also assumed that the entire 80 km radius population participates in swimming and boating.
3.3E+06 = 3.3 x 106.
Page 1 of 1 Rev. 30
parameters and assumptions used are recommended values to be used, in lieu of site specific
, from Regulatory Guide 1.109, Revision 1.
following are site specific parameters or parameters for which there is no recommended e:
normal circulation flow rate (for 3-unit operation) = 4,160 cu ft/sec
= transit time = see Table 11A.2-1 Note: Tp used in calculations was increased, where appropriate, by the distribution or holdup time recommended by Regulatory Guide 1.109, Revision 1.
fractional equilibrium ratio of C14 = 1 (continuous release); = 0.0073 (intermittent containment release); = 0.062 (intermittent steam generator blowdown release) annual release rate of radionuclide i, Ci/yr (Tables 11.2-7 and 11.3-1) fraction of year animals graze on pasture = 0.67 (8 months) fraction of daily feed which is pasture grass when animal is grazing = 1 (100%)
absolute humidity of atmosphere at location of analysis 9.86 g/cu m
= usage factor (hr/year of exposure):
Maximum Individual Adult Teen Child Swimming 100 100 56 Boating 52 52 29 80 km Radius Adult Teen Child Swimming 3.5 10 12 Boating 29 29 16.5 Page 1 of 2 Rev. 30
Rev. 30 iological Release Points:
Millstone stack (continuous release)
Ventilation vent (intermittent release)
Ventilation vent (continuous release)
Turbine building vent (continuous release)
Steam generator blowdown vent (intermittent release)
Condensate polishing building vent (continuous release) eorological Parameters - c/Q = Sec/m3; D/Q = m-2 Growing/Grazing Location Resident Annual Average Season 0 m ENE Maximum resident /Q1 (1) 4.03E-08 (2) 4.59E-08 D/Q1 2.16E-09 1.70E-09
/Q2 5.24E-06 5.70E-06 D/Q2 6.22E-08 5.40E-08
/Q3 3.50E-06 3.98E-06 D/Q3 4.15E-08 3.72E-08
/Q4 1.22E-05 1.54E-05 D/Q4 7.57E-08 7.86E-08
/Q5 1.95E-05 2.45E-05 D/Q5 1.21E-07 1.28E-07
/Q6 1.19E-05 1.51E-05 D/Q6 7.39E-08 7.86E-08 00 m NNE Maximum goat /Q1 6.60E-08 8.00E-08 D/Q1 7.04E-10 7.58E-10
/Q2 3.08E-06 3.56E-06 D/Q2 1.48E-08 1.55E-08 Page 1 of 2 Rev. 30
Location Resident Annual Average Season
/Q3 7.96E-09 1.04E-06 D/Q3 3.94E-09 4.87E-09
/Q4 9.71E-06 1.28E-06 D/Q4 3.81E-09 4.62E-09
/Q5 2.41E-06 3.02E-06 D/Q5 9.84E-09 1.12E-08
/Q6 9.20E-07 1.23E-06 D/Q6 3.76E-09 4.59E-09 0 m ENE Maximum site /Q1 1.27E-08 -
boundary D/Q1 2.34E-09 -
/Q2 8.08E-06 -
D/Q2 9.70E-08 -
/Q3 4.81E-06 -
D/Q3 6.29E-08 -
/Q4 1.90E-05 -
D/Q4 1.22E-07 -
/D5 3.06E-05 -
D/Q5 1.96E-07 -
/Q6 1.86E-05 -
D/Q6 1.19E-07 -
TES:
Numerical superscripts correspond to release point.
4.03E-08 = 4.03 x 10-8.
Page 2 of 2 Rev. 30
BLE 11A.2-4 PARAMETERS AND ASSUMPTIONS USED IN ESTIMATING DOSES TO BIOTA Values Assigned Primary Organisms Secondary Organisms (Fish, Crustaceans, Parameter Mollusks, Algae) Muskrat Heron Duck Raccoon (recirculation factor 0 0 0 0 0 (flow rate, cfs) 4,160 4,160 4,160 4,160 4,160 (mixing ratio) (1) 0.333 0.333 0.333 0.333 0.333 (shore width factor) - 0.5 0.5 0.5 0.5 water immersion) - 1 - - -
(water surface) - - 2 2 -
ective radius (cm) 2 6 11 5 14 mass (kg) - 1 4.6 1 12 food consumption (gpd) aquatic plants - 100 - 100 -
fish - - 600 - -
invertebrates - - - - 200 usage (hr/yr) shoreline - 2,922 2,922 4,383 2,191 water immersion - 2,922 - - -
water surface - - 2,922 4,383 -
oldup time (hr) 0 0 0 0 0 sidence time (mo) 12 12 12 12 12 TE:
Edge of mixing zone and nearest shoreline.
Page 1 of 1 Rev. 30
s appendix presents the results of cost-benefit analyses performed in accordance with Section D of 10 CFR 50, Appendix I.
ments to the liquid and gaseous effluent systems and respective potential reductions to the ual population exposure are taken from the U.S. NRC Regulatory Guide 1.110, Cost-Benefit lysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (Regulatory de 1.110, 1976). The beneficial savings of each augment were calculated by multiplying the ulated dose reduction by $1,000 per man-Rem or $1,000 per man-thyroid-Rem. The cost of owed money was conservatively assumed to be 10 percent. The equations and site specific required for the dose calculations are presented in Part I of this appendix.
ments to the Liquid Effluent Treatment System le 11A.3-1 presents the calculated base case annual total body dose (man-Rem) and thyroid e (man-thyroid-Rem) associated with the operation of the plant liquid radwaste system for the ulation expected to live within an 80-km radius of the plant for the year 2010. Assuming that h augment is capable of reducing the population doses to zero (an extremely conservative mption), the maximum benefit to be derived from any augment would be $1,700 for reducing
-Rem exposures to zero and $16,000 for reducing man-thyroid-Rem exposure to zero.
n analysis of the annualized procurement, installation, operation, and maintenance costs, the t expensive liquid radwaste augment was found to be $19,000 per year for a plant located in northeastern United States. Since the benefit from this augment would be less than the esponding total annualized cost, the cost-benefit ratio is greater than 1. The operation of itional equipment for the purpose of reducing the annual population dose would not be cost ctive. Therefore, the most cost-beneficial system has been included in the current plant gn.
ments to the Gaseous Effluent Treatment System le 11A.3-2 presents the calculated base case annual total body man-Rem and thyroid
-Rem associated with the operation of the gaseous radwaste system for the 80-km radius ulation.
uming that each augment is capable of reducing the population doses to zero, the maximum efit to be derived from any augment would be $4,800 for reducing man-Rem exposures to zero
$7,800 for reducing man-thyroid-Rem exposures to zero.
n analysis of the annualized procurement, installation, operation, and maintenance costs, the t expensive gaseous radwaste augment was found to be $8,700 per year for a plant located in northeastern United States. Since the benefit from this augment would be less than the esponding total annualized cost, the cost-benefit ratio is greater than 1. The operation of itional equipment for the purpose of reducing the annual population dose would not be cost 11A.3-1 Rev. 30
erence for Appendix 11A Part III - Cost-Benefit Analysis ulatory Guide 1.110, 1976. Cost-Benefit Analysis for Radwaste Systems for Light Water-led Nuclear Power Reactor. March 1976.
11A.3-2 Rev. 30
TABLE 11A.3-1 BASE CASE ANNUAL POPULATION DOSES DUE TO LIQUID EFFLUENTS (1)
Total Body Dose Thyroid Dose (man-Pathway (man-Rem) thyroid-Rem) estion of fish 1.2E-00 (2) 9.2E+00 estion of other seafood 2.5E-01 6.5E+00 oreline recreation 2.4E-01 2.4E-01 imming 3.1E-03 3.1E-03 ating 6.2E-03 6.2E-03 tal 1.7E+00 1.6E+01 TES:
Total annual dose from all existing pathways for Millstone 3 operation.
1.2E-00 = 1.2 x 100 Page 1 of 1 Rev. 30
ABLE 11A.3-2 BASE CASE ANNUAL POPULATION DOSES DUE TO GASEOUS EFFLUENTS (1)
Total Body Dose Thyroid Dose (man-Pathway (man-Rem) thyroid-Rem) bmersion 1.3E-01 (2) 1.3E-01 alation 8.8E-01 1.9E+00 nding on contaminated ground 3.3E+00 3.3E+00 estion of fruits, grains, and vegetation 8.6E-02 2.6E-01 estion of cow milk 3.9E-01 2.2E+00 estion of meat 2.1E-02 2.7E-02 tal 4.8E+00 7.8E+00 TES:
Total annual dose from all existing pathways for Millstone 3 operation 1.3E-01 = 1.3 x 10-1 Page 1 of 1 Rev. 30
MPS-3 FSAR FIGURE 11.2 - 3 EXPECTED RADIOACTIVE LIQUID WASTE SOURCE AND DISCHARGE PATHS O.F.= 100 1, OTl£RS D.F.= 1.0 D.F.= 2 Ca, Rb FROJA rorrAltN:Kl" 6UIilliNG SU\4P*- WASTE WASTE TO PRIMARY
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REACTOR PlANl SANPlE SINKS HlGHLEVEl WASTE WASTE.
WASTE WASTE TEST PULtP OOOt£AAl-3 t.t. ~
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(;)
my YfASl£S
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TO PRIMARY o 5 D.F..=10 I. Br D.F.=10 LOr 660 GPO GRADE WATER :l:
O.F.=2C60 Rb D.F.=2Cs. Rb D.f.=~2 I.8r w D.F.::; 'J.o
~ t D.f.=10 AU. 0Tl£RS D.F. = ].0 OJ".=10 AU. 0ll0S D.f.=1 AU. 0Tl£RS OF.::; 1.0 O.F. = LO O.F. = 1.0 (J)
V>
REACTOR COOlANT BlEED 10 OBCICAl C£SlUU 6000N RECOVERY ~ .~ BOO~
& VOLlH: OEGASlFIER REMlVAl P1..NP TEST DOOJ£RAl- FR-TER 6 RECOV£RY
.... H I-t r--- ~ H .-
1440 CPO CONTID. ~~ 100 r-. TANK EVAPORATOO ~ TAm IZER SYSlEU EXCHANGER 150.000 GAL 25GPN 50 CPM 12.000 GAl REACTOR PlANT GAS8lUS ORAlNS 7
300 Q>O D.F.= LO REGtNERANT CtOOCAl.S 8 FILTER 3400 GPO D.f.::; 1.0 D.F.::; 1.0 Ul$C£lLAtEOOS lOfH.EV£l WASTE 9 ...
- LOI LEVEL 40 GPO EfflLENT WASTE ~ PLI.lP H FILlER
- ~~ DRAIN TAN(
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- - VIA REACTOR PLANT AERATED ORA.lNS SYSTEM.
- - NORMALLY DISCHARGED OIRECTLY TO ENVIRONMENT.
- - NOT REQUIRED TO MEET 10 CFR 50 APPENDIX I FIGURE 11.2-3 GUIDELINES. EXPECTED RADIOACTIVE LIQUID WASTE \
D.F. - DECONAMINATION FACTORS ARE CONSISTENT WITH SOURCE AND DISCHARGE PATHS .
NUREG-0017. SECTION 2.2.21.
GPO - ESTIMATED OR EXPECTED FLOWS ARE PROVIDED FOR ILLUSTRAnON OF PROCESSED WASTES. FLOW ARE NOT NECESSARILY CONSISTENT WITH ACTUAL OPERATING FLOW DATA.
CAD FILE: 1123.dqn / 1l23.clt MAY 1998 May 1998 Rev. 20.3
MPS-3 FSAR FIGURE 11.3-2 VENTILATION SYSTEM COMPOSITE DRAWING NORMAL OPERATION.
TO ATMOSPHERE REACTOR PLANT VENTILATION VENT SERVICE BUILDING WASTE DISPOSAL BUILDING AUXILIARY BUILDING FUEL BUILDING Purge TO MILLSTONE STACK CONTAINMENT STRUCTURE Containment Vacuum System MAIN CONDENSER AIR EJECTOR RADIOACTIVE GASEOUS WASTE SYSTEM TO ATMOSPHERE ESF ESF BUILDING BUILDING EXHAUST TURB. BUILD.
ROOF TURBINE EXHAUSTERS BUILDING STEAM GENERATOR BLOWDOWN FLASH TANK VENT TURBINE CONDENSATE GLAND SEAL POLISHING STEAM CONDENSER BUILDING EXHAUST 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 to preserve original design basis.
SINDER PROMOTE/< Co CATI\LYS T on REACTOR ACCEPTA8LE ALTERNATE P~OCESS
~
PLANT SERVICE '2 ~',~E i*04Ci/l'H
$PENT OTHER SERVICE
'U6Ei 01Cl/FTJ 2.GOE +01 Ci/YR~
RESINS (II le:; FT -'/YR jHIPPING CONTAINER r-----+ OFf SITE 1.30E-01Ci/FP i
~
SPENT FILTERS 6.77E~ 02CI/YR SHIELDING
~
MISCELLAN EOU S CASKS AS 1.69Ci/FT-' INCOMPRESSIBLE ~ REQUIRED OPERATION 6.
MAINTENANCE WASTE (2)
NEGLIGIBLE 500 FP/YR
... 900 fT3/YR ACTIVITY
~
RADIOACTIVE SPENT LIQUID WASTE 6.16Ci/YR RESINS (1)(51 ...
SYSTEM 2.04E-02cI/FT3 400 FT::l/YR BORON EVAPORATOR BORON 1.27cl/YR ~ BOTTOMS (3)
~
RECOVERY p-SYSTEM 8.47-C3Ci/rT I~O FT3/YR I
'-CONDENSATE CO:.lDE:-':Sf.TE i POLISHING POL..lSHING ---_~.,[ SPENT RESINS ~-~i&o MILLSTON£ UNIT 2 PROCE5SlEG FACILITY FACILITY - j 0 FT3 IYR (41 I
OPERATIO~J t. ' MISCELLANEOUS 4
MAINTENANCE COMPHESSIBLE
~ WASTES (2) MILLSTO"lE RAOWASF RED ... CTl01' FACILITY NEGL!GIBLE ACTIVITY 1 4~!Y.l F:~/YR NOTES:
J
- 1. Ci FT {YR VALUES BASED UPON VOLUME OF RAW SPENT RESIN.
- 2. Ci F-r {YR VALUES BASED UPON VOLUME OF PACKAGED WASTE.
J
- 3. Ci FT IYR VALUES BASED UPON VOLUME OF RAW BOTTOMS.
- 4. NO CONDENSATE POLISHING SPENT RESINS ARE EXPECTED TO BE GENERATED FOR NORMAL EXPECTED RADIATION LEVELS.
- 5. ALTERNATE METHOD WOULD PRODUCE APPROXIMATELY 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 to preserve original design basis.
BINDER PROMOTER C CATALYST OR ACCEPTABLE REACTOR PLANT ALTERNATE PROCESS SERVICE
- 9. 69E-05 Ci/¥R
.92 E.02Ci/FT SPENT RESINS (I) l OTHER I. 24E-04 Ci/YR 1600 fT~YR SHIPPING CONTAINER r---+ OFf SITE SERVICE
.'-01 Ci/FTs SPENT FILTERS Z .' ("04 C'/YR t
SHIELOING MISCELLANEOUS CASKS AS 2 .63 E-Ol CI/FT3 INCOMPRESSIBLE OPERATION £. REQUIRED p
MAINTENANCE WASTES (2)
NEGLIGIBLE ACTIVITY 1000 FT'/YR - t600 FTs/YR RADIOACTIVE SPENT
.SSE-OlCI/YR LIQUID WASTE RESINS (1)(:\)
r SYSTEM 2 .76 E-02 CI/FT 830 frs/YR BORON EVAPORATOR BORON 5 .SSE-OICi/VR BOTTOMS (3)
RECOVERY SYSTEM 9 .25E-02Ci/FT
.. 600 FrS/YR
~
CONDENSATE POLISHING CONDENSATE POLISHING 1.56£.. 02 C'/YR
. SPENT RESINS t-----1~~ MILLSTONE UNIT 2 rACILlTY -.- PROCE'SSING FACILITY 10.000 fTllYR OPERATION e MISCELLANEOUS tAAINTENANCE COMPRESSI6LE WASTES(Z) 1-----..;-.. MILLSTONE RAOWASTE NEGLIGIBLE .... REDUCTION fACILITY ACTIVITY 6000 fT.5/ YR NOTES:
3
- 1. Ci FT IVR VALUES BASED UPON VOLUME OF RAW SPENT RESIN.
3
- 2. Ci FT /YR VALUES BASED UPON VOLUME OF PACKAGED WASTE.
- 3. ci FT3 /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 FT3 IYR OF RAW EVAPORATOR BOTTOMS. THIS METHOD WOULD NOT BE THE NORMAL OR PREFERRED METHOD OF DISPOSAl.
Rev. 21.3