ML17212A052
| ML17212A052 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/2017 |
| From: | Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17212A038 | List:
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| References | |
| 17-208 | |
| Download: ML17212A052 (151) | |
Text
Millstone Power Station Unit 2 Safety Analysis Report Chapter 11
Table of Contents tion Title Page RADIOACTIVE WASTE PROCESSING SYSTEMS ..................................... 11.1-1
.1 General...................................................................................................... 11.1-1
.2 Design Bases............................................................................................. 11.1-1
.2.1 Functional Requirements .......................................................................... 11.1-1
.2.2 Design Criteria .......................................................................................... 11.1-1
.2.3 System Components ................................................................................. 11.1-3
.3 Liquid Waste Processing System.............................................................. 11.1-3
.3.1 System Descriptions ................................................................................. 11.1-3
.3.2 System Operation...................................................................................... 11.1-7
.4 Gaseous Waste Processing System........................................................... 11.1-9
.4.1 System Descriptions ................................................................................. 11.1-9
.4.2 System Operation.................................................................................... 11.1-10
.4.3 Gaseous Release Radiological Consequences ........................................ 11.1-10
.4.4 Waste Gas System Failure ...................................................................... 11.1-10
.4.4.1 General.................................................................................................... 11.1-10
.4.4.2 Method of Analysis................................................................................. 11.1-11
.4.4.3 Results of Analysis ................................................................................. 11.1-11
.4.4.4 Conclusions............................................................................................. 11.1-11
.5 Solid Waste Processing System .............................................................. 11.1-11
.5.1 System Descriptions ............................................................................... 11.1-11
.6 System Reliability and Availability ........................................................ 11.1-14
.6.1 Special Features ...................................................................................... 11.1-14
.6.2 Tests and Inspections .............................................................................. 11.1-15
.7 References............................................................................................... 11.1-15 RADIATION PROTECTION ........................................................................... 11.2-1
.1 Final Safety Analysis Report Update........................................................ 11.2-1
.2 Design Bases............................................................................................. 11.2-1
.3 Description................................................................................................ 11.2-3
.3.1 Containment Shielding ............................................................................. 11.2-3
.3.2 Auxiliary Building Shielding.................................................................... 11.2-4
.3.3 Control Room Shielding ........................................................................... 11.2-5
.3.4 Spent Fuel Pool Shielding and Fuel Handling Shielding ......................... 11.2-5
.3.5 Piping Systems Shielding ......................................................................... 11.2-5
.4 Health Physics Program............................................................................ 11.2-6
.5 References................................................................................................. 11.2-6 11-i Rev. 35
tion Title Page A SOURCE TERMS FOR RADIOACTIVE WASTE PROCESSING AND RELEASES TO THE ENVIRONMENT ......................................................... 11.A-1 A.1 Reactor Coolant Design Basis Radionuclide Activities .......................... 11.A-1 A.1.1 Development of Reactor Core Radionuclide Activities .......................... 11.A-1 A.1.2 Corrosion Products .................................................................................. 11.A-2 A.1.3 Tritium Production................................................................................... 11.A-2 A.1.4 Fuel Experience ....................................................................................... 11.A-3 A.2 Reactor Coolant Expected Radionuclide Activities................................. 11.A-4 A.3 Calculation of Liquid and Gaseous Effluent Releases............................. 11.A-4 A.3.1 Expected Liquid and Gaseous Radioactive Effluent Releases ................ 11.A-4 A.3.2 Design Basis Liquid and Gaseous Radioactive Effluent Releases .......... 11.A-5 A.4 Solid Waste Processing System ............................................................... 11.A-5 A.4.1 Spent Resins............................................................................................. 11.A-6 A.4.1.1 Spent Resins from CVCS Ion Exchanger ................................................ 11.A-6 A.4.1.2 Spent Resins from Clean Liquid Waste Processing System Demineralizers ......
11.A-6 A.4.1.3 Spent Resins from Aerated Liquid Waste Processing System Demineralizer.....
11.A-6 A.4.1.4 Spent Resins from Spent Fuel Pool Demineralizer ................................. 11.A-6 A.4.1.5 Contaminated Filter Cartridges................................................................ 11.A-7 B RADIOACTIVE WASTE PROCESSING OF RELEASES TO ENVIRONMENT ...
11.B-1 B.1 Bases .........................................................................................................11.B-1 B.2 Liquid Waste Processing System..............................................................11.B-1 B.2.1 Processing of Clean Liquid Waste............................................................11.B-1 B.2.2 Processing of Aerated Liquid Waste ........................................................11.B-2 B.2.3 Processing of Secondary Side Liquid Waste ............................................11.B-2 B.3 Gaseous Waste Processing System...........................................................11.B-3 C DOSES FROM RADIOACTIVE RELEASES AND COST-BENEFIT ANALYSIS .
11.C-1 C.1 Doses to Humans ......................................................................................11.C-1 C.2 Methods for Calculating Doses From Liquid Releases ............................11.C-1 C.2.1 Generalized Equation for Calculating Radiation Doses to Humans via Liquid Pathways ...................................................................................................11.C-1 C.2.2 Doses from Aquatic Foods .......................................................................11.C-2 C.2.3 Doses from Shoreline Deposits.................................................................11.C-2 11-ii Rev. 35
tion Title Page C.2.4 Doses from Swimming and Boating .........................................................11.C-3 C.3 Method for Calculating Doses From Gaseous Releases ...........................11.C-3 C.3.1 Gamma and Beta Doses from Noble Gas Discharged to the Atmosphere .11.C-4 C.3.1.1 Annual Air Doses from Noble Gas Releases (Non-Elevated) ..................11.C-4 C.3.1.2 Annual Total Body Dose from Noble Gas Releases (Non-Elevated).......11.C-5 C.3.1.3 Annual Skin Dose from Noble Gas Releases (Non-Elevated) .................11.C-5 C.3.1.4 Annual Gamma Air Dose and Annual Total Body Dose Due to Noble Gas Releases from Free-Standing Stacks More Than 80 Meters Tall .............11.C-6 C.3.2 Doses from Radioiodines and Other Radionuclides, Exclusive of Noble Gases, Released to the Atmosphere .....................................................................11.C-6 C.3.2.1 Annual Organ Dose Due to External Irradiation from Ground Deposition of Radionuclides............................................................................................11.C-6 C.3.2.2 Annual Organ Dose from Inhalation of Radionuclides in Air..................11.C-6 C.3.2.3 Annual Organ Dose from Ingestion of Atmospherically Released Radionuclides in Food ......................................................................................................11.C-7 C.4 Comparison of Calculated Annual Maximum Individual Doses with Appendix I Design Objectives .....................................................................................11.C-7 C.5 General Expression for Population Doses ................................................11.C-8 C.6 Cost-Benefit Analysis ...............................................................................11.C-8 C.6.1 Procedure Used for Performing Cost-Benefit Analysis............................11.C-9 C.6.2 Augments to the Liquid Radioactive Waste Processing System ............11.C-10 C.6.3 Augments to the Gaseous Radioactive Waste Processing System .........11.C-10 D EXPECTED ANNUAL INHALATION DOSES AND ESTIMATED AIR CONCENTRATIONS OF RADIOACTIVE ISOTOPES FOR MP2 FACILITIES ....
11.D-1 E AIRBORNE ACTIVITY SAMPLING SYSTEM FOR CONTAINMENT, SPENT FUEL AND RADWASTE ATMOSPHERES...................................................11.E-1 11-iii Rev. 35
List of Tables mber Title
-1 Radioactive Waste Processing System Component Description
-2 Sources and Expected Volumes of Solid Wastes
-3 Radioactivity levels of solid wastes (See Note)
-4 Curie Inventory of Solid Waste Shipped from Millstone Unit 2 (See Note)
-5 Assumptions for Waste Gas Decay Tank Accident
-1 Source Terms for Shielding Design A-1 Design-Basis and Expected Primary Coolant Activity Concentrations A-2 Calculated Reactor Core Activities A-3 Expected Annual Effluent Releases (Curies Per Year), by Radionuclide, from Each Release Point A-4 Expected Annual Liquid Effluent Activity Releases (Curies/Year), by Radionuclide, from Each Waste Stream A-5 Expected Annual Liquid Effluent Concentrations (Diluted and Undiluted), by Radionuclide, from Each Waste System A-6 Design Basis Radionuclide Concentrations in Liquid Effluent, in Fractions of 10 CFR Part 20 Concentration Limits A-7 Design-Basis Radionuclide Airborne Concentrations at the Site Boundary From All Gaseous Effluent Release Points Combined, in Fractions of 10 CFR Part 20 Concentration Limits A-8 Total Annual Design-Basis and Expected Releases of Radioactive Liquid Waste to the Environment From All Sources Combined, in Curies Per Year A-9 Total Annual Design Basis and Expected Releases of Airborne Radioactive Waste to the Environment From All Release Points Combined, in Curies Per Year A-10 Basis for Reactor Coolant System Activity NUREG-0017 gale Code Input B-1 Inputs to PWR-GALE Code C-1 Comparison of Calculated Annual Maximum Individual Doses with 10 CFR Part 50 Appendix I Design Objectives C-2 Annual Total Body and Thyroid Doses to the Population Within 50 Miles of the Millstone Site, In Man-Rem, From Expected Liquid and Airborne Effluent Releases 11-iv Rev. 35
mber Title D-1 Containment Building Airborne Concentrations D-2 Auxiliary Building Airborne Concentrations D-3 Turbine Building Airborne Concentrations 11-v Rev. 35
List of Figures ure Title
-1 P&ID Clean Liquid Radwaste System
-2 P&ID Clean Liquid Radwaste System
-3 P&ID Drains (Containment & Auxiliary Building and Auxiliary Yard Sump)
(Sheet 1)
-4 P&ID Aerated Liquid Radwaste System
-5 P&ID Diagram Gaseous Radwaste System
-6 Degasifier Performance Curve
-7 P&ID Spent Resin Radwaste System
-1 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Elevation (-) 45 Feet 6 Inches
-2 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Elevation (-) 29 Feet 6 Inches
-3 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Elevation (-) 5 Feet 0 Inches
-4 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment Building - Elevation 14 Feet 6 Inches and 38 Feet 6 Inches
-5 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Auxiliary Building - Elevation 14 Feet 6 Inches and 25 Feet 6 Inches
-6 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Turbine Building - Elevation 14 Feet 6 Inches
-7 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Auxiliary Building - Elevation 36 Feet 6 Inches
-8 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Section A-A
-9 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary Building - Section B-B
-10 P&ID, Radiation Zones and Access Control Normal Operation With 1.0% Failed Fuel Containment and Auxiliary
-11 Neutron Shield Segment
-12 Neutron Shielding - Sectional View 11-vi Rev. 35
NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
ure Title 2-13 Not Used 2-14 Not Used
-15 Neutron Shield - Thermal Loadings (Typical Annular Section)
-16 Neutron Shield - Thermal Loadings (Plug Shield Section)
A-1 Escape Rate Coefficients B-1 NUREG-0017 Gale Code Input Diagram - Liquid Waste System B-2 NUREG-0017 Gale Code Input Diagram - Airborne 11-vii Rev. 35
RADIOACTIVE WASTE PROCESSING SYSTEMS
.1 GENERAL purpose of the radioactive waste processing systems is to provide controlled handling of all oactive waste, proper discharging of radioactivity in liquid and gaseous effluents, and proper kaging for shipment of solid waste containing radioactivity from Millstone Unit 2.
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).
.2 DESIGN BASES
.2.1 Functional Requirements se systems are required to ensure that the general public and plant personnel are protected inst exposure to radioactive material in accordance with the regulations of 10 CFR Part 20, the recommended guidelines of 10 CFR Part 50, Appendix I. Under normal plant operating ditions, the radioactive waste processing systems are required to limit radionuclide release centrations to unrestricted areas to less than the maximum permissible concentrations as cified in 10 CFR Part 20, Section 1302, and Appendix B.
rim onsite storage facilities accept waste from Millstone Units 1, 2 and 3. Information rding facility design criteria is presented in Section 11.4 of the Millstone Unit 3 Final Safety lysis Report.
.2.2 Design Criteria radioactive waste processing system is designed in accordance with the following criteria:
- a. To protect the general public and plant operating personnel against radiation from materials in accordance with 10 CFR Part 20 and 10 CFR Part 50, Appendix I.
- b. To limit the levels of radioactivity of the effluents in unrestricted areas to as low as reasonably achievable (ALARA).
- c. To provide suitable control of the release of radioactive materials in gaseous and liquid effluents during normal plant operation including anticipated operational occurrences.
11.1-1 Rev. 35
- e. To ensure adequate safety under normal and postulated accident conditions.
- f. To provide the capability to permit inspection and testing of appropriate components.
- g. To provide suitable shielding for radiation protection.
- h. To provide appropriate containment and confinement of radioactive materials.
- i. To provide appropriate monitoring capability to detect excessive radiation levels and to monitor effluent discharge paths.
- j. To provide suitable processing of liquid and gaseous radwaste generated in accordance with the following operating and design criteria:
- 1. Normal operation with expected primary and secondary side activities calculated in accordance with NUREG-0017 Rev. 1 (Reference 11.1-1) using the PWR-GALE code.
- 2. Normal operation with design basis reactor coolant activities based on 1%
failed fuel in accordance with Technical Specifications.
only portions of the radioactive waste processing system designated as Seismic Category I the containment penetration piping and isolation valves. However, the high pressure (150
) portions of the gaseous waste system, consisting of the compressors, decay tanks, rconnecting piping and valves (see Figure 11.1-5), have been designed and analyzed for mic Category I requirements as given in Section 5.8.
analysis had been performed, at the time of initial plant licensing, to determine the site ndary doses due to simultaneous failure of the entire radioactive waste processing system, luding those portions of gaseous waste processing system that were analyzed to Seismic egory I requirements. That analysis was predicated on the original design and operation of the oactive waste processing system. More recently, this chapter was updated to reflect changes to radioactive waste processing system design and operation. The results of this update are nded by those of the original analysis discussed in this subsection.
ure the of the radioactive waste processing system may release gaseous and liquid wastes into auxiliary building. The auxiliary building, a Seismic Category I structure, is designed to tain all liquids within the building.
refore, the total inventory of radioactivity within the gases contained in the radioactive waste cessing system, excluding high pressure portion of gaseous waste system, are assumed to be ased.
11.1-2 Rev. 35
er these conditions, the X/Q value of 9.6 x 10-5 sec/m3 is applicable. The doses at 625 meters e calculated by the method of Safety Guides 3 and 4. Based on the results of the original plant nsing analysis, the site boundary doses due to simultaneous failure of entire radioactive waste cessing system, excluding portions of gaseous waste system, are as follows:
mal Operation Based on 0.1% Failed Fuel Thyroid dose (rem) - 3.9 x 10-4 Whole body dose (rem) - 5.04 x 10-2 mal Operation Based on 1.0% Failed Fuel Thyroid dose (rem) - 4.35 x 10-3 Whole body dose (rem) - 5.78 x 10-1 ce the doses are approximately equal to or less than the limits given in 10 CFR Part 20, tions 105 and 106 and Appendix B (version prior to January 1, 1994), a Seismic Noncategory aste processing system is acceptable.
.2.3 System Components criptions of the radioactive waste processing system components are given in Table 11.1-1.
.3 LIQUID WASTE PROCESSING SYSTEM
.3.1 System Descriptions an Liquid Waste Processing System:
an liquid waste is normally tritiated, non-aerated, low conductivity liquid waste consisting marily of reactor coolant letdown and liquid waste collected from equipment leaks and drains certain valve and pump seal leaks. The clean liquid radioactive waste processing system is wn schematically in Figures 11.1-1 and 11.1-2. The design of the clean liquid radioactive te processing system is based upon processing of radioactive liquids postulated to be released m the reactor coolant system (RCS) during normal reactor operation with design basis reactor lant activities. Operating experience of nuclear power plants indicates that Millstone Unit 2 expect to continue operating with a percentage of fuel failure much less than the postulated gn basis of one percent. Nevertheless, the performance of the clean liquid waste processing em during normal operation is based on both expected and design basis primary side activities.
cussed in Appendix 11.A are the methodologies used to determine the expected and design basis onuclide activity concentrations in the reactor coolant.
clean liquid waste processing system is designed to support the processing of 14 RCS umes per year of reactor coolant waste. However, the quantity of wastes to be generated and 11.1-3 Rev. 35
ure 11.B-1.
clean liquid waste processing system is designed for the processing of reactor coolant wastes currently with the letdown flow from the chemical and volume control system (CVCS). This de of operation, at the maximum clean liquid waste processing flow of 132 gpm, sets the imum system flow rate.
ctor coolant is diverted to the clean liquid radioactive waste processing system from the CS when changes in RCS inventory or boron concentration are necessitated by startups, tdowns, fuel depletion, etc. Reactor coolant at a rate of 44 gpm to 132 gpm is let down from CVCS through a filter to reduce insoluble particulate, after which it flows to the clean liquid oactive waste processing system for further processing.
rces of clean liquid waste in the containment are collected in the primary drain tank. A heat hanger and pump are provided for cooling the primary drain tank content as well as the tent of the pressurizer quench tank. The contents of these tanks are maintained or cooled to
°F to minimize the carryover of radioactive moisture to the gaseous waste processing system to tank venting.
ipment drains, valve leak-offs, and relief valve discharges from components that are located he auxiliary building and that contain liquids with dissolved fission gases are collected in the ipment drain sump tank via the closed drain system, as shown in Figure 11.1-3. This design imizes the uncontrolled release of gaseous radioactivity to the atmosphere.
liquid contents of both the primary drain tank and the equipment drain sump tank are pumped the demineralizers to the coolant waste receiver tank, bypassing the degasifier. The ineralizers and degasifier are discussed below:
- 1. Degasifier Reactor coolant degassing is accomplished by diverting the letdown flow in the CVCS to the degasifier in the clean liquid waste system. Interconnecting piping and valves are provided for degassing the reactor coolant prior to cold shutdown of the reactor for refueling operations. The degasifier is placed in service, prior to cold shutdown, to remove hydrogen, fission product gases, and other dissolved gases from the reactor coolant system liquid and to discharge these gases to the waste gas surge tank in the gaseous waste processing system. (See Section 11.1.4)
The degassed liquid is pumped through the degasifier effluent cooler, which is used to lower the temperature of this liquid to 120°F before it is passed through one of two primary demineralizers. This is done to protect liquid radwaste system demineralizer resins against the damage caused by high liquid temperatures. The degassed reactor coolant is returned to either the volume control tank inlet or to the clean liquid radwaste system primary demineralizers for processing and discharge to the environment.
11.1-4 Rev. 35
pressure, is maintained within the degasifier during the shutdown mode to prevent air in-leakage and the formation of a potentially explosive hydrogen/oxygen mixture.
The degasifier system is provided with cascading steam controls to minimize the adverse effects of varying feed flow rates and temperatures on the dissolved gas removal performance of the unit. The degasifier operates at a pressure of 5 psig and a temperature of 228°F, and its pumps (both operating at full capacity) are controlled automatically by the degasifier level controls.
- 2. Demineralizers The two (2) primary and two (2) secondary demineralizers are of the mixed-bed non-regenerating type, the designs of which are based on an expected operating ion exchange capacity of 12,000 grains of CaCO3. The resin beds for mixed-bed non-regenerating demineralizers are mixtures of cation and anion resins in the H-OH forms. These demineralizers are bypassed automatically upon detection of liquid waste temperatures above 135°F. An additional secondary demineralizer provides the capability to further polish the waste stream, with the capability of placing the two (2) secondary demineralizers in series operation.
The demineralizers of the mixed-bed nonregenerable type are sized for one-year operation between resin replacement. Resin replacement is accomplished remotely by N2 pressure and/or water sluicing to the spent resin tank. All valves required for resin removal are located either outside the demineralizer compartments if non-radioactive, or inside shielded vaults with extension operators to provide protection of plant personnel if radioactive. The demineralizer effluent flows to one of two coolant waste receiver tanks.
two receiver tanks provide storage for approximately two RCS volumes (120,000 gal.
inal) of liquid wastes. A nitrogen gas blanket in each of the tanks is automatically maintained ve atmospheric pressure to prevent air in-leakage. The nitrogen cover gas is vented to the eous waste processing system or displaced into the other receiver tank or monitor tanks as id fills the first receiver tank. No flashing occurs in the receiver tanks, and any transfer of rogen or fission gases from the liquid to the cover gas occurs via the slow process of ecular diffusion.
content of the coolant waste receiver tank is sampled prior to processing. The content of the lant waste receiver tank is then pumped through the secondary demineralizer(s) to one of two lant waste monitor tanks, which offers a final check on the liquid waste to be released. The l storage capacity of the monitor tanks is approximately one RCS volume (60,000 gal.
inal).
11.1-5 Rev. 35
r sampling the monitor tank contents, then the liquid is reprocessed through the ineralizers. If the activity level is within discharge limits, the liquid is pumped through a final r, at a rate selectable over a range of 10 gpm to 132 gpm, to the circulating water system re it is diluted with the water in the discharge conduit. The proper rate of liquid release to the harge conduit is determined by sampling the liquid to be released. The concentrations of the ting isotopes for release are determined. Based on the circulating water flow rate, the harge rate is selected such that the releases to unrestricted areas are within permissible centrations as specified in 10 CFR Part 20, Appendix B. The second isolation valve in the te discharge header to the circulating water system is provided with a panel mounted manual troller for setting the desired flow rate. The clean liquid waste processing system effluent rs the circulating water system discharge box that runs along the south wall of the auxiliary ding, as shown in Figure 1.2-10. The routing of the discharge structure to the quarry is shown he plot plan, Figure 1.2-2. The circulating water is further diluted by the Long Island Sound l flow.
only liquid discharge path from the clean liquid waste processing system to the environment hrough the discharge header, which contains a radiation monitor. The radiation monitor and undant isolation valves are installed between the waste processing system and the circulating er system. This radiation monitor annunciates in the control room on high radioactivity level instrument failure, and will automatically close the isolation valves to prevent further harge. The single radiation monitor serves as a backup system to sampling of the liquid to be ased. All liquid to be released is sampled to confirm that the regulations of 10 CFR Part 20, tions 1301 and 1302, and Appendix B and 10 CFR Part 50, Appendix I are met. The radiation nitor is described in Section 7.5.6.
prevent the addition of waste to a monitor tank, the content of which is being discharged to the ulating water system, a valve interlock system is provided. The pneumatic valves, located in discharge piping from each tank, can be opened and kept open only if the corresponding valve he inlet piping to each tank is closed. This arrangement prohibits the addition of wastes to a viously sampled tank whose content is being released.
ated Liquid Waste Processing System:
design of the aerated liquid waste processing system, shown in Figure 11.1-4, is based upon processing of radioactive liquids (other than those handled by the clean liquid waste cessing system) postulated to be generated annually. However, the expected quantity of wastes e generated and processed annually by this system is approximately 600,000 gallons, based on assumptions used in the 10 CFR 50 Appendix I analysis.
aerated liquid waste system design is based upon the processing of wastes with a oactivity level resulting from normal reactor operation with design basis reactor coolant vity concentrations. The performance of the aerated liquid waste processing system during mal operation is evaluated based upon processing of wastes with radioactivity levels resulting 11.1-6 Rev. 35
ated liquid wastes are collected by the open drains system (shown in Figure 11.1-3) that ties into one of two drain tanks. The system is designed for batch processing. The contents of drain tanks are monitored by sampling and processed through a system that contains a filter, hard-piped demineralizer, and three portable demineralizers.
demineralizer effluent is collected in the aerated waste monitor tank, which offers a final ck on the liquid to be released. If the radioactivity level is within discharge limits, as rmined by sampling, the liquid is then pumped, at a rate selected over a range of 10 gpm to gpm, to the circulating water system, where it is diluted with water in the discharge conduit.
second isolation valve in the waste discharge header to the circulating water system is vided with a panel mounted manual controller for setting the desired flow rate. The liquid ased from the aerated liquid radioactive waste processing system is monitored continuously radioactivity. The radiation monitor annunciates in the control room on high radioactivity and rument failure, and will automatically close two isolation valves in the discharge piping to vent further discharge. The radiation monitor is discussed in Section 7.5.6. Provisions are ished for recycling the monitor tanks content for further processing.
only liquid discharge path from the aerated liquid waste processing system to the ironment is through the discharge pipe containing the radiation monitor. All system leakage, ns, and relief valve flows are collected in the drains system and returned to the aerated liquid te processing system.
.3.2 System Operation an Liquid Waste Processing Systems:
- 1. Normal Operation Letdown from the RCS is automatically diverted to the radioactive waste processing system on detection of a high volume control tank liquid level. The flow of wastes into the radioactive waste processing system, normally at a rate of 44 gpm, is intermittent due to changes in RCS inventory or boron concentration necessitated by startups, shutdowns, fuel depletion, etc.
All other liquid wastes input into the clean liquid waste processing system are intermittent due to plant operation. Liquid wastes are collected in either the equipment drain sump tank (EDST) or the primary drain tank. On high tank level, the equipment drain sump tank contents are pumped to the coolant waste receiver tank. During reactor shutdown for refueling, a greater quantity of waste is anticipated due to draining of system equipment for maintenance or inspection. The operation of the primary drain tank system, for collection of liquid wastes in the containment, is on a batch-type basis. On primary drain tank high tank liquid level, an alarm is annunciated in the main control room to alert the plant operator. The operation of the primary drain tank pumps and opening of the 11.1-7 Rev. 35
operating temperature of 120°F.
The processing of EDST and letdown liquid wastes up to the coolant waste receiver tanks is on a fully automatic, unattended basis. Processing up to the coolant waste receiver tanks consists of filtration, degasification (when needed), and ion exchange. Degassing the RCS prior to shutdown facilitates subsequent opening of the system.
The processing of wastes downstream of the coolant waste receiver tanks occurs on a batch-type basis. The receiver tank content is sampled to determine the amount of further processing required. A remotely actuated recirculation system is provided for each tank for the purpose of taking a representative sample. Normal processing consists of demineralization.
The processed wastes are collected in the monitor tanks, which offer a final check on the liquid to be released. Sampling is employed to determine the required discharge rate to ensure releases are within 10 CFR Part 20, Sections 1301 and 1302 and Appendix B and 10 CFR Part 50, Appendix I limits.
- 2. Abnormal Operation Abnormal operation of the clean liquid waste processing system may involve reactor operation with greater than expected failed fuel and larger than anticipated volume of liquid waste generated, such as for cold-shutdowns and startups late in the core cycle. Since the clean liquid radioactive waste processing system is designed for reactor operation with design basis reactor coolant activities for processing 14 RCS volumes of waste per year and with sufficient storage capacity, system operation for these abnormal occurrences will be the same as for normal operation described above.
ated Liquid Waste Processing Systems:
- 1. Normal Operation The aerated liquid waste processing system is operated on a batch type basis. System processing will vary with the volume of waste generated. The typical processing cycle is expected to be 5 to 7 days.
Normal processing consists of filtration and demineralization. The aerated waste monitor tank provides a final check on the liquid to be released to ensure compliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR 50 Appendix I limits. All potential releases are sampled prior to discharge to the environment.
11.1-8 Rev. 35
Abnormal operating occurrences considered are larger than expected volumes of wastes, such as for steam generator blowdown processing. If the waste generation rate exceeds the capacity of the system, interconnecting piping is provided for pumping the wastes to the clean liquid radioactive waste processing system for treatment.
.4 GASEOUS WASTE PROCESSING SYSTEM
.4.1 System Descriptions gaseous waste processing system, shown in Figure 11.1-5, processes potentially radioactive rogenated waste gases. The system design is based upon the processing of radioactive gases tulated to be released from the RCS during normal operation with design basis reactor coolant vities. The gaseous waste processing system is shown as one of the streams in Figure 11.B-2.
te gases flow to the waste gas header and are collected in the waste gas surge tank. When the e tank pressure increases to approximately 3 psig, one of the two 25 scfm compressors is matically started by pressure instrumentation located on the tank. The surge tank gases are pressed into one of the six waste gas decay tanks, where gases are stored at a maximum sure of 150 psig.
storage capacity of each waste gas decay tank is based on the storage of waste gases expected nter the system during any two months of normal operation. For all the waste gases entering system, including those contributed by operational occurrences, the six decay tanks provide quate storage capacity for a decay time of 90 days. Gases are held in the decay tanks until the oactivity level has been reduced by decay and the gases are suitable for release. Prior to ase, the gases in the decay tanks are sampled to determine compliance with the regulations of CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR Part 50, Appendix I.
decay tanks discharge through an absolute (i.e., HEPA) filter to the Millstone stack. The harge pipe contains a radiation monitor and redundant automatic isolation valves. A radiation nitor annunciates in the control room on high radioactivity level and instrument failure and automatically close the isolation valves to prevent any further release.
release rate for waste gases is selectable over a range of 10 scfm to 50 scfm. The rate is rmined by sampling the content of the decay tank to be discharged. The radionuclide centrations are correlated with the potential site boundary dose and the 10 CFR Part 20, endix B Effluent Concentrations, and the release rate is determined.
radiation monitor in the waste gas discharge header is provided as a backup system to the pling of the decay tank contents. The radiation monitor is described in FSAR Section 7.5.6.
pneumatic valve in the inlet to each decay tank is interlocked with the corresponding valve in tank discharge piping. The discharge valve can be opened and remains open only when the 11.1-9 Rev. 35
.4.2 System Operation
- 1. Normal Operation Waste gases from the sources shown in Figure 11.B-2 are collected in the waste gas surge tank through the waste gas header. As the pressure in the surge tank increases to 3 psig, one compressor is automatically started by pressure instrumentation mounted on the surge tank. If the surge tank pressure continues to increase, the second compressor is automatically started at 5 psig. The compressor discharges the waste gases into one of the six decay tanks selected by the operator. When the decay tank being filled reaches approximately 140 psig pressure, an alarm is annunciated in the control room to alert the operating personnel. The inlet valve to the decay tank is closed remotely by the operator, and one of the five remaining decay tanks is selected to receive the gases.
After an appropriate storage period, and after sampling has confirmed that 10 CFR Part 20, Sections 1301 and 1302 and Appendix B, and 10 CFR 50 Appendix I limits are met, the gases are released on a controlled batch basis to the Millstone stack.
- 2. Abnormal Operation The gaseous waste processing system, under abnormal conditions (e.g., unexpectedly large volumes of gases) will function in a fashion similar to that for normal conditions.
Suitable storage capability is available to store the cover gas from the coolant waste receiver and monitor tanks resulting from back-to-back cold shutdowns and startups late in core life.
.4.3 Gaseous Release Radiological Consequences ual design basis and expected releases of radioactive gaseous waste are presented in le 11.A-9. Gaseous release radiological consequences are presented in Appendix 11.C.
.4.4 Waste Gas System Failure
.4.4.1 General limiting accident considered is the postulated and uncontrolled release to the auxiliary ding of the radioactive xenon and krypton gases stored in one waste gas decay tank. The ibility of such an occurrence is low since the waste gas system is not subjected to pressures ter than 150 psig, or large stresses. The result of a rupture of a gas decay tank is analyzed in er that the maximum hazard, which would result from a malfunction in the radioactive waste cess system, will be defined.
11.1-10 Rev. 35
s assumed that the tank contains the gaseous activity evolved from degassing one system ume of reactor coolant for refueling. The maximum activity would exist prior to cold tdown at the end of an operating cycle during which extended operation with one percent ctive fuel had occurred. Based on this and neglecting decay after degasification, the noble gas vity in the tank is given in Table 11.1-5.
.4.4.3 Results of Analysis Dose (rems)
Organ Exclusion Area Boundary (EAB) LPZ yroid -- --
hole Body 6.4E-01 (1) 6.6E-02 (1)
The current waste gas system failure analysis is based on updated reactor coolant design activity contained in Table 11.A-1. This analysis results in a whole body dose of 3.0E-01 rem at the EAB and 4.0E-02 rem at the LPZ. The current results are bounded by the licensed values listed above.
.4.4.4 Conclusions waste gas decay tank rupture did occur, the dose would be substantially below 10 CFR Part guidelines.
.5 SOLID WASTE PROCESSING SYSTEM
.5.1 System Descriptions solid waste processing system is designed to provide controlled handling of spent resins, taminated filter cartridges, and miscellaneous solid waste. The system is designed for dling solid waste with radioactivity levels resulting from reactor operation with design basis tor coolant source terms. The sources and expected annual volumes of solid wastes are given able 11.1-2.
estimated isotopic curie inventory for each source of solid waste is given in Table 11.1-3. The vity data in Table 11.1-3 are based on analyses done for the original licensing of Millstone t 2 and are not derived from the updated radionuclide activities shown in the Appendices to FSAR chapter.
ign of the solid waste processing system for handling and disposal of each type of solid waste s follows:
11.1-11 Rev. 35
Spent resins from the radioactive waste processing system demineralizers, CVCS ion exchangers, and spent fuel pool demineralizer are replaced in accordance with plant procedures. Resin replacement is accomplished by sluicing the resins from the hard-piped demineralizers and ion exchangers with nitrogen and/or demineralized water to the spent resin tank. The portable demineralizers may be sluiced with an air/demineralized water mixture to the spent resin tank (SRT), or sluiced directly to a shipping cask, bypassing the SRT.
Spent resins are accumulated and stored in the spent resin tank for radioactive decay prior to filling the disposable container located in a shipping cask at elevation (-)45 feet, 6 inches in the auxiliary building. The spent resin tank is sized for storing the total volume of resins resulting from one resin replacement per year per demineralizer. With the storage capacity available, minimum storage time of about six months is expected for resins in the tank. The resins are dewatered by the use of a pump after placement in the disposable container.
Solid waste containers, shipping casks, methods of packaging, and transportation meet applicable federal regulations 10 CFR Part 71and 10 CFR 171-178, and wastes are buried at a licensed burial site in accordance with applicable NRC 10 CFR Part 61. Solid waste treatment design is in compliance with the requirements of 10 CFR Part 20, Sections 1301 and 1302 as it relates to radioactivity in effluents to unrestricted areas.
The spent resin radwaste system is shown in Figure 11.1-7.
Contaminated Filter Cartridges Filter cartridges for all radioactive service filters are normally replaced when the pressure drop across the filter unit exceeds 40 psi. All filters are located in concrete shielded compartments with access provided by a hatch located in the roof of the compartment.
(See FSAR Section 11.2 for shielding design.) For the removal of contaminated cartridges, the concrete hatch plug is removed first. Then contaminated cartridges are removed in accordance with plant procedures.
The contaminated cartridges are transferred to shielded containers, which are then capped and stored in the drumming area for ultimate off-site disposal.
Miscellaneous Solid Wastes Contaminated metallic materials and solid objects are placed in disposable shipping containers for transportation to an off site waste-processing facility or disposal site.
Miscellaneous compressible wastes, such as contaminated clothing, rags, paper, etc., are transported to the Millstone Radwaste Reduction Facility for compacting.
11.1-12 Rev. 35
osition can also be provided depending on curie content and container quantity.
ks used for shipment of filter cartridges and miscellaneous solid wastes are rented as needed.
containers are designed to prevent loss or dispersal of the contents and to maintain self-lding properties under normal conditions of transport. Where surface dose rates exceed those wed by the applicable Department of Transportation (DOT) regulations, the containers are vided with additional shielding. Where the curie content of the containers might exceed the ulatory limit applicable to this type of shipment, special casks licensed by DOT and by NRC, quired, would be used.
shipment of radioactive waste materials is governed by NRC regulations as set forth in CFR Part 71 and regulations of the U. S. DOT contained in 49 CFR Parts 170 through 178.
packaging and shipping of all waste from the Millstone site is in accordance with these ulations and any other applicable regulations that may come into effect. All loading of oactive waste containers is performed by qualified personnel and monitored by radiation ection personnel. The transportation is provided by a carrier authorized by the disposal site rator in accordance with NRC and DOT regulations. The material to be shipped from the site ent to a licensed waste disposal site or to a licensed waste-processing facility.
le 11.1-4 gives the total estimated annual curie inventory of solid wastes to be shipped from the on for off-site burial. This estimated inventory is based on analyses done for the original licensing Millstone Unit 2 and is not derived from the updated radionuclide activities shown in the endices to FSAR Chapter 11.
tem Operation
- 1. Normal Operation Spent resins from the sources listed in Table 11.1-3 are collected in the spent resin tank for radioactive decay. The spent resins are subsequently loaded into the shipping container and are dewatered by the spent resin shipping cask dewatering pump and portable dewatering pumps. After dewatering, the shipping container and cask are sealed and prepared for shipment.
Prior to a differential pressure of 40 psi being exceeded across any of the filters listed in Table 11.1-3, the filter is taken out of service, and allowed to decay. To change the filter cartridge, the concrete shield plug above the filter is removed. The cartridge is removed in accordance with plant procedures, placed in a shielded area and stored for shipment.
- 2. Abnormal Operation No abnormal operations are anticipated.
11.1-13 Rev. 35
.6.1 Special Features radioactive waste processing system design is based upon the processing of wastes postulated e generated from reactor operation with design basis reactor coolant activities and the release hese wastes in accordance with the requirements 10 CFR Part 20, Sections 1301 and 1302 and endix B. For expected radioactivity levels, sufficient processing capability is available to ure all releases from the radioactive waste processing system are in accordance with CFR Part 50, Appendix I.
radioactive waste processing system is designed for as low as reasonably achievable ARA) radioactivity releases to the environment. All radioactive waste processing system ts, equipment drains, leakage, valve stem leadoffs, and relief valve discharges are collected by drains system and reprocessed by the radioactive waste processing system. Those sources taining dissolved fission gases are collected in the equipment drain sump tank and processed lean liquid waste. All other sources are collected and processed by the aerated liquid waste em. All vent gases containing fission gases are discharged to the gaseous waste system for age and decay.
h liquid and gaseous waste processing system is provided with a single discharge path to the ironment. Each discharge header is provided with a radiation monitor and redundant isolation es which are closed on high radiation level or instrument failure to prevent releases not in pliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B.
releases from the radioactive waste processing system to the environment are to be omplished on a batch basis for suitable control. All wastes are first sampled to ensure pliance with 10 CFR Part 20, Sections 1301 and 1302 and Appendix B and 10 CFR Part 50, endix I. Adequate sample points are provided in the radioactive waste processing system for able control of the processing and to ensure that processing equipment performs as designed.
sampling points in the system are piped to the sampling room for analysis. Some local pling points are utilized in the aerated and clean liquid waste systems.
avoid inadvertent releases prior to sampling, the waste gas decay and coolant waste monitor s are provided with inlet and outlet valve interlocks. This control provision precludes the ning and filling of the tanks simultaneously.
storage tanks in the liquid waste system are provided with piping and valves for recirculating tank contents to obtain a representative sample. The valves are designed for remote operation rotect station operating personnel from radiation.
ficient storage capacity is provided for the retention of wastes in the radioactive waste cessing system. The three RCS volume storage capacity in the clean liquid radioactive waste cessing system is sufficient to allow for back-to-back cold shutdowns and startups up to 75 ent of the equilibrium core cycle. The six decay tanks allow for 60-day storage of all waste es, including cover gases from the clean liquid radioactive waste processing system tanks.
11.1-14 Rev. 35
ponent redundancy is provided to ensure adequate processing.
.6.2 Tests and Inspections components of the radioactive waste processing system are nondestructive tested in ordance with the applicable codes and standards as listed in Table 11.1-1. In addition to code uirements, additional testing is required for important components to ensure component grity and performance.
pumps in the radioactive waste processing system are manufacturer shop performance tested emonstrate compliance with design head and capacity requirements.
totype filter cartridge testing was conducted by the filter manufacturer confirming a filter ciency of 92 weight percent when tested with Fine Arizona Air Dust.
assembled degasifier package was shop gas leak tested to ensure leak-tightness of all ponents. The degasifier package was subject to a shop performance test to confirm design ulations and performance. Testing to confirm the performance in the removing of dissolved es was performed with the use of oxygen at feed rate of 60 gpm. Sufficient data were erated with oxygen testing to predict actual performance with respect to hydrogen, krypton xenon removal. The results of these tests are shown in Figure 11.1-6. The testing also firmed the capability of the degasifier to perform effective gas stripping under conditions of antaneous feed rate changes and varying feed temperatures.
system components are visually inspected and manually adjusted if necessary to ensure ect installation and arrangement. The completely installed system is subject to an acceptance in accordance with design requirements. The acceptance test is to check and/or calibrate ps, valves, instrumentation, interlocks, and system operation. In addition, the completely alled system will be checked for environmental considerations, such as correct routing of ns and vent connections, leakage, etc.
vidual system components are located so as to allow access for periodic inspection and testing r component decontamination. Major components are located in individual shielded rooms or partments to maintain access control.
.7 REFERENCES
-1 NUREG-0017 Rev. 1, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors, PWR-GALE Code.
11.1-15 Rev. 35
TABLE 11.1-1 RADIOACTIVE WASTE PROCESSING SYSTEM COMPONENT DESCRIPTION CLEAN LIQUID WASTE PROCESSING SYSTEM mary Drain Tank Pumps e Inline horizontal centrifugal with mechanical seals ntity 2 ign temperature (°F) 250 ign head (TDH) (feet) 150 ign capacity (gpm) 50 SH available (feet) 3 to 33 imum NPSH required (feet) 3 erial:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 tor 7.5 hp es and Standards ASME Section III Class 3 (1971) mic design class 2 ign integrated radiation dosage (rad) 106 asifier Pumps e Inline horizontal centrifugal with mechanical seals ntity 2 ign temperature (°F) 250 ign head (TDH) (feet) 115 ign capacity (gpm) 132 SH available (feet) 2 imum NPSH required (feet) 1.5 erial:
11.1-16 Rev. 35
Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 tor (hp) 7.5 e ASME Section III Class 3 (1971) mic design class 2 ign integrated radiation dosage (rad) 106 lant Waste Receiver Tank and Monitor Tank Pumps e Inline horizontal centrifugal with mechanical seals ntity 2 ign temperature (°F) 175 ign head (TDH) (feet) 115 ign capacity (gpm) 132 SH available (feet) 30 imum NPSH required (feet) 10 erial:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 tor (hp) 7.5 e ASME Section III Class 3 (1971) mic design class 2 ign integrated radiation dosage (rad) 106 ipment Drain Sump Tank Pumps e Vertical wet pit ntity 2 ign temperature (°F) 150 ign head (feet) 125 ign capacity (gpm) 50 11.1-17 Rev. 35
Case Type 316 stainless steel Impeller Type 316 stainless steel Shaft Type 316 stainless steel tor (hp) 10 e Hydraulic Pump Institute mic design class 2 ign integrated radiation dosage (rad) 106 mary Drain Tank and Quench Tank Cooler Pump e Inline horizontal centrifugal with mechanical seals ntity 1 ign temperature (°F) 325 ign head (TDH) (feet) 125 ign capacity (gpm) 100 SH available (feet) 3.3-50 imum NPSH required (feet) 3 erial:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 tor (hp) 10 e ASME Section III Class 3 (1971) mic design class 2 ign integrated radiation dosage (rad) 106 mary Drain Tank e Horizontal ntity 1 ume (gallons net) 1,500 ign pressure (psig) 20 11.1-18 Rev. 35
rating pressure (psig) 3 rating temperature (°F) 120-150 erial ASTM A-240 TP 304 ign code ASME Section III Class 3 (1971) mic design classification 2 lant Waste Receiver Tanks e Vertical ntity 2 ume (gallons net) 60,000 ign pressure (psig) 15 ign temperature (°F) 150 rating pressure (psig) 3 rating temperature (°F) 120 erial ASTM A-240 TP 304 ign code:
ASME Section III Class C (1968 Edition including addenda through Summer 1970) mic design classification 2 lant Waste Monitor Tanks e Vertical ntity 2 ume (gallons net) 30,000 ign pressure (psig) 15 ign temperature (°F) 150 rating pressure (psig) 3 rating temperature (°F) 120 erial ASTM A-240 TP 304 ign code:
ASME Section III Class C (1968 Edition including Addenda through Summer 1970) 11.1-19 Rev. 35
ipment Drain Sump Tank e Vertical ntity 1 ume (gallons net) 500 ign pressure (psig 20 ign temperature (°F) 150 rating pressure (psig) 3 rating temperature (°F) 120 erial ASTM A-240 TP 304 ign code ASME Section III Class 3 (1971) mic design classification 2 asifier e Packed column utilizing internally generated stripping steam.
ntity 1 ign pressure (psig) 50 ign temperature (°F) 250 rating pressure (psig) 5.3 rating temperature (°F) 228 ign capacity (gpm) 40 to 132 ormance See Figure 11.1-6 ign codes:
Components containing radioactive material:
ASME Section III Class C (1968 Edition including addenda through summer 1970), ANSI B31.7 Class III, TEMA R.
Other components:
ASME Section VIII, ANSI B31.1 umn:
Packing One inch Rashig rings.
Height (feet) 4.25 11.1-20 Rev. 35
erials:
Components containing radioactive material: ASTM A-240 TP 304 Other components: Carbon steel mic design classification 2 lant Waste- Demineralizers (Primary and Secondary) e Mixed-bed, non-regenerative ntity 4 (See Note 3) ign pressure (psig) 100 ign temperature (°F) 150 rating pressure (psig) 60 rating temperature (°F) 120 ign flow (gpm) 132 ign code ASME Section III Class 3 (1971) erial ASTM A-240 TP 304 mic design classification 2 lant Waste-Filters e Disposable cartridge ntity 2 ign pressure (psig) 200 ign temperature (°F) 250 rating pressure (psig) 60 rating temperature (°F) 120 ign flow rate (gpm) 132 rating flow (gpm)40-132 er rating (micron) 3 ign filter efficiency (%) 80 ign code:
ASME Section III Class C (1968 Edition including Addenda through summer 1970).
11.1-21 Rev. 35
Vessel ASTM A-312 TP 304 Internals TP 304 SS, Micarta Cartridges Ethylene propylene mic design classification 2 lant Waste-Piping and Valves ng:
Material ASTM A-312 TP-304 or 316/316L Design pressure (psig) 50, 100, 150 Design temperature (°F) 150, 250, 300, 350 Joints 2.5 inches and larger Butt welded except at flanged equipment.
Joints 2 inches and smaller Socket welded except at flanged equipment **
es:
Fabrication ANSI B31.7 Class III
- Testing and Installation ASME Section III Class 3 (1971) *, **
ves: ASTM A-182 F304, F316; ASTM A-351 CF8, CF-8M Ratings:
2.5 inches and larger 150 lb ANSI 2 inches and smaller 600 lb ANSI ***
e:
ASME Draft for Pumps and Valves for Nuclear Service (1968) *, **
Portions of the Clean Liquid Waste Processing System have been replaced with piping and piping components designed, constructed and tested to the ANSI B31.1 Power Piping Code with augmented requirements per the guidance in Regulatory Guide 1.143.
Portions of the Clean Liquid Radwaste System have been designed, constructed and tested to the ANSI B31.1 Piping Code with augmented quality requirements per Regulatory Guide 1.143.
600 pound ANSI rating represents minimum requirements. 800 pound ANSI rating valves are utilized on a case-by-case basis.
11.1-22 Rev. 35
AERATED LIQUID WASTE PROCESSING SYSTEM ated Waste Drain and Monitor Tanks e Vertical ntity 3 ume (gallons net) 5,000 ign pressure Atmospheric ign temperature (°F) 150 rating pressure Atmospheric rating temperature (°F) 120 erial ASTM A-240 TP 304 ign codes ASME Section III - ASME Section VIII (Original Fabrication) mic design classification 2 ated Waste Demineralizer e Mixed-bed, non-regenerative ntity 1 ign pressure (psig) 100 ign temperature (°F) 150 rating pressure (psig) 60 rating temperature (°F) 120 ign flow (gpm) 132 mal operating flow (gpm) 50 ign code ASME Section III Class 3 (1971) erial ASTM A-240 TP 304 mic design classification 2 ated Waste Portable Demineralizers e Various, non-regenerative, sluicible ntity 3 ign Pressure 150 psig at 180°F 11.1-23 Rev. 35
ers e Disposable cartridge ntity 2 ign pressure (psig) 200 ign temperature (°F) 250 rating pressure (psig) 60 rating temperature (°F) 120 ign flow rate (gpm) 132 rating flow (gpm)40-132 er rating (micron) 3 ign filter efficiency (%) 80 ign code:
ASME Section III Class C (1968 Edition including Addenda through summer 1970).
mic design 2 Aerated Waste Drain Tank Pump e Inline horizontal centrifugal with mechanical seals ntity 1 ign temperature (°F) 180 ign head (TDH) (feet) 238 ign capacity (gpm) 90 SH available (feet) 27 imum NPSH required (feet) 9 sepower (hp) 15 erial:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 es and standards ASME Section III Class 3 (1971) 11.1-24 Rev. 35
ign integrated radiation dosage (rads) 106 Aerated Waste Drain Tank and Monitor Tank Pumps e Inline horizontal centrifugal with mechanical seals lity 2 ign temperature (°F) 150 ign head (TDH) (feet) 135 ign capacity (gpm) 50 SH available (feet) 27 imum NPSH required (feet) 5 sepower (hp) 5 erial:
Case ASTM A-351 Gr CF8M Impeller ASTM A-351 Gr CF8M Shaft ASTM A-276 TP 316 es and standards ASME Section III Class 3 (1971) mic design classification 2 ign integrated radiation dosage (rads) 106 ng and Valves ng:
Material ASTM A-316 TP 304 or 316L Design pressure (psig) 50 and 100 Design temperature (°F) 150 Joints:
2.5 inches and larger Butt welded except at flanged equipment 2 inches and smaller Socket welded except at flanged equipment es:
Fabrication ANSI B31.7 Class III
- Testing and Installation ASME Section III Class 3 (1971)
- 11.1-25 Rev. 35
Material ASTM A-182 F 304, F316 Ratings:
2.5 inches and larger 150 lb ANSI 2 inches and smaller 600 lb ANSI, 800 lb ANSI (Intermediate Class) e ASME Draft for Pumps and Valves for Nuclear Service (1968)
- tions of the Aerated Liquid Waste Processing System have been replaced with piping and ng components designed, constructed and tested to the ANSI B31.1 Power Piping Code with mented requirements per the guidance in Regulatory Guide 1.143.
11.1-26 Rev. 35
GASEOUS WASTE PROCESSING SYSTEM ste Gas Compressors e Diaphragm, single state ntity 2 acity (scfm) 25 at 14.7 psia suction pressure ign discharge pressure (psig) 150 ign pressure (psig) 165 erials:
Heads - ASTM A-105 Grade 2, with ASTM A-240 TP 304 in contact with waste gases.
Diaphragm TP 301 stainless steel or 316 stainless steel Support cylinder ASTM A-105 Grade 2 tor (hp) 40 mic design classification See Note 1 ign integrated radiation level (rads) 3.5 x 105 e:
ASME Section VIII, ASME Draft for Pump and Valves for Nuclear Service, ANSI B31.7.
te Gas Surge Tank ntity 1 e Vertical ign pressure (psig) 20 ign temperature (°F) 150 mal operating pressure (psig) 3 to 5 mal operating temp (°F) 150 ume (ft3) 582 erial ASTM A-240 TP 304 mic design classification 2 e:
ASME Section III Class C (1968 Edition including Addenda through 1970).
11.1-27 Rev. 35
e Vertical ntity 6 ign pressure (psig) 165 ign temperature (°F) 150 mal operating pressure (psig) 5 to 150 mal operating temp (°F) 120 ume (ft3) 582 erial ASTM A-515 Grade 70 mic design classification See Note 1 e:
ASME Section III Class C (1968 Edition including Addenda through summer 1970).
te Gas Filter e Disposable cartridge with HEPA filter and demister ntity 1 ign flow (scfm) 50 mal operating flow (scfm) 5 - 25 ign pressure (psig) 165 ign temperature (°F) 150 mal operating pressure (psig) 4 mal operating temp (°F) 120 ign efficiency 99.97% of particles 0.3 micron and larger erial:
Filter housing ASTM A-240 TP 304 Internals TP 304 stainless steel Cartridge Glass fiber e ASME Section III Class 3 (1971) mic design classification See Note 1 ng and Valves 11.1-28 Rev. 35
Material ASTM A-106 Grade B Design pressure (psig) 50, 150 and 200 Design temperature (°F) 150, 200 and 300 Joints:
2.5 inches and larger Butt welded except at flanged equipment.
2 inches and smaller Socket welded except at flanged equipment.
es:
Fabrication ANSI B31.7 Class III Testing and Installation ASME Section III Class 3 (1971)
Design seismic classification 2 (See Note 1) ves:
Materials ASTM A-216, Grade WCB Ratings:
2.5 inches and larger 150 lb ANSI butt weld ends 2 inches and smaller 3,000 lb ANSI socket weld ends e: ASME Draft for Pup and Valves for Nuclear Service (1968).
11.1-29 Rev. 35
SOLID WASTE PROCESSING SYSTEM nt Resin Storage Tank e Vertical ntity 1 ume (ft3) 380 ign pressure (psig) 75 rating pressure (psig) 35 ign temperature (°F) 175 rating temperature (°F) 120 erial ASTM A-240 TP 304 ign code ASME Section III Class 3 (1971) mic design classification 2 nt Resin Shipping Cask Dewatering Pump e Horizontal Centrifugal ntity 1 acity (gpm) 10 d (feet) 92 erial Stainless steel e 1: The high pressure components of the gaseous waste processing system are designated as Seismic Class 2 equipment. However, the components are designed to meet Seismic Class 1 loadings due to the nature of the equipment service.
e 2: The supply lines to the solidification concentrates pump were cut in accordance with PDCR 2-54-95.
e 3: Secondary Demineralizer may be filled with either ion-specific resin or mixed bed resin.
11.1-30 Rev. 35
Waste Generating Quantity Per Source Operation Year Per Year ent resins One resin replacement per 351 cu ft 7*
demineralizer/year ntaminated filter cartridges One cartridge replacement 12 12 **
per filter/year cartridges scellaneous solid wastes One 55 gallon drum/week - 52 **
- 50 cubic foot containers
- 55 gallon drums ated Waste hardpiped demineralizer at 42 cu. ft. x 1 = 42 cubic feet ated Waste sluicible portable demineralizers at 15 cubic feet/each x 3 = 45 cubic feet nt Fuel Pool demineralizer at 42 cubic feet x 1 = 42 cubic feet mary Liquid Radwaste demineralizers at 42 cubic feet x 2 = 84 cubic feet ondary Liquid Radwaste demineralizer at 42 cubic feet x 1 = 42 cubic feet down Ion Exchangers at 32 cubic feet x 3 = 96 cubic feet 351 cubic feet e: The data in this Table is not derived from the updated 10 CFR 50 Appendix I analysis. The best source of Solid Waste Volumes can be found in the Annual Radiological Effluent Release Report.
11.1-31 Rev. 35
ION EXCHANGER AND DEMINERALIZER RESINS Chemical and Volume Chemical and Volume Chemical and Volume Control System Control System Control System Primary Clean Liquid Purification Ion Purification Ion Deborating Ion Waste Demineralizer Aerated Liquid Waste Spent Fuel Po Exchanger Resin Exchanger Resin Exchanger Resin Demineralizer Resin Demineralizer Resin up Deminerali Isotope (Curies) (Li Removal) (Curies) (Curies) (Curies) (Curies) (Curies Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal M Cr-51 Mn-54 Mn-56 Co-58 Fe-59 Co-60 4.78 x 10- 0.478 9.6 x 10-3 9.6 x 10-2 1.2 x 10-3 1.2 x 10-2 1.12 x 10- 1.9 x 10-3 8.72 x 10- 1.01 x 10-2 1.23 12 2 4 5 Rb-88 1.366 13.66 0.246 2.46 -- -- 3.1 x 10-3 5.22 x 10-2 2.4 x 10-3 0.279 67.2 67 Rb-89 3.06 x 10- 0.306 5.5 x 10-3 5.5 x 10-2 -- -- 6.9 x 10-5 1.17 x 10-3 5.38 x 10- 6.25 x 10-3 1.69 16 2 5 Sr-89 13.8 138.0 2.398 23.98 -- -- 3.11 x 10- 0.525 2.41 x 10- 2.80 0.134 1.
2 2 Sr-90 3.052 30.52 0.471 4.71 -- -- - 0.114 6.77 x 10 5.25 x 10- 0.610 6.88 x 10-3 6.
3 3 Y-90 -- -- 2.24 x 10- 0.224 -- -- 4.31 x 10- 7.26 x 10-4 3.34 x 10- 3.88 x 10-3 2.69 x 10-2 0.
2 5 5 Sr-91 6.9 x 10-2 0.691 1.26 x 10- 0.126 -- -- 1.57 x 10- 2.65 x 10-3 1.22 x 10- 1.41 x 10-2 9.39 x 10-2 0.
2 4 4 11.1-32 Rev
ION EXCHANGER AND DEMINERALIZER RESINS Chemical and Volume Chemical and Volume Chemical and Volume Control System Control System Control System Primary Clean Liquid Purification Ion Purification Ion Deborating Ion Waste Demineralizer Aerated Liquid Waste Spent Fuel Po Exchanger Resin Exchanger Resin Exchanger Resin Demineralizer Resin Demineralizer Resin up Deminerali Isotope (Curies) (Li Removal) (Curies) (Curies) (Curies) (Curies) (Curies Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal M Y-91 -- -- 7.71 x 10- 0.771 -- -- 1.48 x 10- 2.5 x 10-3 1.15 x 10- 1.33 x 10-2 2.93 29 2 4 4 Zr-95 Mo-99 -- -- 52.46 524.6 -- -- 0.101 1.70 7.82 x 10- 9.08 53.5 53 2
Ru-103 7.84 78.4 1.57 x 10- 0.157 -- -- 1.51 x 10- 0.255 1.17 x 10- 1.36 0.109 1.
2 2 2 Ru-106 2.02 20.2 0.363 3.63 -- -- 4.58 x 10- 7.72 x 10-2 3.55 x 10- 0.412 6.54 x 10-3 6.
3 3 I-129 9.36 x 10- 9.36 x 10-3 1.56 x 10- 1.56 x 10-3 -- -- 2.1 x 10-6 3.54 x 10-5 1.63 x 10- 1.89 x 10-4 1.9 x 10-6 1.
4 4 6 Te-129 5.42 x 10- 0.542 1.09 x 10- 0.109 1.3 x 10-3 1.3 x 10-2 1.28 x 10- 2.15 x 10-3 9.88 x 10- 1.15 x 10-2 0.662 6.
2 2 4 5 I-131 1,680.0 16,800.0 336.0 3,360.0 19.2 192.0 3.91 65.9 3.03 352 105 1, I-132 5.04 50.4 1.007 10.07 0.121 1.21 1.19 x 10- 0.20 9.19 x 10- 1.07 26.9 26 2 3 Te-132 49.3 493.0 9.86 98.6 0.944 9.44 0.116 1.95 8.96 x 10- 10.4 8.7 87 2
I-133 247.0 2,470.0 4.94 49.4 5.91 59.1 0.496 8.35 0.384 44.6 149 1, Cs-134 388.0 3,800.0 517.0 5,170.0 -- -- 1.74 29.3 1.35 156 2.64 26 11.1-33 Rev
ION EXCHANGER AND DEMINERALIZER RESINS Chemical and Volume Chemical and Volume Chemical and Volume Control System Control System Control System Primary Clean Liquid Purification Ion Purification Ion Deborating Ion Waste Demineralizer Aerated Liquid Waste Spent Fuel Po Exchanger Resin Exchanger Resin Exchanger Resin Demineralizer Resin Demineralizer Resin up Deminerali Isotope (Curies) (Li Removal) (Curies) (Curies) (Curies) (Curies) (Curies Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal Maximum Normal M I-134 1.06 10.6 0.212 2.12 2.54 x 10-2 0.254 2.49 x 10- 4.2 x 10-2 1.93 x 10- 0.225 16.3 16 3 3 Te-134 3.19 x 10- 0.319 6.4 x 10-3 6.4 x 10-2 7.65 x 10-4 7.65 x 10-4 7.5 x 10-5 1.26 x 10-3 5.82 x 10- 6.76 x 10-3 0.691 6.
2 3 I-135 35.9 359.0 7.24 72.4 0.869 8.69 8.46 x 10- 1.43 6.56 x 10- 7.62 71.2 71 2 2 Cs-136 -- -- 18.12 181.2 -- -- 3.48 x 10- 0.587 2.7 x 10-2 3.14 0.672 6.
2 Cs-137 1,695.0 16,950.0 1,728.0 17,280.0 -- -- 6.58 111 5.10 593 8.44 84 Cs-138 -- -- 0.14 1.40 -- -- 2.69 x 10- 4.54 x 10-3 2.09 x 10- 2.42 x 10-2 18.2 18 4 4 Ba-140 4.08 40.8 0.743 7.43 -- -- 9.27 x 10- 0.156 7.19 x 10- 0.835 0.161 1.
3 3 La-140 0.469 4.69 9.37 x 10- 0.937 -- -- 1.08 x 10- 1.82 x 10-2 8.38 x 10- 9.74 x 10-2 0.154 1.
2 3 4 Pr-143 3.49 34.9 0.697 6.97 -- -- 8.05 x 10- 0.136 6.24 x 10- 0.725 0.154 1.
3 3 Ce-144 26.2 262.0 5.24 52.4 -- -- 6.04 x 10- 1.02 4.68 x 10- 5.44 0.109 1.
2 2 11.1-34 Rev
Chemical and Volume Reactor Vessel Hea Control System Filter Clean Liquid Waste Filter Aerated Liquid Waste Spent Fuel Pool Clean-up Decontamination Syst Cartridges Cartridges Filter Cartridges Filter Cartridges Filter Cartridges Isotope (Curies/Cartridges) (Curies/Cartridge) (Curies/Cartridges) (Curies/Cartridge) (Curies/Cartridges)
Maximu Normal Maximum Normal Maximum Normal m Normal Maximum Normal Maximu Cr-51 0.209 13.05 0.001 0.043 -- 0.452 0.003 0.185 0.039 0.039 Mn-54 0.935 0.67 0.002 0.002 0.004 0.024 0.004 0.003 0.028 0.028 Mn-56 -- 0.486 -- 0.002 -- 0.018 -- -- -- --
Co-58 64.0 107.0 0.123 0.347 0.192 3.702 0.50 0.84 4.78 4.78 Fe-59 0.192 0.069 0.001 0.001 -- 0.002 0.002 0.001 0.022 0.022 Co-60 0.025 61.0 -- 0.166 -- 1.766 0.068 0.142 0.533 0.533 Br-84 Rb-88 Rb-89 Sr-89 Sr-90 Y-90 Sr-91 Y-91 Zr-95 -- -- -- -- -- -- -- -- -- --
Mo-99 Ru-103 Ru-106 11.1-35 Rev
Chemical and Volume Reactor Vessel Hea Control System Filter Clean Liquid Waste Filter Aerated Liquid Waste Spent Fuel Pool Clean-up Decontamination Syst Cartridges Cartridges Filter Cartridges Filter Cartridges Filter Cartridges Isotope (Curies/Cartridges) (Curies/Cartridge) (Curies/Cartridges) (Curies/Cartridge) (Curies/Cartridges)
Maximu Normal Maximum Normal Maximum Normal m Normal Maximum Normal Maximu I-129 Te-129 I-131 I-132 Te-132 I-133 Cs-134 I-134 Te-134 I-135 Cs-136 Cs-137 Cs-138 Ba-140 La-140 Pr-143 Ce-144 Note:The activity data in this Table are based on analysis done for the original licensing of Millstone Unit 2 and are not derived from th updated radionuclide activities shown in the Appendices to FSAR Chapter 11.
11.1-36 Rev
MILLSTONE UNIT 2 (SEE NOTE)
Source Normal Curies per Year Maximum Curies per Year sins 4,215 42,871 ter Cartridges 35 149 ncentrates Less than 1 5 TAL 4,250 43,025 e: The activity data in this Table are based on analysis done for the original licensing of Millstone Unit 2 and are not derived from the updated radionuclide activities shown in the Appendices to FSAR Chapter 11. The best source of Solid Waste data can be found in the Annual Radiological Effluent Release Report.
11.1-37 Rev. 35
TABLE 11.1-5 ASSUMPTIONS FOR WASTE GAS DECAY TANK ACCIDENT umption (1) Maximum Noble Gas Activity in Waste Gas Decay Tank.
Isotope Activity, curies Kr-85m 278 Kr-85 942 Kr-87 160 Kr-88 494 Xe-131m 486 Xe-133m 708 Xe-133 45,600 Xe-135m 26.2 Xe-135 1,280 Xe-137 28.9 Xe-138 101 is: Maximum activity in tanks based on 1% degraded fuel, core power = 2700 MWt and degassing one system volume.
umption (2) 2 Hour Ground Level Release.
is: Regulatory Guide 1.24 umption (3) Ground Level X/Q:
EAB hr. = 3.66 E -04 (sec/m3)
LPZ 4.80 E -05 (sec/m3) is: 95% maximum X/Qs during the years 1974 -1981.
11.1-38 Rev. 35
FIGURE 11.1-1 P&ID CLEAN LIQUID RADWASTE SYSTEM figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-39 Rev. 35
FIGURE 11.1-2 P&ID CLEAN LIQUID RADWASTE SYSTEM figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-40 Rev. 35
IGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 1) figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-41 Rev. 35
IGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 2) figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-42 Rev. 35
IGURE 11.1-3 P&ID DRAINS (CONTAINMENT & AUXILIARY BUILDING AND AUXILIARY YARD SUMP) (SHEET 3) figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-43 Rev. 35
FIGURE 11.1-4 P&ID AERATED LIQUID RADWASTE SYSTEM figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-44 Rev. 35
FIGURE 11.1-5 P&ID DIAGRAM GASEOUS RADWASTE SYSTEM figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-45 Rev. 35
FIGURE 11.1-6 DEGASIFIER PERFORMANCE CURVE 11.1-46 Rev. 35
FIGURE 11.1-7 P&ID SPENT RESIN RADWASTE SYSTEM figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
11.1-47 Rev. 35
.1 FINAL SAFETY ANALYSIS REPORT UPDATE hould be noted that the information in this section provided the basis and description of the lding design prior to operation. As the station has operated, dose rates have fluctuated within various areas as a function of time but for the most part have remained well within the zone gnation limits presented below. Minor changes in equipment layout from that indicated in ures 11.2-1 through 11.2-10 have been made, but since the values and figures given in this ion provided the design basis for the major shielding structures, they are not being changed.
ny minor shielding changes (mostly through the use of portable shielding) have been rporated in the interest of maintaining occupational exposures ALARA. The only major lding change was the installation of a reactor cavity neutron shield.
values and figures given in this section should not be used to estimate area dose rates. Actual lth physics surveys are available and should be consulted to get a more accurate and up-to-date cation of radiological conditions.
endices 11.D and 11.E provide some of the design basis considerations on the control and ected levels of airborne radioactivity in the plant. Again, health physics survey data should be sulted for more realistic indications of actual airborne radioactivity levels. The occupational osure due to airborne radioactivity has been insignificant in comparison to the direct radiation osure.
.2 DESIGN BASES shielding is designed to perform two primary functions: to ensure that during normal ration the radiation dose to operating personnel and to the general public is within the ation exposure limits set forth in 10 CFR Part 20 and to ensure that operating personnel are quately protected in the event of a reactor incident so that the incident can be terminated hout undue hazard to the general public. The shielding design is based on operating at design er level of 2700 MWt with reactor coolant system activity levels corresponding to one ent failed fuel. The shielding design is governed by the limits for radiation levels as follows:
Zone Description Maximum Dose Rate (mrem/hr) ncontrolled, unlimited access (Zone A) 0.5 ontrolled, unlimited (40 hr/week) (Zone B) 1.0 ontrolled, limited access (6-2/3 hr/week) (Zone C) 15 ontrolled, limited access (1 hr/wk) (Zone D) 100 ormally inaccessible (Zone E) > 100 11.2-1 Rev. 35
imum external dose rates are based on one percent failed fuel. Since the expected amount of ed fuel is much less than this, the average external dose rates will be much lower. The estricted areas in the plant where construction workers and plant visitors may be are all ation Zone A areas which have a maximum dose rate of 0.5 mrem/hr. Only plant workers are wed in radiation zones with dose rates above this level, and their stay times are increasingly ted as the dose rate increases.
anticipated average dose rates will vary within each zone and with operating conditions. In eral, the dose rate normally around process equipment is expected to be at least an order of nitude lower than indicated by the applicable zone. This is due to the extremely conservative mptions made concerning the quantity of failed fuel. The direct dose at the site boundary is ligible both in the case of normal operation and under postulated incident conditions.
elding wall thicknesses, as shown in Figures 11.2-1 through 11.2-10, were found using the ry of the Reactor Shielding Design Manual by T. Rockwell, Reactor Physics Constants L-5800, Table of Isotopes by Perlman and Lederer, Engineering Compendium on Radiation elding and Nuclear Engineering Handbook by H. Etherington. The geometric considerations e determined by each physical situation, such as source configurations and distances to dose nts.
umptions used included those of TID 14844 and the following:
- 1. If the source was liquid, then self-attenuation by water was used. If the source was gaseous, then no self-attenuation was assumed. If the source was solid, the volume percentages of each solid were used for self-attenuation (for example, solidified waste in drums).
- 2. Shields included the tank or pipe walls which were assumed to have an attenuation coefficient similar to that for iron.
- 4. Sources were given in Mev/cm3 sec for each of 7 standard energy bins: 0.4, 0.8, 1.3, 1.7, 2.2, 2.5, > 2.5 Mev, and 6.1 and 7.1 Mev for nitrogen 16. One percent failed fuel was assumed.
- 5. Flux to dose conversion factors used were those given by Henderson in XDC-59-8-179.
- 6. The shield thicknesses were determined by the amount of shielding required to achieve a dose rate less than the upper limit for the radiation zone in which the dose point was located.
11.2-2 Rev. 35
.3 DESCRIPTION elding throughout Millstone Unit 2 is designed in accordance with the criteria specified in tion 11.2.1. Figures 11.2-1 through 11.2-10 show the equipment placement, the quantity of lding used and the appropriate radiation zones. All components containing radioactive fluids shown on the figures except the refueling water storage tank (RWST), located as shown on ure 1.2-2. The maximum direct dose rate from the content of the RWST is 5.6 x 10-3 mr/hr, ed on reactor operation with one percent failed fuel. The maximum total site boundary dose to the RWST is approximately 1.66 mrem/yr. For normal operating conditions, the site ndary doses will be approximately one-tenth of those values given above.
.3.1 Containment Shielding containment shielding consisting of the primary shielding, the reactor cavity neutron lding, the secondary shielding, and the containment wall, is shown on Figure 1.2-6.
mary shielding is provided to limit radiation emanating from the reactor vessel. The primary lding is designed to:
- a. Attenuate the neutron flux to limit the activation of components and structural steel.
- b. Limit the radiation level after shutdown to permit access to the reactor coolant system equipment.
- c. To reduce, in conjunction with the secondary shield, the radiation level from sources within the reactor vessel to allow limited access to the containment during normal operation.
primary shield consists of a minimum of five feet of reinforced concrete surrounding the tor vessel. The cavity between the primary shield and the reactor vessel insulation is air-led to prevent overheating and dehydration of the concrete primary shield wall.
eactor cavity neutron shield reduces the operating neutron and gamma dose rates in the tainment. A dose reduction factor of 40 on the operating floor of the containment was the gn goal for the shield. With the postulation of leak-before-break applied to the reactor coolant ng, the neutron shield design does not consider dynamic effects resulting from a design basis CA.
permanent reactor cavity neutron shield consists of borated concrete blocks contained inside nless steel containers. The concrete blocks are supported by the shielding support bars, which n between the reactor vessel flange and the embedment ring. The borated concrete blocks are nches high and 23.25 inches wide, located all around the reactor vessel beneath the permanent 11.2-3 Rev. 35
tor vessel (Figures 11.2-11 through 11.2-12). Thermal analysis has shown that the borated crete blocks inside the stainless steel containers could reach a temperature of up to 455°F in e sections of the concrete blocks during normal operation.
shielding is provided with personnel access openings, located directly above the neutron ctor wells. During reactor operation these openings are plugged with borated concrete shield gs to further minimize neutron flux. The shield plugs are removed through the access openings he permanent cavity seal for maintenance of the neutron detector instrumentation.
shielding in combination with the reactor cavity seal described in Section 4.3.10 will provide quate airflow channels to maintain sufficient cavity cooling during reactor operation.
addition to the borated concrete shielding, borated polyethylene layers are installed roximately at the reactor vessel flange elevation to reduce streaming between the borated crete shielding and the primary shield wall. The thickness of the polyethylene shielding ranges m 1 to 3 inches contained in stainless steel liner between the ribs of the support members. The ated polyethylene layers are not insulated. This shielding is cooled by cavity cooling air sing under the shielding at temperatures less than 150°F. Together, the borated concrete blocks the borated polyethylene layers are capable of reducing the neutron radiation by a dose factor pproximately 16 to 27.
ondary shielding is provided to reduce the activity from the reactor coolant system to radiation ls which allow limited access to the containment during normal operation and to supplement ary shielding. Nitrogen 16 is the major source of radioactivity in the reactor coolant during ration and controls the thickness of the secondary shield. The secondary shielding consists of a imum of 3.5 feet of reinforced concrete surrounding the reactor coolant piping, pumps, steam erators and pressurizer. The N-16 activity concentration at the reactor vessel outlet nozzles is x 10-5 Ci/cc at full power.
containment is a reinforced prestressed concrete structure with 3.75 feet thick cylindrical ls and a 3.25 feet thickness dome. In conjunction with the primary and secondary shield, it will t the radiation level outside the enclosure building due to sources inside the structure to no e than 0.5 mrem/hr at full power operation. The structure is also designed to protect plant onnel from radiation sources inside the structure following a postulated incident.
.3.2 Auxiliary Building Shielding function of the auxiliary building shielding is to protect personnel working near various em components, such as those in the CVCS, the radioactive waste processing system, pling system, and the spent fuel pool cooling system. Controlled access to the auxiliary ding is allowed during reactor operation. Each equipment compartment is individually lded to reduce the radiation level in it and adjacent compartments as reflected by the zone gnations. Source terms used in the design of shielding for major components throughout the 11.2-4 Rev. 35
.3.3 Control Room Shielding layout of the control room is shown in Figure 1.2-7. In conformance with General Criterion mber 19 of 10 CFR Part 50, the control room shielding is designed to ensure that the dose will exceed 5 rem for the duration of the incident (see Subsection 14.18.3.3). The walls of the trol room are two feet thick. Under normal operating conditions, the control room is a Zone A on with expected dose rates well below the indicated limits.
.3.4 Spent Fuel Pool Shielding and Fuel Handling Shielding l handling shielding is designed to facilitate the removal and transfer of spent fuel assemblies m the reactor vessel to the spent fuel pool. it is designed to protect personnel against the ation emitted from the spent fuel and control rod assemblies.
refueling cavity above the reactor vessel is flooded to Elevation 36 feet 6 inches to provide a porary water shield above the components being withdrawn from the reactor vessel.
water height is approximately 24 feet above the reactor vessel flange. This height assures a imum of 108 inches of water above the active portion of a withdrawn fuel assembly at its hest point of travel. Under these conditions, the dose rate from the spent fuel assembly is less 1.0 mrem/hr at the water surface.
n removal of the fuel assembly from the reactor vessel, it is moved to the spent fuel pool by fuel transfer mechanism, via the fuel transfer tube. Concrete shielding is provided around tor internals storage and the steam generator for personnel protection during refueling. The nt fuel pool in the auxiliary building is permanently flooded to provide a minimum of 108 es of water above the active portion of a fuel assembly when being withdrawn from the fuel sfer tube and raised by the fuel pool platform crane, prior to insertion in the spent fuel storage
. The minimum water height above stored fuel assemblies is approximately 24.5 feet during ration of the spent fuel pool cooling system, described in Section 9.5.2.1, to avoid air ainment at the pump suction intakes. Otherwise, a minimum height of 23 feet prevails in the nt fuel pool and also, while fuel is in movement, above the reactor pressure vessel, to satisfy ion product retention assumptions for fuel handling accident calculations (see tion 14.7.4.2). The sides of the spent fuel pool are six feet thick concrete to ensure a dose rate ess than 0.035 mrem/hr on the outer surface of the spent fuel pool.
.3.5 Piping Systems Shielding piping systems containing radioactive material are routed and/or provided with shielding in ordance with the radiation zones given in Section 11.2.1. Consideration is given to ntenance and inspection requirements for components located in shielded compartments ugh which these lines are routed.
11.2-5 Rev. 35
.4 HEALTH PHYSICS PROGRAM rmation regarding the health physics program organization is presented in Section 12.5 of lstone 3 Final Safety Analysis Report (Reference 11.2-1). That information is contained herein eference.
.5 REFERENCES
-1 Millstone Unit 3, Final Safety Analysis Report, Section 12.5 - Health Physics Program 11.2-6 Rev. 35
Purification Letdown Ion Letdown Volume Pre-Filter Exchanger Post-filter Control Tank Degasifier sotope (Curies) (Curies) (Curies) Demineralizer (Curies)
Br-84 - 1.23 - 840.6 4.24x10-3 Kr-85m - - - 34.0 0.0 Kr-85 - - - 87.1 0.0 Kr-87 - - - 18.4 0.0 Kr-88 - - - 58.2 0.0 Rb-88 - 38.20 - 4.63 0.23 Rb-89 - 0.82 - 0.11 4.02x10-3 Sr-89 - 339.80 - 0.01 5.24x10-4 Sr-90 - 85.90 - 5.0x10-4 2.63x10-5 Y-90 - 1.79 - 2.1x10-2 1.05x10-3 Sr-91 - - - 6.7x10-3 3.32x10-4 Y-91 - - - 5.22 0.26 Mo-99 - - - 40.86 2.06 Ru-103 - 222.0 - 8.4x10-3 4.24x10-4 Ru-106 - 59.4 - 4.9x10-4 2.49x105 Te-129 - 1.54 - 5.0x10-2 2.52x10-3 I-129 - 2.4x10-2 - 1.44x107 7.29x10-9 I-131 - 4.26x104 - 8.01 4.04x10-1
-131m - - - 50.8 0.0 Te-132 - 1.40x103 - 0.65 3.27x10-2 I-132 - 129.0 - 2.03 1.03x10-1 I-133 - 6.32x102 - 10.90 5.51x10-1 Xe-133 - - - 5012.3 0.0 Te-134 - 0.92 - 4.7x102 2.39x10-3 11.2-7 Rev. 35
Purification Letdown Ion Letdown Volume Pre-Filter Exchanger Post-filter Control Tank Degasifier sotope (Curies) (Curies) (Curies) Demineralizer (Curies)
I-134 - 27.55 - 1.13 5.72x10-2 Cs-134 - - - 8.19 4.13X10-1 I-135 - 933.4 - 5.04 2.54x10-1 Xe-135 - - - 172.4 0.0 Cs-136 - - - 0.28 1.45x10-2 Cs-137 - - - 39.78 2.0 Xe-138 - - - 8.16 0.0 Cs-138 - - - 1.24 3.27x10-2 Ba-140 - 104.2 - 1.23x10-2 6.23x10-4 La-140 - 13.1 - 1.19x10-2 6.00x10-4 Pr-143 - - - 1.08x10-2 5.45x10-3 Ce-143 - - - 7.74x10-3 3.91x10-3 Co-60 263.0 - 2.63 1.98x10-4 1.0x10-5 Fe-59 0.45 - 4.5x10-3 3.83x10-6 2.09x10-7 Co-58 690.0 - 6.90 1.42x10-3 7.18x10-6 Mn-56 3.27 - 3.27x10-2 4.14x10-3 2.09x10-5 Mn-54 3.69 - 3.69x10-2 4.95x10-6 2.50x10-7 Cr-51 88.10 - 0.88 4.32x10-4 2.18x10-7 Zr-95 1.54 - 0.15 3.42x10-6 1.73x10-7 11.2-8 Rev. 35
TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGNS Aerated Aerated Waste Gas Waste Gas Waste Drain Aerated Waste Waste sotope Surge Tank Decay Tank Tank Demineralizer Monitor Tank Br-84 - - 8.8x10-3 2.78x10-5 8.8x10-4 Kr-85m 4.09x102 2.80x102 - - -
Kr-85 3.13.102 2.61x103 - - -
Kr-87 6.97x101 1.41x101 - - -
Kr-88 4.80x102 2.10x102 - - -
Rb-88 - - 0.486 8.63x10-4 0.243 Rb-89 - - 1.2x10-2 1.85x10-5 6x10-3 Sr-89 - - 1.08x10-3 7.68x10-3 5.4x10-4 Sr-90 - - 5.47x10-5 1.94x10-3 2.74x10-5 Y-90 - - 2.20x10-4 4.05x10-5 2.2x10-4 Sr-91 - - 7.04x10-4 - 3.52x10-4 Y-91 - - 5.49x10-2 - 5.49x10-2 Mo-99 - - 0.430 - 0.430 Ru-103 - - 8.84x10-4 5.02x10-3 8.84x10-5 Ru-106 - - 5.19x10-5 1.34x10-3 5.19x10-6 Te-129 - - 5.24x10-3 3.48x10-5 5.24x10-4 I-129 - - 1.52x10-8 5.42x10-7 1.52x10-9 I-131 - - 0.842 9.63x10-1 8.42x10-2
-131m 1.80x103 1.37x104 - - -
Te-132 - - 6.81x10-2 3.16x10-2 6.81x10-3 I-132 - - 0.214 2.92x10-3 2.14x10-2 I-133 - - 1.15 1.43x10-2 1.15x10-1 Xe-133 1.71x105 1.17x106 - - -
11.2-9 Rev. 35
Aerated Aerated Waste Gas Waste Gas Waste Drain Aerated Waste Waste sotope Surge Tank Decay Tank Tank Demineralizer Monitor Tank Te-134 - - 4.98x10-3 2.08x10-5 4.98x10-4 I-134 - - 0.119 6.23x10-4 1.19x10-2 Cs-134 - - 0.861 - 0.430 I-135 - - 0.530 2.11x10-2 5.3x10-2 Xe-135 3.41x102 4.86x102 - - -
Cs-136 - - 3.03x10-2 - 1.52x10-2 Cs-137 - - 4.18 - 2.09 Xe-138 6.75 2.99x10-1 - - -
Cs-138 - - 1.31x10-1 - 6.6x10-2 Ba-140 - - 1.30x10-3 2.35x10-3 1.30x10-4 La-140 - - 1.25x10-3 2.96x10-4 1.25x10-4 Pr-143 - - 1.14x10-3 - 1.14x10-4 Ce-143 - - 8.14x10-4 - 8.14x10-5 Co-60 - - 2.08x10-4 - 2.08x10-5 Fe-59 - - 4.03x10-6 - 4.03x10-7 Co-58 - - 1.50x10-3 - 1.50x10-4 Mn-56 - - 4.35x10-3 - 4.35x10-4 Mn-54 - - 5.29x10-6 - 5.20x10-7 Cr-51 - - 4.54x10-4 - 4.54x10-3 Zr-95 - - 3.60x10-6 - 3.6x10-7 11.2-10 Rev. 35
TABLE 11.2-1 SOURCE TERMS FOR SHIELDING DESIGNS Coolant Primary Waste Secondary Coolant Waste Primary sotope Demineralizer Receiver Demineralizer Monitor Tanks Drain Tank Br-84 4.87x10-1 0.106 4.87x10-2 5.30x10-3 2.65x10-1
-85m - - - - -
Kr-85 - - - - -
Kr-87 - - - - -
Kr-88 - - - - -
Rb-88 4.87x10-1 5.84 4.87x10-2 2.92x10-2 1.50x10+1 Rb-89 1.04x10-2 0.145 1.04x10-3 7.24x10-3 3.62x10-1 Sr-89 4.34 1.30x10-2 4.34x10-2 6.48x10-4 3.24x10-2 Sr-90 1.10 6.56x10-4 1.10x10-1 3.28x10-5 1.64x10-3 Y-90 4.71x10-2 2.63x10-2 4.71x10-3 1.32x10-3 6.59x10-3 Sr-91 2.28x10-2 8.45x10-3 2.28x10-3 4.22x10-4 2.11x10-2 Y-91 2.50x102 6.59 2.50x10-1 3.29x10-1 1.65 Mo-99 9.47x101 5.16x101 9.47 2.58 1.29x101 Ru-103 2.83 1.06x10-2 2.83x10-1 5.30x10-4 2.65x10-2 Ru-106 7.59x10-1 6.22x10-4 7.59x10-2 3.11x10-4 1.56x10-3 Te-129 1.96x10-2 6.29x10-2 1.96x10-3 3.15x10-3 1.57x10-1 I-129 3.07x10-4 - 3.04x10-5 - -
I-131 5.43x102 1.01x101 5.43x101 5.05x10-1 2.53x101
-131m - - - - -
Te-132 1.77x101 0.818 1.77 4.09x10-2 2.04 I-132 1.64 2.566 1.65x101 1.28x10-1 6.42 I-133 8.06x101 1.378x101 8.06 6.89x10-1 3.45x101 Xe-133 - - - - -
Te-134 1.16x10-2 5.97x10-2 1.16x10-3 2.99x10-3 1.49x10-1 11.2-11 Rev. 35
Coolant Primary Waste Secondary Coolant Waste Primary sotope Demineralizer Receiver Demineralizer Monitor Tanks Drain Tank I-134 3.52x10-10 1.43 3.52x10-2 7.15x10-2 3.58 Cs-134 1.50x104 1.03x101 1.5x103 5.17x10-1 2.58x101 I-135 3.52x10-1 6.36 3.52x10-2 3.18x10-1 1.59x101 Xe-135 - - - - -
Cs-136 3.06x101 3.63x10-1 3.06 1.82x10-2 9.08x10-1 Cs-137 8.38x104 5.0x101 8.38x103 2.50 1.252 Xe-138 - - - - -
Cs-138 2.34x10-1 8.18x10-1 2.34x10-2 4.09x10-2 3.92 Ba-140 1.33 1.56x10-2 1.33x10-1 7.79x10-4 3.89x10-2 La-140 1.68x10-1 1.5x10-2 1.68x10-2 7.51x10-4 3.75x10-2 Pr-143 1.24 1.24x10-1 1.24x10-1 6.81x10-3 3.41x10-2 Ce-143 1.09x101 9.77x10-2 1.09 4.88x10-3 2.44x10-2 Co-60 - - - - 6.25x10-3 Fe-59 - - - - 1.21x10-4 Co-58 - - - - 4.49x10-2 Mn-56 - - - - 1.31x10-1 Mr-54 - - - - 1.56x10-4 Cr-51 - - - - 1.36x10-2 Zr-95 - - - - 1.08x10-4 11.2-12 Rev. 35
FIGURE 11.2-1 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 45 FEET 6 INCHES 11.2-13 Rev. 35
FIGURE 11.2-2 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 29 FEET 6 INCHES 11.2-14 Rev. 35
FIGURE 11.2-3 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - ELEVATION (-) 5 FEET 0 INCHES 11.2-15 Rev. 35
FIGURE 11.2-4 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT BUILDING -
ELEVATION 14 FEET 6 INCHES AND 38 FEET 6 INCHES 11.2-16 Rev. 35
FIGURE 11.2-5 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL PERATION WITH 1.0% FAILED FUEL AUXILIARY BUILDING - ELEVATION 14 FEET 6 INCHES AND 25 FEET 6 INCHES 11.2-17 Rev. 35
FIGURE 11.2-6 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL TURBINE BUILDING - ELEVATION 14 FEET 6 INCHES 11.2-18 Rev. 35
FIGURE 11.2-7 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL PERATION WITH 1.0% FAILED FUEL AUXILIARY BUILDING - ELEVATION 36 FEET 6 INCHES 11.2-19 Rev. 35
FIGURE 11.2-8 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - SECTION A-A 11.2-20 Rev. 35
FIGURE 11.2-9 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY BUILDING - SECTION B-B 11.2-21 Rev. 35
IGURE 11.2-10 P&ID, RADIATION ZONES AND ACCESS CONTROL NORMAL OPERATION WITH 1.0% FAILED FUEL CONTAINMENT AND AUXILIARY 11.2-22 Rev. 35
MPS-2 FSAR FIGURE 11.2-11 NEUTRON SHIELD SEGMENT Hanging Outer Beam Cylinder Access Port Inner Cylinder Rev. 22.6
MPS-2 FSAR FIGURE 11.2-12 NEUTRON SHIELDING - SECTIONAL VIEW FLOW AREA THROUGH 4 ROUND ACCESS PORTS:
SUPPORT PLATE RIB POLY SHIELDING FLOW AREA REACTOR VESSEL CONCRETE SHIELD WALL BORATED CONCRETE WALL LINER REFLECTIVE INSULATION NUKON INSULATION NUKON INSULATION FLOW AREA Rev. 22.6
FIGURE 11.2-13 NOT USED 11.2-25 Rev. 35
FIGURE 11.2-14 NOT USED 11.2-26 Rev. 35
MPS-2 FSAR FIGURE 11.2-15 NEUTRON SHIELD - THERMAL LOADINGS (TYPICAL ANNULAR SECTION)
THERMAL ANALYSIS MODEL: TYPICAL SHIELDING SECTION ADIABATIC BULK AIR TEMP = 140 F RV FLANGE FILM COEFFICIENT AIR GAP BORATED CONCRETE FILM COEFFICIENT FILM COEF.
TYPE 304 SS PLATE NUKON INSULATION ADIABATIC T = 600 F SURFACE TEMPERATURE Rev. 22.6
MPS-2 FSAR FIGURE 11.2-16 NEUTRON SHIELD - THERMAL LOADINGS (PLUG SHIELD SECTION)
THERMAL ANALYSIS MODEL: PLUG SHIELDING SECTION ADIABATIC BULK AIR TEMP = 140 F RV FLANGE 1 FILM COEFFICIENT AIR GAP (TYP)
AIR GAP (TYP) 1 BORATED CONCRETE FILM COEFFICIENT FILM COEF.
TYPE 304 SS PLATE NUKON INSULATION ADIABATIC T = 600 F SURFACE TEMPERATURE Rev. 22.6
A.1 REACTOR COOLANT DESIGN BASIS RADIONUCLIDE ACTIVITIES calculate reactor coolant design-basis radionuclide activities, a calculation of reactor core rce term (discussed in Section 11.A.1.1 below) was performed. Reactor coolant activities were ulated taking into account (a) leakage from the fuel pellets to the primary coolant due to fuel cladding failure, (b) removal by radioactive decay, (c) removal by the CVCS purification em, (d) removal of primary coolant for boron reduction, and (e) discharge of primary coolant he liquid waste processing system. Included in the resultant reactor coolant design-basis onuclide inventory are activated corrosion products (crud) and tritium (discussed in tions 11.A.1.2 and 11.A.1.3, respectively). Table 11.A-1 provides a tabulation of the design-s reactor coolant radionuclide activities. Bases for the calculation of these activities are ented in Table 11.A-10.
A.1.1 Development of Reactor Core Radionuclide Activities conservatism, the model used to calculate reactor core radionuclide activities employs an end-ycle equilibrium fuel cycle condition. Cycle characteristics are consistent with nominal two-r fuel cycles (with an 80% capacity factor included). The ORIGEN-S computer program is d to model the buildup and decay of core fission product radionuclides.
IGEN-S calculates the following three categories of core radionuclide activities:
- 1. Fission Products
- 2. Light Elements
- 3. Actinides light elements are neutron activation products of fuel assembly metallic structural ponents. The ORIGEN-S model used in the calculation of Millstone Unit 2 reactor core onuclide activities is predicated on a standard 14-by-14 fuel pin arrangement. Actinides ulate the buildup, decay, and loss due to fissioning. Only the 239Np actinide data are used.
ORIGEN-S results are calculated in terms of gram-atoms per metric tonne of initial heavy al. These results are converted, as necessary, to activities (in Curies) in the core using the owing equation:
NA ore = --------------- { M u1 ( B1C1 + B1C2 + B1C3 ) + M u2 ( B2C1 + B2C2 ) }
ACF 11.A-1 Rev. 35
ual to 3.7 x 1010 disintegrations/second per Curie), Mu1 is equal to 19.208 metric tonnes of nium in Batch 1, Mu2 is equal to 13.72 metric tonnes of uranium in Batch 2, and where B1C1, 2, B1C3, B2C1, and B2C2 are batch/cycle-specific values (in gram-atoms per metric tonnes ranium) calculated by ORIGEN-S.
reactor core radionuclide activities thus calculated and used as basic source terms are lated in Table 11.A-2. Also presented in Table 11.A-2 are the bases for the calculation of e reactor core radionuclide activities.
A.1.2 Corrosion Products activity concentrations of activated corrosion products (crud) in the reactor coolant have been ulated based on the model in the ANSI/ANS standard1 dealing with the radioactive source for normal operation of light water reactors. The starting point for this calculation was the tor coolant radionuclide activity concentrations specified in the standard for a reference PWR h U-tube steam generators (such as those used at Millstone Unit 2). These activity centrations were altered by adjustment factors that were prepared in accordance with the SI/ANS model to reflect the operating parameter differences between Millstone Unit 2 and the rence PWR, and were then further adjusted to the Technical Specification radionuclide centration limits.
A.1.3 Tritium Production ium may be produced in the coolant or enter the coolant from a number of sources. One source om fissioning of uranium within the fuel, yielding tritium as a tertiary fission product. Since aloy fuel cladding reacts with tritium to form zircaloy hydride, no tritium diffuses through the ding 2, 3. Therefore, the tritium released to the coolant from the fuel originates only from ctive fuel.
ium is also produced by the reaction of neutrons with boron in the control element assemblies As). Data from operating plants using B4C control rods indicates that no tritium is released m the control rods. The tritium may combine with carbon to form hydrocarbons and/or with um to form lithium hydride, thereby preventing diffusion through the inconel cladding.
ther possibility is that the low internal temperature of the B4C control rods (relative to ANSI/ANS-18.1-1984, American National Standard Radioactive Source Term for Normal ration of Light Water Reactors, dated 12/31/1984 James M. Smith, Jr., The Significance of Tritium in Water Reactors, GE, APED, 9/19/67 Joseph W. Ray, et. al., Investigation of Tritium Generation and Release in PM Nuclear Power nts, BMI-1787, 10/31/66 11.A-2 Rev. 35
tritium produced in the CEAs is released to the coolant.
ther source of tritium is the activation of boron, lithium, deuterium, and nitrogen within the tor coolant. Boron in the form of boric acid is used in the coolant for reactivity control.
ium is produced in the coolant as a result of neutron-boron reaction and may also be added as H control agent. Deuterium is a naturally occurring constituent of water. Nitrogen may be ent due to aeration of the coolant during shutdown and due to aerated makeup water.
expected tritium releases from the combined liquid and vapor pathways was assumed to be Ci/year per MWt, based on review of the tritium release rates at a number of PWRs evaluated UREG-0017, Rev. 1. The quantity of tritium released in the liquid pathway is based on the ulated volume of liquid released, excluding secondary system wastes, with a primary coolant centration of 1.0 Ci/ml. It was assumed that the remainder was released as a gas from ding ventilation system exhaust.
A.1.4 Fuel Experience t operation of stainless steel-clad fuel rods in the Connecticut Yankee reactor showed fuel ure rates on the order of 0.01%.
aloy-clad UO2 fuel in the Obrigheim reactor in Germany sustained a fuel failure rate just over
% in its first cycle, but this had fallen in the second cycle to essentially zero (0.001%). The fuel ure rate in the Dresden 1 reactor over a nine-year period had averaged < 0.1%, with the rate e recently being even lower. Fuel in the Mihama reactor in Japan and in the Point Beach tor had exceeded the burn up at which failures in fuel of similar design were observed in na, without exhibiting increases in coolant activity (indicative of fuel defects).
l failure rates in the current generation of reactors can be controlled to very low levels.
espread fuel defects in certain reactors have been observed, the cause having been attributed uel clad contamination. Appropriate corrective actions have been devised to ensure that the urrence of such fuel defects will be greatly minimized in the future. Nevertheless, there is ays the possibility, despite careful testing and manufacture, that other defects will become arent in new fuel designs in the future that, because of statistical considerations or ecognized or uncontrollable environmental differences, could not be foreseen. The design nements continuously introduced in nuclear power reactors and their fuel as a natural outcome a dynamic industry will, on rare occasions, introduce such defects. Existing licensing ulations limit coolant activity to that associated with 1% failed fuel, even during these sitory and infrequent periods.
fuel failure rate attainable under more normal conditions has been demonstrated to be nearer
%. Over the lifetime of an operating reactor, the latter rate is expected to predominate.
11.A-3 Rev. 35
ected reactor coolant radionuclide activities are based on data generated from operating ts, field and laboratory tests, and plant-specific design considerations. These activities are t into the PWR-GALE Code (henceforth referred to as GALE), which is a computerized hematical model for calculating the expected releases of radioactive material in liquid and eous effluents from pressurized water reactors (PWRs). The expected reactor coolant onuclide activities are tabulated in Table 11.A-1, juxtaposed with the design-basis reactor lant radionuclide activities.
A.3 CALCULATION OF LIQUID AND GASEOUS EFFLUENT RELEASES A.3.1 Expected Liquid and Gaseous Radioactive Effluent Releases Millstone Unit 2 expected liquid and gaseous radioactive effluent releases are calculated to rmine compliance with 10 CFR Part 50 Appendix I. This calculation is performed using LE. As previously stated, the calculation is based on data generated from operating reactors, d and laboratory tests, and plant-specific design considerations incorporated to reduce the ntity of radioactive materials that may be released to the environment during normal ration, including anticipated operational occurrences. The calculation performed by GALE is ed on (a) American Nuclear Society (ANS) 18.1 Working Group recommendations for stment factors, (b) the release and transport mechanisms that result in the appearance of oactive material in liquid and gaseous waste streams, (c) plant-specific design features used to uce the quantities of radioactive materials that are ultimately released to the environment, and information received on the operation of nuclear power plants. The principal mechanisms that ct the concentrations of radioactive materials in the primary coolant are the following:
fission product leakage to the primary coolant from defects in the fuel cladding and fission product generation in tramp uranium, corrosion products activated in the core (i.e., crud),
radioactivity removed in the reactor coolant treatment systems, and activity removed because of primary coolant leakage.
descriptions of the Millstone Unit 2 liquid and gaseous waste processing systems, as well as GALE input parameters unique to those systems, are presented in Appendix 11.B. The owing tables provide tabulations of the GALE results:
Table 11.A-3: Expected Annual Airborne Effluent Releases (Curies per Year), by Radionuclide, from Each Release Point Table 11.A-4: Expected Annual Liquid Activity Releases (Curies/Year), by Radionuclide, from Each Waste Stream 11.A-4 Rev. 35
Table 11.A-8: Total Annual Design Basis And Expected Releases Of Radioactive Liquid Waste To The Environment From All Sources Combined, In Curies Per Year Table 11.A-9: Total Annual Design Basis And Expected Releases Of Airborne Radioactive Waste To The Environment From All Release Points Combined, In Curies Per Year A.3.2 Design Basis Liquid and Gaseous Radioactive Effluent Releases Millstone Unit 2 design basis liquid and gaseous radioactive effluent releases are calculated etermine compliance with 10 CFR Part 20. To perform this calculation, the design basis mary coolant activity concentrations are used, in conjunction with their corresponding ected primary coolant activity concentrations, to determine isotopic scaling factors. These ing factors are then applied to the expected liquid and gaseous effluent release inventory, as ulated by GALE, in order to determine the design basis liquid and gaseous radioactive uent releases.
calculated design basis liquid and gaseous radioactive effluent releases are provided in the owing tables:
Table 11.A-6: Design Basis Radionuclide Concentrations in Liquid Effluent, in Fractions of 10 CFR Part 20 Concentration Limits Table 11.A-7: Design Basis Radionuclide Airborne Concentrations at the Site Boundary from All Gaseous Effluent Release Points Combined, in Fractions of 10 CFR Part 20 Concentration Limits Table 11.A-8: Total Annual Design Basis And Expected Releases Of Radioactive Liquid Waste To The Environment From All Sources Combined, In Curies Per Year Table 11.A-9: Total Annual Design Basis And Expected Releases Of Airborne Radioactive Waste To The Environment From All Release Points Combined, In Curies Per Year A.4 SOLID WASTE PROCESSING SYSTEM information in this section is based on analyses done for the original licensing of Millstone t 2 and is not derived from the updated radionuclide activities shown in the Appendices to R Chapter 11.
11.A-5 Rev. 35
A.4.1.1 Spent Resins from CVCS Ion Exchanger radioactivity buildup on the CVCS ion exchangers is based on a continuous letdown flow of gpm. One of the two purification ion exchangers and the deborating ion exchanger will be ced to the solid waste processing system at the end of each core cycle. Each of the two fication ion exchangers operates for two cycles, one cycle as a Li removal exchanger and the t cycle as the continuous purification exchanger, such that one of the two exchangers will be he Li removal cycle while the other is for continuous purification.
decontamination factor for ion exchangers is assumed to be 10 for all soluble isotopes except Mo, and Cs. For Y, Mo, and Cs, a removal factor of 10 with a 20% usage factor is used for the emoval cycle for each purification ion exchanger.
buildup of activity on the CVCS ion exchangers resins (32 ft3 each) is given in Table 11.A-3.
A.4.1.2 Spent Resins from Clean Liquid Waste Processing System Demineralizers ivity buildup on the clean liquid waste processing system demineralizers is based on the cessing of 14 system volumes per year of reactor coolant wastes. An average processing rate ed on annual volumes of liquid wastes is assumed, with a decontamination factor of 103 for ble isotopes. Operation is assumed to be divided equally between the two primary ineralizers. The loading on the secondary demineralizer resin bed(s) will be much less than on the primary demineralizers and has not been tabulated.
activity buildup on the clean liquid waste processing system demineralizers resins (42 ft3 h) is given in Table 11.1-3.
A.4.1.3 Spent Resins from Aerated Liquid Waste Processing System Demineralizer ivity buildup on the aerated liquid waste demineralizer resin is based on the processing of the umes of aerated liquid radwaste generated annually. An average processing rate based on ual volumes of liquid waste is assumed, with a total DF of 500 taken for soluble isotopes.
resin activity buildup on the hard piped demineralizer resin (42 ft3) is given in Table 11.1-3.
resin activity buildup on the portable demineralizers (15 ft3 each) is not addressed, since this vity is bounded by the resin activity buildup on hard-piped demineralizer.
A.4.1.4 Spent Resins from Spent Fuel Pool Demineralizer activity buildup on resins from the spent fuel demineralizer is based on the processing of nt fuel pool water and refueling water with a radioactivity level corresponding to one-tenth of tor coolant. The assumption is made that complete mixing of reactor coolant, spent fuel pool 11.A-6 Rev. 35
activity buildup on the demineralizer resin (42 ft3) is given in Table 11.1-3.
A.4.1.5 Contaminated Filter Cartridges ldup of activity on filter cartridges is based on processing of liquid radioactive waste. A ontamination factor (DF) of 10 is taken for each filter. This DF is consistent with operating erience from nuclear power stations utilizing three micron filter cartridges.
activity buildup on the filter cartridges is given in Table 11.1-3.
11.A-7 Rev. 35
ABLE 11.A-1 DESIGN-BASIS AND EXPECTED PRIMARY COOLANT ACTIVITY CONCENTRATIONS Noble Gases DESIGN-BASIS PRIMARY EXPECTED PRIMARY NUCLIDE COOLANT (Ci/gm) COOLANT (Ci/gm)
Kr-85m 1.33E+00 1.40E-01 Kr-85 4.50E+00 2.12E-02 Kr-87 7.64E-01 1.41E-01 Kr-88 2.36E+00 2.53E-01 Xe-131m 2.32E+00 1.17E-01 Xe-133m 3.39E+00 3.03E-02 Xe-133 2.17E+02 6.88E-01 Xe-135m 1.25E-01 1.25E-01 Xe-35 6.13E+00 6.72E-01 Xe-137 1.38E-01 3.28E-02 Xe-138 4.81E-01 1.15E-01 Halogens DESIGN-BASIS PRIMARY EXPECTED PRIMARY NUCLIDE COOLANT (Ci/gm) COOLANT (Ci/gm)
I-131 5.57E+00 4.51E-02 I-132 1.18E+00 2.04E-01 I-133 7.29E+00 1.39E-01 I-134 7.77E-01 3.30E-01 I-135 3.65E+00 2.55E-01 Br-84 4.84E-02 1.55E-02 11.A-8 Rev. 35
Activated Corrosion Products (Crud)
DESIGN-BASIS PRIMARY EXPECTED PRIMARY NUCLIDE COOLANT (Ci/gm) COOLANT (Ci/gm)
Na-24 1.40E+00 4.63E-02 Cr-51 9.50E-02 3.09E-03 Mn-54 4.90E-02 1.60E-03 Fe-55 3.70E-02 1.20E-03 Fe-59 9.20E-03 2.99E-04 Co-58 1.40E-01 4.59E-03 Co-60 1.60E-02 5.29E-04 Zn-65 1.60E-02 5.09E-04 W-187 7.60E-02 2.47E-03 Np-239 6.70E-02 2.19E-03 Other Particulate Radionuclides DESIGN-BASIS PRIMARY EXPECTED PRIMARY NUCLIDE COOLANT (Ci/gm) COOLANT (Ci/gm)
Rb-88 5.04E-02 1.84E-01 Sr-89 3.74E-03 1.40E-04 Sr-90 3.83E-04 1.20E-05 Sr-91 1.83E-03 9.42E-04 Y-91m 1.43E-04 4.46E-04 Y-91 4.83E-03 5.19E-06 Y-93 1.53E-03 4.12E-03 Zr-95 1.20E-02 3.89E-04 Nb-95 6.39E-03 2.79E-04 Mo-99 8.11E-01 6.36E-03 Tc-99m 1.77E-03 4.60E-03 Ru-103 5.57E-03 7.49E-03 Ru-106 1.94E-03 8.99E-02 11.A-9 Rev. 35
DESIGN-BASIS PRIMARY EXPECTED PRIMARY NUCLIDE COOLANT (Ci/gm) COOLANT (Ci/gm)
Ag-110m 1.28E-05 1.30E-03 Te-129m 2.20E-02 1.90E-04 Te-129 7.99E-03 2.33E-02 Te-131m 4.75E-02 1.48E-03 Te-131 8.51E-03 7.45E-03 Te-132 4.46E-01 1.69E-03 Cs-134 1.92E+00 7.56E-03 Cs-136 5.10E-01 9.21E-04 Cs-137 1.58E+00 1.00E-02 Ba-140 6.49E-03 1.30E-02 La-140 5.14E-03 2.48E-02 Ce-141 6.16E-03 1.50E-04 Ce-143 4.02E-03 2.77E-03 Ce-144 4.93E-03 3.89E-03 11.A-10 Rev. 35
TABLE 11.A-2 CALCULATED REACTOR CORE ACTIVITIES Nobel Gases Nuclide Reactor Core Activities (Curies) 1 Kr-85m 1.859E+07 Kr-85 8.557E+05 Kr-87 3.736E+07 Kr-88 5.193E+07 Xe-131m 9.452E+05 Xe-133m 4.636E+06 Xe-133 1.482E+08 Xe-135m 3.021E+07 Xe-135 4.288E+07 Xe-137 1.344E+08 Xe-138 1.266E+08 Halogens Nuclide Reactor Core Activities (Curies) 1 Br-84 1.658E+07 I-131 7.120E+07 I-132 1.036E+08 I-133 1.479E+08 I-134 1.644E+08 I-135 1.400E+08 11.A-11 Rev. 35
Cesium and Rubidium Nuclide Reactor Core Activities (Curies) 1 Rb-88 5.320E+07 Cs-134 1.202E+07 Cs-136 3.475E+06 Cs-137 9.888E+06 Other Nuclides Nuclide Reactor Core Activities (Curies) 1 Cr-51 2.413E+06 Mn-54 9.754E+04 Fe-55 4.496E+05 Fe-59 3.382E+04 Co-58 8.739E+05 Co-60 8.888E+05 Zn-65 6.496E-02 Np-239 1.347E+09 Sr-89 7.321E+07 Sr-90 7.406E+06 Sr-91 9.077E+07 Y-91 9.435E+07 Y-91m 5.255E+07 Y-93 7.297E+07 Zr-95 1.250E+08 Nb-95 1.255E+08 Mo-99 1.343E+08 Tc-99m 1.188E+08 Ru-103 1.093E+08 Ru-106 3.752E+07 11.A-12 Rev. 35
Nuclide Reactor Core Activities (Curies) 1 Ag-110m 2.479E+05 Te-129 2.136E+07 Te-129m 4.324E+06 Te-131 6.042E+07 Te-131m 1.373E+07 Te-132 1.025E+08 Ba-140 1.314E+08 La-140 1.359E+08 Ce-141 1.213E+08 Ce-143 1.127E+08 Ce-144 9.545E+07 The reactor core radionuclide activities tabulated above are based on the following: (a) a core power level of 2,700 MWt and (b) a three-region equilibrium cycle core, with an end-of-cycle core average burn-up of 36,142 MWD/MTU, the three regions having operated at a specific power of 31.74 MWt/MTU for 705, 1333 and 1550 EFPD, respectively.
11.A-13 Rev. 35
BLE 11.A-3 EXPECTED ANNUAL EFFLUENT RELEASES (CURIES PER YEAR),
BY RADIONUCLIDE, FROM EACH RELEASE POINT Release Points: Iodines Turbine Building (Ci/yr) Unit 2 Vent (Ci/yr) Millstone Stack (Ci/yr) 31 0.00E+00 1.77E-01 1.00E-03 33 1.80E-04 5.51E-01 1.00E-03 otals 1.80E-04 7.28E-01 2.00E-03 Release Points: Noble Gases Turbine Building (Ci/yr) Unit 2 Vent (Ci/yr) Millstone Stack (Ci/yr)
-85m 0.00E+00 3.00E+00 1.00E+00 r85 0.00E+00 3.80E+01 5.60E+02
-87 0M.00E+00 3.00E+00 1.00E+00
-88 0.00E+00 5.00E+00 4.00E+00
-131m 0.00E+00 2.70E+01 2.30E+01
-133m 0.00E+00 1.00E+00 2.00E+00
-133 0.00E+00 8.10E+01 7.10E+01
-135m 0.00E+00 3.00E+00 1.00E+00
-135 0.00E+00 1.90E+01 1.20E+01
-137 0.00E+00 0.00E+00 0.00E+00
-138 0.00E+00 2.00E+00 1.00E+00 otals 0.00E+00 1.82E+02 6.76E+02 11.A-14 Rev. 35
Release Points: Others (Totals Allocated to Unit 2 Vent Release for Conservatism)
Turbine Building (Ci/yr) Unit 2 Vent (Ci/yr) Millstone Stack (Ci/yr) 3 none 1.10E+02 none 14 none 7.30E+00 none
-41 none 3.40E+01 none otals none 1.51E+02 none 11.A-15 Rev. 35
Release Points: Particulates Turbine Building (Ci/yr) Unit 2 Vent (Ci/yr) Millstone Stack (Ci/yr) 51 none 9.70E-05 1.40E-05
-54 none 5.68E-05 2.10E-06
-57 none 8.20E-06 0.00E+00
-58 none 4.79E-04 8.70E-06
-60 none 1.13E-04 1.40E-05
-59 none 2.75E-05 1.80E-06 89 none 1.59E-04 4.40E-05 90 none 6.29E-05 1.70E-05 95 none 1.00E-05 4.80E-06
-95 none 4.23E-05 3.70E-06
-103 none 1.66E-05 3.20E-06
-106 none 7.50E-07 2.70E-06
-125 none 6.09E-07 0.00E+00
-134 none 4.74E-05 3.30E-05
-136 none 3.25E-05 5.30E-06
-137 none 8.92E-05 7.70E-05
-140 none 4.00E-06 2.30E-05
-141 none 1.33E-05 2.20E-06 otals none 1.26E-03 2.57E-04 11.A-16 Rev. 35
(CURIES/YEAR), BY RADIONUCLIDE, FROM EACH WASTE STREAM BORON RS MISC. WASTES SECONDARY TURB. BLDG. DETERGENT CLIDE (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) rrosion and Activation Products:
-24 2.46E-04 9.26E-03 4.62E-01 5.94E-05 0.00E+00 51 3.39E-04 1.96E-03 3.81E-02 6.23E-06 4.70E-04
-54 2.11E-04 1.04E-03 1.94E-02 3.13E-06 3.80E-04
-55 1.61E-04 7.84E-04 1.47E-02 2.36E-06 7.20E-04
-59 3.55E-05 1.92E-04 3.54E-03 5.76E-07 2.20E-04
-58 5.71E-04 2.97E-03 5.63E-02 9.13E-06 7.90E-04
-60 7.12E-05 3.47E-04 6.59E-03 1.06E-06 1.40E-03
-65 6.71E-05 3.32E-04 6.27E-03 1.01E-06 0.00E+00 187 2.36E-05 7.16E-04 2.61E-02 3.72E-06 0.00E+00
-239 5.48E-05 9.81E-04 2.45E-02 3.85E-06 0.00E+00 sion Products:
84 4.64E-10 1.90E-05 3.21E-02 4.26E-09 0.00E+00
-88 3.25E-10 1.44E-03 2.42E-01 3.36E-11 0.00E+00 89 1.69E-05 8.99E-05 1.68E-03 2.74E-07 8.80E-06 90 1.62E-06 7.85E-06 1.47E-04 2.36E-08 1.30E-06 91 2.71E-06 1.23E-04 8.95E-03 9.88E-07 0.00E+00 91m 1.75E-06 7.86E-05 1.37E-03 6.33E-07 0.00E+00 91 1.45E-06 6.86E-06 6.47E-05 1.35E-08 8.40E-06 93 1.29E-05 5.68E-04 3.82E-02 4.31E-06 0.00E+00 95 4.80E-05 2.51E-04 4.74E-03 7.68E-07 1.10E-04
-95 3.95E-05 1.85E-04 3.32E-03 5.31E-07 1.90E-04
-99 1.90E-04 3.02E-03 7.26E-02 1.15E-05 6.00E-06 99m 1.80E-04 3.02E-03 3.73E-02 8.30E-06 0.00E+00
-103 8.72E-04 4.79E-03 9.12E-02 1.49E-05 2.90E-05
-106 1.19E-02 5.87E-02 1.11E+00 1.78E-04 8.90E-04
-110m 1.71E-04 8.47E-04 1.58E-02 2.55E-06 1.20E-04 29m 2.16E-05 1.21E-04 2.29E-03 3.74E-07 0.00E+00 11.A-17 Rev. 35
BORON RS MISC. WASTES SECONDARY TURB. BLDG. DETERGENT CLIDE (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) 29 1.40E-05 2.47E-04 9.12E-02 6.49E-07 0.00E+00 131m 1.85E-05 4.97E-04 1.60E-02 2.36E-06 0.00E+00 131 3.38E-06 9.50E-05 1.31E-02 4.32E-07 0.00E+00 31 2.91E-03 2.64E-02 4.51E-01 1.43E-04 1.60E-04 132 5.87E-05 8.39E-04 1.91E-02 3.06E-06 0.00E+00 32 8.64E-05 5.45E-03 1.07E+00 6.24E-05 0.00E+00 33 1.04E-03 3.66E-02 1.24E+00 3.42E-04 0.00E+00 34 5.82E-07 1.41E-03 9.69E-01 2.92E-06 0.00E+00
-134 1.14E-01 2.47E-01 9.36E-02 1.54E-05 1.10E-03 35 3.76E-04 2.22E-02 1.91E+00 3.48E-04 0.00E+00
-136 9.16E-03 2.81E-02 1.11E-02 1.84E-06 3.70E-05
-137 1.52E-01 3.28E-01 1.25E-01 2.05E-05 1.60E-03
-140 1.14E-03 7.90E-03 1.51E-01 2.48E-05 9.10E-05
-140 1.48E-03 1.29E-02 2.76E-01 4.42E-05 0.00E+00
-141 1.69E-05 9.53E-05 1.79E-03 2.92E-07 2.30E-05
-143 3.86E-05 9.81E-04 2.96E-02 4.42E-06 0.00E+00
-144 5.15E-04 2.54E-03 4.78E-02 7.70E-06 3.90E-04 Others 1.56E-01 3.73E-01 6.69E-02 2.20E-04 2.31E-04 cept tium TAL 4.54E-01 1.19E+00 8.90E+00 1.56E-03 8.98E-03 11.A-18 Rev. 35
FROM EACH WASTE SYSTEM MISC. TURB. MISC. TURB.
BORON RS, WASTES, SECONDARY, BLDG., DETERGENT, BORON RS, WASTES, SECONDARY BLDG., DETE UNDILUTED UNDILUTED UNDILUTED UNDILUTED UNDILUTED DILUTED DILUTED , DILUTED DILUTED DIL NUCLIDE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (C Corrosion and Activation Products:
Na-24 5.45E-08 6.04E-06 1.57E-06 5.97E-09 0.00E+00 2.64E-13 9.94E-12 4.96E-10 6.38E-14 0.00 Cr-51 7.50E-08 1.28E-06 1.30E-07 6.26E-10 6.30E-07 3.64E-13 2.10E-12 4.09E-11 6.69E-15 5.05 Mn-54 4.67E-08 6.78E-07 6.60E-08 3.15E-10 5.09E-07 2.27E-13 1.12E-12 2.08E-11 3.36E-15 4.08 Fe-55 3.56E-08 5.11E-07 5.00E-08 2.37E-10 9.65E-07 1.73E-13 8.42E-13 1.58E-11 2.53E-15 7.73 Fe-59 7.86E-09 1.25E-07 1.20E-08 5.79E-11 2.95E-07 3.81E-14 2.06E-13 3.80E-12 6.18E-16 2.36 Co-58 1.26E-07 1.94E-06 1.91E-07 9.18E-10 1.06E-06 6.13E-13 3.19E-12 6.04E-11 9.80E-15 8.48 Co-60 1.58E-08 2.26E-07 2.24E-08 1.07E-10 1.88E-06 7.65E-14 3.73E-13 7.07E-12 1.14E-15 1.50 Zn-65 1.49E-08 2.16E-07 2.13E-08 1.02E-10 0.00E+00 7.20E-14 3.56E-13 6.73E-12 1.08E-15 0.00 W-187 5.22E-09 4.67E-07 8.88E-08 3.74E-10 0.00E+00 2.53E-14 7.69E-13 2.80E-11 3.99E-15 0.00 Np-239 1.21E-08 6.40E-07 8.33E-08 3.87E-10 0.00E+00 5.88E-14 1.05E-12 2.63E-11 4.13E-15 0.00 Fission Products:
Br-84 1.03E-13 1.24E-08 1.09E-07 4.28E-13 0.00E+00 4.98E-19 2.04E-14 3.45E-11 4.57E-18 0.00 Rb-88 7.19E-14 9.39E-07 8.23E-07 3.38E-15 0.00E+00 3.49E-19 1.55E-12 2.60E-10 3.61E-20 0.00 Sr-89 3.74E-09 5.86E-08 5.71E-09 2.75E-11 1.18E-08 1.81E-14 9.65E-14 1.80E-12 2.94E-16 9.45 Sr-90 3.59E-10 5.12E-09 5.00E-10 2.37E-12 1.74E-09 1.74E-15 8.43E-15 1.58E-13 2.53E-17 1.40 Sr-91 6.00E-10 8.02E-08 3.04E-08 9.93E-11 0.00E+00 2.91E-15 1.32E-13 9.61E-12 1.06E-15 0.00 Y-91m 3.87E-10 5.13E-08 4.66E-09 6.36E-11 0.00E+00 1.88E-15 8.44E-14 1.47E-12 6.80E-16 0.00 Y-91 3.21E-10 4.47E-09 2.20E-10 1.36E-12 1.13E-08 1.56E-15 7.37E-15 6.95E-14 1.45E-17 9.02 Y-93 2.86E-09 3.70E-07 1.30E-07 4.33E-10 0.00E+00 1.39E-14 6.10E-13 4.10E-11 4.63E-15 0.00 Zr-95 1.06E-08 1.64E-07 1.61E-08 7.72E-11 1.47E-07 5.15E-14 2.70E-13 5.09E-12 8.25E-16 1.18 Nb-95 8.74E-09 1.21E-07 1.13E-08 5.34E-11 2.55E-07 4.24E-14 1.99E-13 3.56E-12 5.70E-16 2.04 19 Rev
MISC. TURB. MISC. TURB.
BORON RS, WASTES, SECONDARY, BLDG., DETERGENT, BORON RS, WASTES, SECONDARY BLDG., DETE UNDILUTED UNDILUTED UNDILUTED UNDILUTED UNDILUTED DILUTED DILUTED , DILUTED DILUTED DIL NUCLIDE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (C Mo-99 4.21E-08 1.97E-06 2.47E-07 1.16E-09 8.04E-09 2.04E-13 3.24E-12 7.79E-11 1.23E-14 6.44 Tc-99m 3.98E-08 1.97E-06 1.27E-07 8.34E-10 0.00E+00 1.93E-13 3.24E-12 4.00E-11 8.91E-15 0.00 Ru-103 1.93E-07 3.12E-06 3.10E-07 1.50E-09 3.89E-08 9.36E-13 5.14E-12 9.79E-11 1.60E-14 3.11 Ru-106 2.63E-06 3.83E-05 3.77E-06 1.79E-08 1.19E-06 1.28E-11 6.30E-11 1.19E-09 1.91E-13 9.56 Ag-110m 3.79E-08 5.52E-07 5.37E-08 2.56E-10 1.61E-07 1.84E-13 9.09E-13 1.70E-11 2.74E-15 1.29 Te-129m 4.78E-09 7.89E-08 7.79E-09 3.76E-11 0.00E+00 2.32E-14 1.30E-13 2.46E-12 4.02E-16 0.00 Te-129 3.10E-09 1.61E-07 3.10E-07 6.52E-11 0.00E+00 1.50E-14 2.65E-13 9.79E-11 6.97E-16 0.00 Te-131m 4.10E-09 3.24E-07 5.44E-08 2.37E-10 0.00E+00 1.99E-14 5.34E-13 1.72E-11 2.53E-15 0.00 Te-131 7.48E-10 6.20E-08 4.45E-08 4.34E-11 0.00E+00 3.63E-15 1.02E-13 1.41E-11 4.64E-16 0.00 I-131 6.44E-07 1.72E-05 1.53E-06 1.44E-08 2.14E-07 3.12E-12 2.83E-11 4.84E-10 1.54E-13 1.72 Te-132 1.30E-08 5.47E-07 6.50E-08 3.08E-10 0.00E+00 6.30E-14 9.01E-13 2.05E-11 3.29E-15 0.00 I-132 1.91E-08 3.55E-06 3.64E-06 6.27E-09 0.00E+00 9.28E-14 5.85E-12 1.15E-09 6.70E-14 0.00 I-133 2.30E-07 2.39E-05 4.22E-06 3.44E-08 0.00E+00 1.12E-12 3.93E-11 1.33E-09 3.67E-13 0.00 I-134 1.29E-10 9.19E-07 3.30E-06 2.94E-10 0.00E+00 6.25E-16 1.51E-12 1.04E-09 3.14E-15 0.00 Cs-134 2.52E-05 1.61E-04 3.18E-07 1.55E-09 1.47E-06 1.22E-10 2.65E-10 1.00E-10 1.65E-14 1.18 I-135 8.32E-08 1.45E-05 6.50E-06 3.50E-08 0.00E+00 4.04E-13 2.38E-11 2.05E-09 3.74E-13 0.00 Cs-136 2.03E-06 1.83E-05 3.77E-08 1.85E-10 4.96E-08 9.84E-12 3.02E-11 1.19E-11 1.98E-15 3.97 Cs-137 3.36E-05 2.14E-04 4.25E-07 2.06E-09 2.14E-06 1.63E-10 3.52E-10 1.34E-10 2.20E-14 1.72 Ba-140 2.52E-07 5.15E-06 5.14E-07 2.49E-09 1.22E-07 1.22E-12 8.48E-12 1.62E-10 2.66E-14 9.77 La-140 3.28E-07 8.41E-06 9.39E-07 4.44E-09 0.00E+00 1.59E-12 1.39E-11 2.96E-10 4.75E-14 0.00 Ce-141 3.74E-09 6.21E-08 6.09E-09 2.94E-11 3.08E-08 1.81E-14 1.02E-13 1.92E-12 3.14E-16 2.47 Ce-143 8.54E-09 6.40E-07 1.01E-07 4.44E-10 0.00E+00 4.14E-14 1.05E-12 3.18E-11 4.75E-15 0.00 Ce-144 1.14E-07 1.66E-06 1.63E-07 7.74E-10 5.23E-07 5.53E-13 2.73E-12 5.13E-11 8.27E-15 4.19 20 Rev
MISC. TURB. MISC. TURB.
BORON RS, WASTES, SECONDARY, BLDG., DETERGENT, BORON RS, WASTES, SECONDARY BLDG., DETE UNDILUTED UNDILUTED UNDILUTED UNDILUTED UNDILUTED DILUTED DILUTED , DILUTED DILUTED DIL NUCLIDE (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (Ci/ml) (C All Others 3.45E-05 2.43E-04 2.28E-07 2.21E-08 3.10E-07 1.68E-10 4.01E-10 7.18E-11 2.36E-13 2.48 Except Tritium TOTAL 1.01E-04 7.73E-04 3.03E-05 1.57E-07 1.20E-05 4.88E-10 1.27E-09 9.56E-09 1.67E-12 9.64 21 Rev
EFFLUENT, IN FRACTIONS OF 10 CFR PART 20 CONCENTRATION LIMITS FRACTION OF 10 CFR PART 20 MAXIMUM NUCLIDE PERMISSIBLE CONCENTRATIONS 1 Na-24 7.79E-05 Cr-51 6.77E-07 Mn-54 6.90E-06 Fe-55 7.03E-07 Fe-59 2.20E-06 Co-58 2.03E-08 Co-60 5.52E-06 Zn-65 2.29E-06 W-187 1.27E-05 Np-239 8.54E-06 Br-84 2.77E-07 Rb-88 1.84E-07 Sr-89 1.73E-05 Sr-90 1.83E-05 Sr-91 2.73E-07 Y-91m 1.72E-10 Y-91 2.76E-06 Y-93 5.18E-07 Zr-95 2.87E-06 Nb-95 9.34E-07 Mo-99 5.27E-05 Tc-99m 2.82E-09 Ru-103 9.79E-07 Ru-106 2.78E-06 Ag-110m 5.99E-09 Te-129m 1.03E-05 Te-129 4.28E-08 11.A-22 Rev. 35
FRACTION OF 10 CFR PART 20 MAXIMUM NUCLIDE PERMISSIBLE CONCENTRATIONS 1 Te-131m 9.77E-06 Te-131 1.99E-07 I-131 2.17E-01 Te-132 1.89E-04 I-132 8.52E-04 I-133 7.32E-02 I-134 1.25E-04 Cs-134 1.39E-02 I-135 7.69E-03 Cs-136 3.24E-04 Cs-137 5.26E-03 Ba-140 2.86E-06 La-140 3.23E-06 Ce-141 9.79E-07 Ce-143 1.21E-06 Ce-144 7.07E-06 H-3 3.54E-04 Sum of MPC Fractions 2 3.19E-01 Based on 10 CFR Part 20, Appendix B, Table II, Column 2, prior to 1994.
In the course of plant operation, individual isotopic fractions of maximum permissible concentration (MPC) might vary from the values tabulated above. The purpose of this table is to demonstrate that the design of the liquid radioactive waste processing system is adequate at design conditions such that the sum of the fractions of MPCs in the pre-1994 version of 10 CFR Part 20 will not exceed the limit of 1.0.
11.A-23 Rev. 35
ABLE 11.A-7 DESIGN-BASIS RADIONUCLIDE AIRBORNE CONCENTRATIONS T THE SITE BOUNDARY FROM ALL GASEOUS EFFLUENT RELEASE POINTS COMBINED, IN FRACTIONS OF 10 CFR PART 20 CONCENTRATION LIMITS Iodines Fraction of 10 CFR Part 20 Maximum Permissible Nuclide Concentration (1)
I-131 4.13E-04 I-133 7.65E-04 Noble Gases Fraction of 10 CFR Part 20 Maximum Permissible Nuclide Concentration (1)
Kr-85m 7.03E-05 Kr-85 7.87E-02 Kr-87 2.01E-04 Kr-88 7.77E-04 Xe-131m 4.60E-04 Xe-133m 2.07E-04 Xe-133 2.93E-02 Xe-135m 1.85E-05 Xe-135 5.23E-04 Xe-137 none Xe-138 1.16E-04 Particulates Fraction of 10 CFR Part 20 Maximum Permissible Nuclide Concentration (1)
Cr-51 7.83E-09 11.A-24 Rev. 35
Fraction of 10 CFR Part 20 Maximum Permissible Nuclide Concentration (1)
Mn-54 3.35E-07 Co-57 2.53E-10 Co-58 1.38E-06 Co-60 2.43E-06 Fe-59 8.26E-08 Sr-89 9.92E-07 Sr-90 2.37E-06 Zr-95 8.57E-08 Nb-95 6.50E-08 Ru-103 9.19E-10 Ru-106 6.99E-11 Sb-125 1.61E-10 Cs-134 9.39E-06 Cs-136 6.49E-07 Cs-137 9.94E-06 Ba-140 2.49E-09 Ce-141 2.28E-08 Ar-41 1.57E-04 C-14 1.35E-05 H-3 1.02E-04 of MPC Fractions (2) 1.12E-01.
ased on 10 CFR Part 20, Appendix B, Table II, Column 1, prior to 1994.
In the course of plant operation, individual isotopic fractions of maximum permissible concentration (MPC) might vary from the values tabulated above. The purpose of this table is to demonstrate that the design of the gaseous radioactive waste processing system is adequate at design conditions such that the sum of the fractions of MPCs in the pre-1994 version of 10 CFR Part 20 will not exceed the limit of 1.0.
11.A-25 Rev. 35
ADIOACTIVE LIQUID WASTE TO THE ENVIRONMENT FROM ALL SOURCES COMBINED, IN CURIES PER YEAR ANNUAL DESIGN BASIS LIQUID ANNUAL EXPECTED LIQUID NUCLIDE EFFLUENT RELEASE (Ci/yr) EFFLUENT RELEASE (Ci/yr) tivated Corrosion Products (Crud):
Na-24 1.45E+01 4.80E-01 Cr-51 1.26E+00 4.10E-02 Mn-54 6.43E-01 2.10E-02 Fe-55 5.24E-01 1.70E-02 Fe-59 1.23E-01 4.00E-03 Co-58 1.89E+00 6.20E-02 Co-60 2.57E-01 8.50E-03 Zn-65 2.14E-01 6.80E-03 W-187 8.31E-01 2.70E-02 Np-239 7.95E-01 2.60E-02 sion Products:
Br-84 1.03E-01 3.30E-02 Rb-88 6.85E-02 2.50E-01 Sr-89 4.81E-02 1.80E-03 Sr-90 5.11E-03 1.60E-04 Sr-91 1.79E-02 9.20E-03 Y-91m 4.82E-04 1.50E-03 Y-91 7.73E-02 8.30E-05 Y-93 1.45E-02 3.90E-02 Zr-95 1.60E-01 5.20E-03 Nb-95 8.70E-02 3.80E-03 Mo-99 9.81E+00 7.70E-02 Tc-99m 1.58E-02 4.10E-02 Ru-103 7.29E-02 9.80E-02 Ru-106 2.59E-02 1.20E+00 Ag-110m 1.67E-04 1.70E-02 11.A-26 Rev. 35
COMBINED, IN CURIES PER YEAR ANNUAL DESIGN BASIS LIQUID ANNUAL EXPECTED LIQUID NUCLIDE EFFLUENT RELEASE (Ci/yr) EFFLUENT RELEASE (Ci/yr)
Te-129m 2.89E-01 2.50E-03 Te-129 3.19E-02 9.30E-02 Te-131m 5.46E-01 1.70E-02 Te-131 1.49E-02 1.30E-02 I-131 6.05E+01 4.90E-01 Te-132 5.28E+00 2.00E-02 I-132 6.35E+00 1.10E+00 I-133 6.82E+01 1.30E+00 I-134 2.33E+00 9.90E-01 Cs-134 1.17E+02 4.60E-01 I-135 2.86E+01 2.00E+00 Cs-136 2.71E+01 4.90E-02 Cs-137 9.79E+01 6.20E-01 Ba-140 7.98E-02 1.60E-01 La-140 6.00E-02 2.90E-01 Ce-141 8.22E-02 2.00E-03 Ce-143 4.50E-02 3.10E-02 Ce-144 6.59E-02 5.20E-02 H-3 9.90E+02 9.90E+02 TAL 4.25E+02 1.10E+01 THOUT ITIUM 11.A-27 Rev. 35
BORNE RADIOACTIVE WASTE TO THE ENVIRONMENT FROM ALL RELEASE POINTS COMBINED, IN CURIES PER YEAR ANNUAL DESIGN BASIS ANNUAL EXPECTED AIRBORNE EFFLUENT RELEASE AIRBORNE EFFLUENT UCLIDE (Ci/yr) RELEASE (Ci/yr)
I-131 2.23E+01 1.80E-01 I-133 2.89E+01 5.50E-01 Kr-85m 3.80E+01 4.00E+00 Kr-85 1.28E+05 6.00E+02 Kr-87 2.17E+01 4.00E+00 Kr-88 8.38E+01 9.00E+00 Xe-131m 9.94E+02 5.00E+01 Xe-133m 3.36E+02 3.00E+00 Xe-133 4.74E+04 1.50E+02 Xe-135m 4.00E+00 4.00E+00 Xe-135 2.83E+02 3.10E+01 Xe-137 0.00E+00 0.00E+00 Xe-138 1.25E+01 3.00E+00 Cr-51 3.38E-03 1.10E-04 Mn-54 1.81E-03 5.90E-05 Co-57 8.20E-06 8.20E-06 Co-58 1.49E-02 4.90E-04 Co-60 3.93E-03 1.30E-04 Fe-59 8.92E-04 2.90E-05 Sr-89 5.35E-03 2.00E-04 Sr-90 2.56E-03 8.00E-05 Zr-95 4.63E-04 1.50E-05 11.A-28 Rev. 35
POINTS COMBINED, IN CURIES PER YEAR ANNUAL DESIGN BASIS ANNUAL EXPECTED AIRBORNE EFFLUENT RELEASE AIRBORNE EFFLUENT UCLIDE (Ci/yr) RELEASE (Ci/yr)
Nb-95 1.05E-03 4.60E-05 Ru-103 1.49E-05 2.00E-05 Ru-106 7.55E-08 3.50E-06 Sb-125 6.10E-07 6.10E-07 Cs-134 2.03E-02 8.00E-05 Cs-136 2.10E-02 3.80E-05 Cs-137 2.69E-02 1.70E-04 Ba-140 1.35E-05 2.70E-05 Ce-141 6.16E-04 1.50E-05 Ar-41 3.40E+01 3.40E+01 C-14 7.30E+00 7.30E+00 H-3 1.10E+02 1.10E+02 11.A-29 Rev. 35
BLE 11.A-10 BASIS FOR REACTOR COOLANT SYSTEM ACTIVITY NUREG-0017 GALE CODE INPUT wer Level, MWt 2754 cent Failed Fuel 1.0 CS Purification Ion Exchanger Decontamination Factors:
Iodine 100 Cs, Rb 2 others 100 ification Flow Rate (CVCS Purification Ion Exchanger), gpm 60 ctive Purification Flow Rate for Lithium and Cesium Removal, gpm 0 ion Product Escape Rate Coefficients, sec-1 Noble Gases 6.5 x 10-8 Halogens, Cs 2.3 x 10-8 Te, Mo 1.4 x 10-9 All others 1.4 x 10-11 d and Bleed Liquid Waste, gal./day 2670 11.A-30 Rev. 35
FIGURE 11.A-1 ESCAPE RATE COEFFICIENTS 11.A-31 Rev. 35
B.1 BASES cussed in this section are the bases used for determining the expected and design-basis annual ases of radionuclides from the radioactive waste processing system to the environment. The cription of radioactive waste processing system design and operation provided in this endix is a representation of how the design and operation have been modeled in the CFR 50 Appendix I, NUREG-0017, GALE Code analysis. GALE, the NRC computer gram, was used to calculate expected releases of radioactive material in liquid, gaseous and orne effluents. The inputs to GALE are summarized in Table 11.B-1. The total expected ual releases are given in Appendix 11.A, Tables 11.A-3 and 11.A-4.
ected releases, with anticipated operational occurrences included, are calculated (using LE) for the purpose of ascertaining that the radioactive waste processing systems and building tilation systems have sufficient capability to ensure that annual releases will be within the ts specified by 10 CFR Part 50, Appendix I. Design basis releases are calculated (as discussed Appendix 11.A) to demonstrate that the radioactive waste processing systems and building tilation effluents are within the limits of 10 CFR Part 20, Sections 105 and 106 and endix B (version prior to January 1, 1994).
ual system operation (discussed in FSAR Section 11.1) may differ from the model input to LE. However, meeting the applicable criteria is ensured through the use of process and uent radiation monitoring and sampling systems used in conjunction with the Radiological uent Monitoring & Offsite Dose Calculation Manual (REMODCM).
B.2 LIQUID WASTE PROCESSING SYSTEM B.2.1 Processing of Clean Liquid Waste expected releases during normal operation from the clean liquid waste processing system mprised of primary coolant waste and shim bleed) are based on processing and discharging of roximately 1,200,000 gallons per year of reactor coolant wastes with quantities as shown in ure 11.B-1.
clean liquid waste processing system is designed to process reactor coolant waste as well as letdown flow from the CVCS. The following three sources of liquid waste contribute to the n liquid waste processing system:
The primary drain tank, containing clean liquid waste from containment The equipment drain sump tank, containing clean liquid waste from the auxiliary building Shim bleed from the primary coolant letdown that has been processed by a deborating or purification ion exchanger 11.B-1 Rev. 35
id waste system for further processing. The remainder of the primary coolant letdown is sed on to the volume control tank (VCT). In the volume control tank, primary coolant liquid is assed and the gases are purged and sent to the gaseous waste processing system with the aining liquid returned to the primary coolant system.
three liquid sources, listed above, are sent to a mixed-bed ion exchanger. Following cessing in the mixed-bed ion exchangers, the liquid is sent to the coolant waste receiver tanks.
m there, the liquid is then sent to a secondary mixed-bed ion exchanger for further processing, r which this liquid is passed into the coolant waste monitor tanks. From these tanks, the liquid leased to the environment via the discharge canal.
s process stream is shown schematically in Figure 11.B-1. Shown in that figure are the tank umes, flow rates, and decontamination factors (DFs) associated with this liquid waste process am that are inputs to GALE.
B.2.2 Processing of Aerated Liquid Waste tributions to the aerated liquid waste processing system include the following:
Waste water from on-site laundry Drainage from hand wash sinks Equipment and area decontamination waste water Spent fuel pool liner drainage Primary coolant sampling system drainage Auxiliary building floor drainage Primary coolant leakage from miscellaneous sources se contributions are directed into aerated waste drain tanks. The aerated liquid waste collected hese drain tanks is then sent to a mixed-bed ion exchanger for processing, after which it is sent n aerated waste monitor tank. The liquid content of this tank is released to the environment via discharge canal.
B.2.3 Processing of Secondary Side Liquid Waste ondary side liquid waste is comprised of the following three contributions:
Steam generator blowdown 11.B-2 Rev. 35
Condensate demineralizer regenerant solutions from condensate polishing facility m generator blowdown is sent to a blowdown flash tank, where one-third of the liquid wdown flashes to steam and can be released as gaseous effluent. Turbine building floor nage flows to the turbine building sump and is released to the environment via the discharge al when radioactivity is detected. With respect to the liquid waste from the condensate shing facility, the condensate demineralizer regenerant solutions are sent to the waste tralization sumps. The liquid in the waste neutralization sumps is released to the environment the discharge canal.
se three contributions to the secondary side liquid waste (identified above) are shown ematically in Figure 11.B-1.
B.3 GASEOUS WASTE PROCESSING SYSTEM borne releases from the gaseous radioactive waste processing system, main condenser air tor, and containment vents are discharged via the Millstone stack. Containment purges, as l as airborne releases from buildings such as the auxiliary and fuel building, are discharged via Unit 2 enclosure building roof vent. Turbine building releases are discharged via the Unit 2 ine building roof vent. Steam generator blowdown is discharged via a separate blowdown t located on the Unit 2 enclosure building roof. For demonstrating compliance with the uirements of 10 CFR Part 20, Sections 105 and 106, and Appendix B (version prior to January 994) and 10 CFR Part 50, Appendix I, maximum dispersion factors derived from annual rage meteorological data were assumed in the 10 CFR Part 50 Appendix I analysis.
sources and pathways of airborne releases are the following:
Auxiliary Building/Fuel Building Ventilation: Airborne releases from these sources are normally processed by HEPA filter and monitored prior to release to the environment via the Unit 2 enclosure building roof vent.
Steam Generator Blowdown Tank Vent: Approximately one-third of the steam generator blowdown flashes to steam and is released to the environment at the Unit 2 enclosure building roof.
Containment Building: Containment purge air is processed by a HEPA filter and monitored before being released to the environment via the Unit 2 enclosure building roof vent. Releases of routine venting of containment are processed by charcoal adsorber and HEPA filter and monitored before being released to the environment via the Millstone stack.
Gaseous Waste Gas Processing System: Gases from liquid waste processing system tanks and gases stripped in the VCT from the primary coolant letdown flow are stored in six gas 11.B-3 Rev. 35
Main Condenser/Air Ejector: Non-condensable gases from steam flow to the main condenser are monitored and released to the environment via the Millstone stack.
Turbine Building: Steam leakage into the turbine building is released to the environment at the Unit 2 turbine building roof vent.
sources and pathways to the gaseous radioactive waste processing system (identified above) shown schematically in Figure 11.B-2. These contributions are included in the inputs to LE and are summarized in Table 11.B-1.
11.B-4 Rev. 35
TABLE 11.B-1 INPUTS TO PWR-GALE CODE DESCRIPTION UNITS VALUE ermal Power Level MWt 2,754 ss of Primary Coolant 103 lbm 461.1 mary Coolant Letdown Rate GPM 60 tdown Cation Demineralizer Flow Rate GPM 0 mber of Steam Generators 2 tal Steam Flow 106 lbm/hr 11.8 ss of Liquid in Each Steam Generator 103 lbm 132.257 tal Blowdown Rate 103 lbm/hr 73.7 owdown Treatment Method 0, 1, or 2 2 ndensate Demineralizer Regeneration Time days 56 ndensate Demineralizer Flow Fraction fraction 0.813 ual to average flow for all condensate demineralizers ÷ al steam flow)
EAN LIQUID WASTE (see notes) im Bleed Rate:
im Bleed Flow Rate GPD 2,670 for Iodine 10,000 for Cs, Rb 200 for Other Nuclides 5,000 llection Time days 16.66 cess and Discharge Time days 0.2865 ction Discharged 1 uipment Drains (Coolant Waste) Inputs:
w Rate GPD 600 A (Fraction of Primary Coolant Activity) 0.145 for Iodine 1,000 for Cs, Rb 20 for Other Nuclides 1,000 llection Time days 16.66 11.B-5 Rev. 35
DESCRIPTION UNITS VALUE cess and Discharge Time days 0.2865 ction Discharged 1 ean Wastes Inputs: None (see notes)
RATED LIQUID WASTE (see note) rty Wastes (Aerated Waste) Inputs:
w Rate GPD 1,110 A (Fraction of Primary Coolant Activity) 0.0427 for Iodine 100 for Cs, Rb 2.0 for Other Nuclides 100 llection Time days 2.61 cess and Discharge Time days 0.0598 ction Discharged 1 SC. other liquid waste inputs (see note) am Generator Blowdown:
owdown Fraction Processed 1 for Iodine 1 for Cs, Rb 1 for Other Nuclides 1 llection Time days 0 cess and Discharge Time days 0 ction Discharged 1 generant Inputs:
generant Flow Rate GPD 850 for Iodine 1 for Cs, Rb 1 for Other Nuclides 1 llection Time days 0 cess and Discharge Time days 0 ction Discharged 1 11.B-6 Rev. 35
DESCRIPTION UNITS VALUE sc. other GALE Code INPUTS:
ntinuous Stripping of Full Letdown Flow 2 ld-up Time For Xenon days 90 ld-up Time for Krypton days 90 l Time for Decay Tanks for Gas Stripping days 90 RBORNE INPUTS:
ste Gas System Particulate (HEPA) Filter Removal 0%
iciency el Handling Building Charcoal Filter Removal Efficiency 0%
el Handling Building HEPA Filter Removal Efficiency 99%
xiliary Building Charcoal Filter Removal Efficiency 0%
xiliary Building HEPA Filter Removal Efficiency 99%
ntainment Volume 106 ft3 1.899 ntainment Atmosphere Clean-up Charcoal Removal 0%
iciency ntainment Atmosphere Clean-up HEPA Removal 0%
iciency ntainment Atmosphere Clean-up Rate 103 CFM 0 ntainment High Volume (Shutdown) Purge Charcoal 0%
moval Efficiency ntainment High Volume (Shutdown) Purge HEPA Filter 99%
moval Efficiency
. of Purges per Year (2 purges internal to GALE code, see 0 es) ntainment Low Volume Purge (Normal) Charcoal 70%
iciency (i.e.: containment venting, see note) ntainment Low Volume Purge (Normal) Filter Efficiency 99%
.: containment venting, see note) ntainment Low Volume Purge (Normal) Rate CFM 10
.: containment venting, see note) ction of Iodine Released from Blowdown Tank Vent 0.05 11.B-7 Rev. 35
DESCRIPTION UNITS VALUE rcent of Iodine Removed From (Main Condenser) Air 0 ctor Release TERGENT WASTES:
tergent Waste PF (Purification Factor), (DF)-1 0.1 TES:
The component and system designations used in this table reflect the terminology used in NUREG-0017, GALE Code, and may differ from that used by Millstone Unit 2.
The GALE Code has internal calculations that impact the input data. The data provided as GALE Code inputs in this table and Figures 11.B-1 and 11.B-2 may differ from data listed in other tables in the FSAR. A specific example of this are the tank volumes listed on Figure 11.B-1 versus Table 11.1-1. The GALE Code applies a 0.8 or 0.4 factor to tank volumes to calculate collection, process, discharge and decay times.
11.B-8 Rev. 35
MPS-2 FSAR FIGURE 11.B-1 NUREG 0017 GALE CODE INPUT DIAGRAM - LIQUID WASTE SYSTEML Primary Drain Tank T17 w PRIMARY COOLANT 1,617 gallons l- equipment drains en MIXEO...aEO 500 GPO illI0.oo1 PCA DEMINERALIZERS MIXEO-BED Cool2lntWasf:e
~ T22A or B Coolant Wast~
pump seals leakof"f 1-----1~~E3":~~~~~~1- .~, Monitor Tanks
... Recei....rTanks c 20 GPO@O.1 peA DFs .,. T14A or B DFs T15AorB
- 5 I & OTHERS --100 (2) tI oe~ gallons ALL'" 10 (2) 35,300 gaJlons system equipment dr-zins a Cs & Rb-2 80GPO@1.0peA
~ Tank .... 2.670 GPD
...J ~iI'------.TO GASEOUS WASTE SYSTEM T75 o 769 gallons; '"
OEBORATING or PUAJFICATION ION-EXCHANGERS VOLUME RETURNED TO
-:P"RJ_M=A:-:R_Y_C_O_O_L.;.ANT,-_L_E_TD_O_W_N
_ _~ T10A, T 1 O.B orT11 CONTROL 0F TANK 113 FLASHES TO STEAM eo GPM PRIMARY COOLANT to Unlt2 Enclosure Building Roof 1-100 GASEOUS EFFLUENT see FIG. 11.B-2 ell & Rb-2 OTHERS = 50 PER NUREG 0017, PARA 2272 2f3 DISCHARGE BLOWDOWN ...
73,700 ~ (195 GPM) @ SECONDARY LIQUID ACTIVITY TANK LIQUID EFFLUENT T45 DECONTAMINATlON on-site laundry 300 GPD @ NUREG 0017 activity hand wash sinks 200 GPO @ NUREG 017 activity Equipment & area decontamination MIXED-8ED W 40 GPD @ NUREG 0017 activity OENINERALlZERS I-1"24 sub.total ~ 540 GPD @ NUREG 0017 activity OFs *
~ I & OTHERS:; 100 Spent Fuel Pool liner drains Cs&Rb-2 C
w 700 GPD ~ 0.001 PCA TO ENVIRONMENT W FOR DETERGENT WASTE DF's - 10 Primary coolant sampling system drains VIA DISCHARGE CANAL 200 GPD @0.05 PCA
~ Aerated Waste Auxilary building floor drains
~ 200 GPO ~ 0.1 PCA Monitor Tank Primary Coolant leakage, miscellaneous sources Aerated Waste T21 10 Gpd@1.67 peA Drain Tanks 5,384 gallons T20A or B S-U-b-.t-ot-a-I-:-1-,-1-10-G-PD-<1i!-.-0-.0-4-Z-7-P-C-A--------i~ (2) 5,384 gallons D1PHOENIX RESIN DURASIL 230 PORTABLE PORTABLE PORTABLE OEMINERALIZER OEMINERALIZER OEMINERALIZER TURBINE BUILDING TURBINE BUILDING FLOOR DRAINS SUMP NUREG 0017 code calculated @7,200 GPD AND PCA CONDESATE POLISHING FACILITY WASTE condensate demineralizer reqenerant solutions NEUTRILIZATION .....
-=====--======-========--------------------i SUMPS 1------------------------------------------------1p~
850 GPO @AII calculated secondary side activity released per NUREG 0017 code TK-10 or 11 NOTES:
PCA - FRACTION OF PRIMARY COOLANT ACTIVITY
_____________ = ALTERNATE PROCESS PATH THE VALUES IN THIS FIGURE ARE THE ASSUMPTIONS USED IN THE 10CFR50 APPENDIX I ANALYSIS ACTUAL VALUES MAY VARY.
Rev. 22
MNPS-2 FSAR Steam Gene rat or Blowd own Tank ven t 24,566 Ibmlh r steam (1/3 Steam Gene rator Blowdown) (rom FIG. 11.8 -1 Aux iliary Building / Fuel Building .. Unit 2, Enclosure Build ing roof vent
~-------------t~~
16 0 Ibm/day Primary Coolant Leaka ge P5A Purg e HLE2 ' .... Unit 2, Enc losure Building roof vent (satisfies RG 1.140) i- ..
,..~
2 per year eft. =99%
---l Charco al Adso rbe r HEPA Containme nt Build ing Vent ing .... I---~ U n i t 1 Stack, 375 ft. above grad e L29A & B L29A & B 3%/ day Nobl e Gas 68 per year- (satisfies RG 1.1 40) (sa tisfie s RG 1.140) 0.008%/day lodines eft. = 70% eft . 99 %
=
Primary Coolant Leak age
. ._---------~---- - ---_. _- -
Ga .eous W aste Proce ssing System Ga s Decay T-~nks-M - HC1~A from VCT , Degas ifie r, various liqu id wa ste tank ve nts (6) 582 ft3 (does not satisfy RG 1.140) Unit 1 Sta ck , 375 ft. above grade 49 ft3/day eft . 0%
=
Main Condenser / Ai r Ejec tor 75 Ibm/hr Prima ry to Seco ndary leak age NOTE : TH E VALUES IN THI S FIGU RE ARE THE ASS UMP TIONS US ED IN TH E 10CFR 50 APP ENDIX I ANALYS IS. A C T U/~ L VAL UES MAY VARY .
FIGURE 11.8-2 NUREG 0017 GALE CODE INPUT DIAGRAM - AIRBORNE APRIL 1999
C.1 DOSES TO HUMANS analysis of annual doses to the maximum individual and to the population residing within an mile radius of Millstone Unit 2 are based on the methodology and equations presented in U. S.
C Regulatory Guide 1.109 Rev. 1, Calculation of Annual Doses to Man from Routine Releases eactor Effluents for The Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I.
automate the Millstone Unit 2 pathways-to-man dose analyses, U.S. NRC Computer Code DTAP II was used for the liquid effluent releases, and U.S. NRC Computer Code GASPAR used for the gaseous effluent releases. The Regulatory Guide 1.109 equations germane to id and gaseous releases are presented in Sections 11.C.2 and 11.C.3, respectively. In the lstone Unit 2 liquid and gaseous pathways-to-man dose analyses, the NRC default values vided in Regulatory Guide 1.109 were used as input to LADTAP II and GASPAR whenever
-specific data were unavailable.
ensuing sections present the equations/methodologies employed in LADTAP II and SPAR to calculate the dose contributions corresponding to the release/uptake pathways sidered in the Millstone Unit 2 pathways-to-man dose analyses.
C.2 METHODS FOR CALCULATING DOSES FROM LIQUID RELEASES DTAP II was used to calculate the radiation exposure to man from ingestion of aquatic foods, reline deposits, swimming, and boating. Doses are calculated for both the maximum vidual and for the population and are summarized for each pathway by age group and organ.
DTAP II also calculates the doses to certain representative biota other than man in the aquatic ironment, such as fish, invertebrates, algae, muskrats, herons, and ducks, using the models ented in WASH-1258.
equations and assumptions used in the LADTAP II analysis are presented in the ensuing sections. Final dose results based on pathway doses calculated by LADTAP II are presented in le 11.C-1.
C.2.1 Generalized Equation for Calculating Radiation Doses to Humans via Liquid Pathways Raipj = (Cjp) . (Uap) . (Daipj) re:
Raipj = the annual dose to organ j of an individual of age group a from nuclide i via pathway p, in mrem/yr Cip = the concentration of nuclide i in the media of pathway p, in pCi/l, pCi/kg, or pCi/m2 Uap = the exposure time or intake rate (usage) associated with pathway p for age group a, in hr/yr, yr-1, or kg/yr (as appropriate) 11.C-1 Rev. 35
mrem/pCi ingested, mrem/hr per pCi/m2 from exposure to deposited activity in sediment or on the ground, or mrem/hr per pCi/liter due to exposure from boating and swimming C.2.2 Doses from Aquatic Foods
- it p Rapj = (1,100)[(Uap)(Mp)/F] . iQil. Bip . Daipj . e re:
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 point of harvest of aquatic food, dimensionless F = the flow rate of the liquid effluent in ft3/sec 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 hours 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, in hours.
1,100 = the factor to convert from (Ci/yr)/(ft3/sec) to pCi/l the other symbols are as previously defined.
C.2.3 Doses from Shoreline Deposits
- it p - it b Rapj = 110,000 x [(Uap)(Mp)(W)/F] x iQiTiDaipj x e x (1- e )
re:
W = the shoreline width factor that describes the geometry of the exposure, dimensionless Ti = the radiological half-life of nuclide i, in days tb = the period of time for which sediment or soil is exposed to the contaminated water, in hours 11.C-2 Rev. 35
proportionality constant used in the sediment radioactivity model other symbols are as previously defined.
C.2.4 Doses from Swimming and Boating doses from swimming and boating were calculated using the methodology described in SH 1258 (Atomic Energy Commission 1973).
equation for calculation of the external dose to skin and the total body dose from swimming ter immersion) or boating (water surface) is:
- it p Rapj = 1,000 x [(Uap)(Mp)/(F)(Kp)] x iQiDaipj x e re:
Kp = 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.
C.3 METHOD FOR CALCULATING DOSES FROM GASEOUS RELEASES PAR implements the air release dose models of the NRC Regulatory Guide 1.109 for noble es (semi-infinite plume only) and for radioiodine and particulate emissions. GASPAR putes both population and individual doses using site data, meteorological data, and onuclide release source terms as inputs. Location-dependent meteorological data for selected viduals are specified as input data. The site data include the population distribution and the ntities of milk, meat, and vegetation produced. The meteorological data include dispersion
, /Q decayed, /Q decayed and depleted, and deposition D/Q. Population doses, individual es, and cost benefit results are calculated.
re are two basic types of calculations, the individual dose calculation and the population dose ulation. Seven pathways by which the nuclides travel to man are considered. These are plume, und, inhalation, vegetation, cows milk, goats milk, and meat. For the individual dose ulations, man is subdivided into the four age groups of infant (0 to 1 year), child (1 to 11 rs), teenager (12 to 18 years), and adult (over 18 years). Each of these calculations takes into ount eight body organs: total body, gastrointestinal (GI) tract, bone, liver, kidney, thyroid, g, and skin.
equations and assumptions used in the GASPAR analysis are presented in the ensuing sections. Final dose results by GASPAR, based on activity releases tabulated in Table 11.A-3, presented in Table 11.C-1.
11.C-3 Rev. 35
C.3.1.1 Annual Air Doses from Noble Gas Releases (Non-Elevated)
Annual Gamma Air Dose Equation:
D (r,) = 3.17x10 Q i X Q (r,) DF i 4
i Annual Beta Air Dose:
D (r,) = 3.17x10 Q i X Q (r,) DF i 4
i re:
D(r,) = the annual gamma air doses at the distance r in the sector at angle from the discharge point in mrad/year D(r,) = the annual beta air doses at the distance r in the sector at angle from the discharge point in mrad/year Qi = the release rate of the radionuclide i, in Ci/year
/Q(r,) = the annual average gaseous dispersion factor at distance r in sector , in sec/
m3 (corrected for radioactive decay)
DFi = the gamma air dose factors for a uniform semi-infinite cloud of radionuclide i, in mrad*m3/pCi*yr DFi = the beta air dose factors for a uniform semi-infinite cloud of radionuclide i, in mrad*m3/pCi*yr 3.17 x 104 = the number of pCi per Ci divided by the number of seconds per year 11.C-4 Rev. 35
T D (r,) = S F i (r,) DFB i i
re:
DT(r,) = the total body dose due to immersion in a semi-infinite cloud at distance r in sector , in mrem/year SF = the attenuation 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 distance r in sector , in pCi/ m3 DFBi = the total body dose factor for a semi-infinite cloud of the radionuclide i, which includes radiation attenuation through a depth of 5 cm into the body, in mrem*m3/
pCi*yr C.3.1.3 Annual Skin Dose from Noble Gas Releases (Non-Elevated)
DS (r,) = 1.11 S F i (r,) DF i + i (r,) DFS 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*m3/pCi*yr 1.11 = the average ratio of tissue to air energy absorption coefficients other parameters are as previously defined.
11.C-5 Rev. 35
ulatory Guide 1.109 provides gamma air dose and total body dose equations that are licable to elevated releases of noble gases from free-standing stacks that are more than 80 ers tall. However, GASPAR does not include these equations, and any attempt to use SPAR to calculate cloud shine doses from elevated releases would lead to non-conservative lts within approximately 1 kilometer from the point of release. What was done in the lstone Unit 2 analysis was to simulate an elevated release by making a separate GASPAR run g the stack release radionuclide inventory as input, but with ground-level release /Qs used in of stack release /Qs. This conservative approach was taken to calculate noble gas gamma e doses (both gamma air doses and total body doses) to maximum individuals.
C.3.2 Doses from Radioiodines and Other Radionuclides, Exclusive of Noble Gases, Released to the Atmosphere C.3.2.1 Annual Organ Dose Due to External Irradiation from Ground Deposition of Radionuclides G
DGj (r,) = 8,760 S F C i (r,) DFG ij i
re:
DGj (r,) = the annual dose to 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 (assumed to be 0.7), dimensionless CGi (r,) = the ground plane concentration of radionuclide i at distance r in sector , in pCi/m2 DFGij = the open field ground plane dose conversion factor for organ j from radionuclide i, in mrem*m2/pCi*hr 8,760 = the number of hours in a year C.3.2.2 Annual Organ Dose from Inhalation of Radionuclides in Air A
D ja (r,) = R a i (r,) DFA ija i
11.C-6 Rev. 35
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 m3/yr i (r,) = the annual average concentration of radionuclide i in air at location (r,), in pCim3 DFAija = the inhalation dose factor for radionuclide i, organ j, and age group a, in mrempCi C.3.2.3 Annual Organ Dose from Ingestion of Atmospherically Released Radionuclides in Food D m m Dja (r,) = DFI ija U Va f g C Vi (r,) + U a C i (r,)
i F F L L
+ U a C i (r,) + U a f 1 C i (r,)
re:
Cvi(r,), Cmi(r,), CLi(r,), CFi(r,) = the concentrations of radionuclide i in produce (non-leafy vegetables, leafy vegetables, fruits, and grains), milk, and meat, 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 DFIija = the ingestion dose factor for radionuclide i, organ j, and age group a in mrem/pCi fg, fl = 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 C.4 COMPARISON OF CALCULATED ANNUAL MAXIMUM INDIVIDUAL DOSES WITH APPENDIX I DESIGN OBJECTIVES omparison of calculated annual maximum individual doses with 10 CFR Part 50 Appendix I gn objectives is provided in Table 11.C-1.
11.C-7 Rev. 35
general expression for calculating the annual population-integrated dose is:
P D j = 0.001 P d D jda f da d a re:
DPj = the annual population-integrated dose to organ j (total body or thyroid), in man-rem or thyroid man-rem Pd = the population associated with sub-region d Djda = the annual dose to organ j (total body or thyroid) of an average individual of age group a in sub-region d, in mrem/yr fda = the fraction of the population in sub-region d that is in age group a 0.001 = the conversion factor from mrem to rem equation above, used in conjunction with the preceding equations with parameters adjusted each age group, is used to calculate the population doses. The population doses due to annual ases of expected liquid and airborne radionuclide concentrations are presented in Table 11.C-C.6 COST-BENEFIT ANALYSIS sented in this section are the cost-benefit analyses and their results, performed in accordance h the requirements set forth in 10 CFR Part 50, Appendix I, Section II.D. In these analyses, ential augments to the liquid and gaseous radioactive waste processing systems are examined cost-effectiveness using the methodology and data provided in U. S. NRC Regulatory Guide 0 Rev. 0 (March 1976), Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled lear Power Reactors. The beneficial savings of each augment is calculated by multiplying the ulated dose reduction by $1,000 per man-rem. The cost of borrowed money is conservatively med to be 9%(1). Provided in Appendix 11.B is information pertinent to the model input to LE representing the design and configuration of the Millstone Unit 2 radioactive waste cessing systems.
This is predicated on Northeast Utilities' June 1998 Cost of Capital Study, in which it was cluded that the weighted cost of borrowed money for NU, CL&P, WMECO, PSNH, and bined system is 9.3%. Based on this conclusion, it is conservative in the cost-benefit analysis ound down to 9%.
11.C-8 Rev. 35
procedure employed for performing the cost-benefit analyses presented in Sections 11.C.6.2 11.C.6.3 below is taken directly from Regulatory Guide 1.110 Rev. 0. This procedure is marized in the following steps:
- 1. Obtain the direct cost of equipment and materials from Table A-1 in Regulatory Guide 1.110.
- 2. Multiply the direct labor cost obtained from Table A-1 in Regulatory Guide 1.110 by the appropriate labor cost correction factor, obtained from Table A-4 in Regulatory Guide 1.110, to obtain the corrected labor cost for the geographical region in which the plant is located (Millstone Unit 2 being in Region I).
- 3. Add the costs obtained from the previous two steps to obtain the total direct cost (TDC).
- 4. Obtain the appropriate indirect cost factor (ICF) from Table A-5 in Regulatory Guide 1.110.
- 5. Determine the total capital cost (TCC) using the following equation:
- 6. Obtain the appropriate capital recovery factor (CRF) from Table A-6 in Regulatory Guide 1.110.
- 7. Determine the annual fixed cost (AFC) using the following equation:
AFC = TCC x CRF
- 8. Obtain the annual operating cost (AOC) and the annual maintenance cost (AMC) from Regulatory Guide 1.110 Tables A-2 and A-3, respectively.
- 9. Determine the total annual cost (TAC) using the following equation:
TAC = AFC + AOC + AMC
- 10. Determine the benefit of each augment by multiplying the dose reduction to be achieved by $1,000 per man-rem.
- 11. The benefit calculated in the previous step minus TAC provides the net benefit of adding the augment to the radwaste system. A positive net benefit means that adding the augment would be cost-beneficial; conversely, a negative net benefit means that it would not be cost-beneficial to add the augment to the radwaste system.
11.C-9 Rev. 35
le 11.C-2 presents the calculated base case annual total body dose (man-rem), annual imum organ dose (man-rem), and annual thyroid dose (man-rem) associated with the ration of the Millstone Unit 2 liquid radioactive waste processing system for the population ected to live within a 50 mile radius of the plant. In the cost-benefit analysis performed for the id radioactive waste processing system, the augment that was chosen was a 50-gpm ineralizer on the blow-down waste stream. This demineralizer was selected because it was ged to be the least expensive augment that could provide any benefit in the reduction of ulation dose. Assuming that this augment is capable of reducing the population doses to zero extremely conservative assumption), the maximum annual benefit to be realized would be roximately $4,880. However, off-setting this benefit is a total annual cost associated with this ment that is estimated to be approximately $43,800.
clusion: It would not be cost-beneficial to add additional liquid radwaste processing ipment to the existing system.
salient inputs to the liquid radioactive waste processing system cost-benefit analysis (taken m the tables in Regulatory Guide 1.110) are summarized below:
Equipment/material cost: $43,000 Labor cost: $29,000 Labor cost correction factor: 1.6 Indirect cost factor (ICF): 1.58, based on the fact that Millstone is a three unit site, with each unit having its own radwaste system Capital recovery factor (CRF): 0.0973 Annual operating cost (AOC): $25,000 Annual maintenance cost (AMC): $5,000 Dose reduction goal: 4.88 man-rem (assuming that the goal is to reduce to zero the largest population dose resulting from liquid effluent released to the environment, that population dose having been calculated to be the maximum organ dose)
C.6.3 Augments to the Gaseous Radioactive Waste Processing System le 11.C-2 presents the calculated base case annual total body dose (man-rem) and annual oid dose (man-rem) associated with the operation of the Unit 2 gaseous radioactive waste cessing system for the population expected to live within a 50 mile radius of the plant. In the
-benefit analysis performed for the gaseous radioactive waste processing system, the augment was chosen was a pair of 30,000 cfm charcoal/HEPA filtration systems on the Unit 2 11.C-10 Rev. 35
ment is capable of reducing the population doses to zero (an extremely conservative mption), the maximum annual benefit to be realized would be approximately $19,600.
wever, off-setting this benefit is a total annual cost associated with this augment that is mated to be approximately $127,500.
clusion: It would not be cost-beneficial to add the two 30,000-cfm charcoal/HEPA filtration ems to the existing gaseous radioactive waste processing system. The salient inputs to the eous radioactive waste processing system cost-benefit analysis (taken from the tables in ulatory Guide 1.110) are summarized below:
Equipment/material cost: $314,000 Labor cost: $102,000 Labor cost correction factor: 1.6 Indirect cost factor (ICF): 1.58, based on the fact that Millstone is a three-unit site, with each unit having its own radwaste system Capital recovery factor (CRF): 0.0973 Annual operating cost (AOC): $18,000 Annual maintenance cost (AMC): $36,000 Dose reduction goal: 19.6 man-rem (assuming that the goal is to reduce to zero the largest population dose resulting from liquid effluent released to the environment, that population dose having been calculated to be the thyroid dose) 11.C-11 Rev. 35
NDIVIDUAL DOSES WITH 10 CFR PART 50 APPENDIX I DESIGN OBJECTIVES 10 CFR PART 50 APPENDIX I DESIGN CALCULATED OBJECTIVE DOSE 1 ANNUAL DOSE rborne Effluent:
Gamma Air Dose 10 mrad 0.2 mrad Beta Air Dose 20 mrad 0.078 mrad Total Body Dose 5 mrem 0.15 mrem Skin Dose 15 mrem 0.25 mrem Maximum Organ Dose (Thyroid) 15 mrem 8.3 mrem uid Effluent:
Total Body Dose 3 mrem 0.06 mrem Maximum Organ Dose (GI-LLI) 10 mrem 0.91 mrem Per reactor.
11.C-12 Rev. 35
TABLE 11.C-2 ANNUAL TOTAL BODY AND THYROID DOSES TO THE PULATION WITHIN 50 MILES OF THE MILLSTONE SITE, IN MAN-REM, FROM EXPECTED LIQUID AND AIRBORNE EFFLUENT RELEASES Annual Total Population Doses from Airborne Effluent Releases:
- Total Body Dose: 0.607 man-rem
- Thyroid Dose: 19.6 man-rem Annual Total Population Doses from Liquid Effluent Releases:
- Total Body Dose: 1.89 man-rem
- Thyroid Dose: 2.86 man-rem
- Maximum Organ Dose (GI-LLI): 4.88 man-rem 11.C-13 Rev. 35
expected annual inhalation doses to plant personnel and estimated air concentrations were puted as outlined below:
ilibrium airborne concentrations:
Equilibrium concentration = (Production/Losses) = fA/
where: f = flow rate of activity source (Ci/sec)
A = activity of source (Ci/cc)
= loss rates = d = e + R e = leakage or exhaust R = recirculation for iodines, if any es:
whole body dose is determined using a semi-infinite cloud model (Safety Guide 4). The oid inhalation dose is found using the inhalation model and dose conversion factors of TID 44 and Safety Guide 4. The other organ doses are found by the fraction of allowable MPC, ch yields the limiting doses.
ole Body Doses g Dose:
CI Lung D I Lung = ----------------------------- x MPD MPC I Lung where: CI = Equilibrium concentration of the Ith isotope MPCILung = Maximum permissible concentration for the Ith isotope for lung based on a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> workweek taken from ICRP II for the airborne case MPDLung = Maximum permissible yearly dose to lung For Cs 138 and Rb 88, DILung is computed as follows:
DLung = 2.81 x 105 x T x E x C where: Y = radiological half life 11.D-1 Rev. 35
C = Concentration in Ci/cc of CS138 or 88Rb.
Y Min. E(MeV/dis) 138 32.2 1.39 88 17.8 1.68 total lung dose DL is found as follows:
L n D = D I Lung I= 1 roid Dose:
DIThyroid = CI x t x BR x DCFI where: DIThyroid = Thyroid dose in REMs CI = Concentration of the Iodine isotope I t = Exposure time in day = 300 days BR = Breathing rate in cc/day = 2 x 107 cc/day DCF = Dose conversion factors for Iodine in REMs/Ci per TID 14844 total thyroid dose is found as follows:
T n D = I = 1 D I Thyroid es and Results:
rce terms are based on reactor operation with 0.1 percent failed fuel.
tainment Building:
- 1. Containment volume - 1.92 x 106 ft3
- 2. Purge containment for a minimum of five hours at 32,000 CFM prior to personnel entering the containment (see FSAR Section 9.9.2).
11.D-2 Rev. 35
atmosphere will be a small fraction of the 40 GPD rate used in the analysis. The 40 GPD leakage rate was taken for Twelfth AEC Air Cleaning Conference-Analysis of Power Reactor Gaseous Waste System.
- 4. Assume 75 days of activity buildup in the containment.
- 5. A decontamination factor of 10, used for iodines and particulates, was taken from Twelfth AEC Air Cleaning Conference-Analysis of Power Reactor Gaseous Waste System.
results indicate the whole body dose rate will be 0.85 mr/hr. After 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the dose rate reases to 0.44 mrem/hr. Concentrations, given in Table 11.D-1, are below the 10 CFR Part 20 wable airborne concentrations.
iliary Building (Excludes nonradioactive areas and fuel handling area):
- 1. Assume 20 GPD leakage of 120°F fluid and 1 GPD leakage of 550°F fluid with reactor coolant concentrations of activity.
- 2. Ventilation air flow is 40,000 CFM for radioactive areas of auxiliary building. This corresponds to approximately five air changes per hour.
- 3. Iodine and particulate partition factors assumed as follows:
Coolant at 120°F 4 Coolant at 550°F 1
- 4. 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of exposure per year.
es based on 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, 52 weeks per year exposure are as follows:
Lung Dose - 0.046 REM Whole Body Dose - 0.468 REM Thyroid Dose - 0.61 REM borne concentrations are given in Table 11.D-2.
iliary Building (Fuel handling area):
- 1. Sources arise due to evaporation from the spent fuel pool surface.
- 2. Sources assumed as one-tenth of reactor coolant concentrations (0.1% failed fuel).
Only tritium is considered since separation factors for iodines and particulates in 11.D-3 Rev. 35
- 3. Evaporation rate from spent fuel pool at 120°F is 0.37 gpm.
- 5. Only whole body dose is considered as the critical organ for tritium.
- 6. Exposure calculated for 2000 hrs/year.
airborne equilibrium concentration for tritium is 1.48 x 10-8Ci/cc. The whole body dose is mr/yr.
bine Building
- 1. Sources arise due to leakage from the secondary systems into the turbine building.
- 2. Primary coolant system operating at 0.1% failed fuel.
- 3. Primary to secondary leakage of 100 gallons per day.
- 4. 5 gpm continuous blowdown per steam generator.
- 5. Processed blowdown liquid is discharged.
- 6. Feedwater cleanup factor is 0.
- 7. Steam leakage from the secondary side is 1700 lb/hr.
- 8. Liquid leakage from the secondary side is 15 gpm.
- 9. Volume of the turbine building 2,882,000 ft3.
- 10. Exhaust flow rate is 200,000 cfm.
es based on 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, 50 weeks per year exposure are as follows:
Lung Dose - 7.8 x 10-3 REM Whole Body Dos - 2.4 x 10-5 REM Thyroid Dose - 0.314 REM borne concentrations are given in Table 11.D-3.
11.D-4 Rev. 35
TABLE 11.D-1 CONTAINMENT BUILDING AIRBORNE CONCENTRATIONS ISOTOPE CONCENTRATION Ci/cc I-131 1.31 x 10-8 I-132 9.75 x 10-10 I-133 7.46 x 10-9 I-134 3.99 x 10-10 I-135 2.95 x 10-9 Xe-131M 6.35 x 10-8 Xe-133M --
Xe-133 4.60 x 10-6 Xe-135M --
Xe-135 8.61 x 10-8 Xe-137 --
Xe-138 1.24 x 10-9 Kr-83M --
Kr-85M 1.52 x 10-8 Kr-85 1.35 x 10-7 Kr-87 6.07 x 10-9 Kr-88 2.43 x 10-8 Kr-89 --
11.D-5 Rev. 35
ISOTOPE CONCENTRATION µCi/cc Kr-85M 8.90 x 10-9 Kr-85 5.45 x 10-9 Kr-87 4.52 x 10-9 Kr-88 1.53 x 10-8 Xe-131M 9.10 x 10-9 Xe-133 1.11 x 10-6 Xe-135 4.56 x 10-8 Xe-138 1.42 x 10-9 I-129 2.96 x 10-18 I-131 1.63 x 10-10 I-132 4.23 x 10-11 I-133 2.31 x 10-10 I-134 2.21 x 10-11 I-135 1.09 x 10-10 Br-84 1.53 x 10-12 Rb-88 7.24 x 10-11 Rb-89 1.72 x 10-12 Sr-89 2.08 x 10-13 Sr-90 1.07 x 10-14 Sr-91 1.44 x 10-13 Te-129 9.24 x 10-13 Te-132 1.35 x 10-11 Te-134 9.08 x 10-13 Be-140 2.50 x 10-13 Ru-103 1.69 x 10-13 11.D-6 Rev. 35
ISOTOPE CONCENTRATION µCi/cc Ru-106 1.02 x 10-15 La-140 2.39 x 10-13 Ce-144 1.69 x 10-13 Pr-143 2.39 x 10-13 Mo-99 8.31 x 10-11 Y-90 4.17 x 10-14 Y-91 4.51 x 10-12 Cs-134 4.10 x 10-12 Cs-136 1.05 x 10-12 Cs-137 1.31 x 10-11 Cs-138 2.27 x 10-11 Cr-51 1.56 x 10-14 Mn-54 1.31 x 10-14 Fe-59 8.73 x 10-15 Co-58 1.91 x 10-12 Co-60 2.13 x 10-13 Zr-95 3.83 x 10-16 11.D-7 Rev. 35
ISOTOPE CONCENTRATION Ci/cc H-3 5.14E-10 Cr-51 1.59E-15 Mn-54 1.21E-15 Fe-59 9.08E-16 Co-58 2.01E-13 Co-60 2.29E-14 Br-84 1.09E-14 Kr-85M 4.61E-12 Kr-85 2.88E-12 Kr-87 2.19E-12 Kr-88 7.80E-12 Rb-88 5.74E-14 Rb-89 1.17E-15 Sr-89 2.17E-14 Sr-90 1.15E-15 Sr-91 3.24E-15 Y-90 2.85E-15 Y-91 4.72E-13 Zr-95 4.04E-17 Mo-99 5.75E-12 Ru-103 1.75E-14 Ru-106 1.09E-16 Te-129 2.96E-15 Te-132 9.90E-13 Te-134 1.80E-15 I-129 1.56E-18 I-131 7.22E-11 I-132 1.29E-12 I-133 4.32E-11 11.D-8 Rev. 35
ISOTOPE CONCENTRATION Ci/cc I-134 2.60E-13 I-135 8.81E-12 Xe-131M 4.83E-12 Xe-133 5.89E-10 Xe-135 2.41E-11 Xe-138 5.88E-13 Cs-134 4.40E-13 Cs-136 1.01E-13 Cs-137 1.41E-12 Cs-138 3.36E-14 Ba-140 2.40E-14 La-140 1.34E-14 Ce-144 1.81E-14 Fr-143 2.24E-14 Where E-14 = 10-14, etc.
11.D-9 Rev. 35
lant control of personnel exposure from airborne radioactivity will be effected by a continuing gram of sampling for airborne activity and administrative controls through radiation work mits.
- 1. Containment atmosphere is continuously monitored by two sample lines inside containment. The sample is pulled through a fixed particulate filter, a charcoal cartridge and a gas chamber before being pumped back to the containment.
Particulate activity is measured by a beta scintillator through a derivative (rate of change) circuit. Any rapid changes in beta activity will be alarmed in the control room. Gaseous iodine is absorbed on the charcoal cartridge. The particulate filter is changed on a routine schedule to prevent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with the particulate filter. The removed filter and cartridge may be counted in accordance with approved station procedures, if needed, to evaluate the airborne radioactivity concentration during the collection period. Gaseous activity is measured with a beta scintillator type detector. Increased activity above pre-set levels will alarm in the control room.
Upon indication of a significant increase in activity in the containment building, the Shift Supervisor will notify radiation protection. Radiation protection will obtain and analyze samples of containment air to determine the isotopes responsible for the increased activity. During power operation, if conditions warrant, radiation protection personnel may enter the containment and take portable air samples in selected areas in order to locate the source of the activity.
Appropriate respiratory equipment will be indicated on the radiation work permits.
During shutdown and maintenance conditions, air samples will be taken with the portable sampling equipment throughout the containment building on a scheduled basis and as required, to determine the presence of any unusual concentrations of airborne activity. Appropriate respiratory protection, if required, will be indicated on the radiation work permits.
While work is being performed on the reactor vessel head and there is a potential for release of airborne activity, continuous radiation protection monitoring of the work area will be in effect. This monitoring will consist of particulate and iodine sampling. Workers will be instructed to leave the area if they suspect any unusual problem.
The spent fuel pool exhaust monitor is a continuous sampling system that takes suction upstream to the HEPA filter. This sample passes through a particulate filter, charcoal cartridge and gaseous 4 pi beta, gamma detection chamber. The particulate filter is changed on a routine schedule to prevent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with 11.E-1 Rev. 35
radioactivity concentration during the collection period. The gaseous channel is monitored and alarmed in the control room. An alarm condition will require notification of radiation protection to determine the source of the increased activity. Work area air sampling will be established consistent with spent fuel pool ventilation line up and work in progress. These samples will supplement the spent fuel pool exhaust monitor. During periods when the spent fuel pool exhaust or the containment atmosphere monitors are not available, air samples are taken with portable sampling equipment in these areas on a scheduled basis per the Millstone Effluent Control Program to determine the presence of any unusual concentrations of airborne activity.
- 2. The area with the highest potential for creating airborne activity is the sampling room. All other radwaste areas contain equipment that is closed and therefore of a low potential for creating airborne activity.
The sampling room (primary sample sink area) has a separate ventilation system via the sample hood. The hood design is such that at least 100 linear feet per minute flow is maintained over the working area in the hood to ensure that there is no contamination or airborne activity spread into the sampling room. Visual indication is available to the personnel frequenting the sampling room to indicate the condition of the exhaust system. A shutdown exhaust system would require health physics evaluation of airborne conditions prior to entry.
- 3. The radwaste airborne radioactivity monitoring system consists of four radiation monitors, each capable of continuously monitoring airborne radioactive particulates. In addition, each monitor contains an iodine sampling assembly consisting of a replaceable charcoal cartridge mounted in series and downstream of the particulate monitor. The particulate filter is changed on a routine schedule to prevent excessive dust loading on the filter paper. The charcoal cartridge is normally changed in conjunction with the particulate filter. The removed filter and cartridge may be counted in accordance with approved station procedures, if needed, to evaluate the airborne radioactivity concentration during the collection period.
Since the waste gas system of the radwaste areas has the potential for the release of gaseous activity, the airborne radiation monitor sampling the ventilation exhaust duct servicing that area also contains a gaseous radioactivity monitor in series with the particulate and iodine monitor. The gaseous monitor will be used to detect and measure significant noble gas releases from the waste gas system.
Sampling for each monitor is accomplished by extracting a representative sample from the radwaste ventilation exhaust system by using an isokinetic nozzle designed in accordance with ANSI N13.1-1969, Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities.
11.E-2 Rev. 35
analyzed to demonstrate the capability of detection and measurement of airborne contaminants in the areas being served by that section of the ventilation exhaust system.
For this analysis, a concentration of 3x10-9 microcuries per cubic centimeter was used. To determine the radioactive concentration at the sampling point, this value was multiplied by the ratio of the flow for the compartment having the lowest airflow to the total flow at the sample point.
Point A (RM-8999 P&ID 25203-26029)
The airflow at this sampling point is 15,495 scfm. The ventilation exhaust system at this point services the storage area, the offices, the heat exchanger pump area, the coolant waste tank areas, and the evaporator rooms.
The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 800 scfm. Assuming the concentration of 3x10-9 microcuries per cubic centimeter exists in this room with the least flow, the radioactive concentration CA at this point is:
-9 800 scfm C A = 3 x 10 Ci cc ---------------------------------
15, 495 scfm
- 10
= 1.548 x 10 Ci cc Manufacturers data for the radiation monitoring equipment states that the monitoring equipment placed in a 1 mrem/hr background of 1 Mev gamma energies produces a count rate of 78 cpm. Concentrations of 1x10-11Ci/cc (limiting isotope Cs-137) produces a count rate 107 cpm. Therefore, the count rate for the concentration at point A is equal to 1656 cpm.
Point B (RM-8998 P&ID 25203-26029)
The airflow at this sampling point is 18,335 scfm. The ventilation exhaust system at this point services the sampling room, the letdown heat exchanger room, the coolant waste receiver area, and the volume control tank area.
The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 160 scfm. Assuming the value of 3x10-9Ci/cc exists in this area, the radioactive concentration CB at this point is:
11.E-3 Rev. 35
160 scfm Ci cc ----------------------------------
-9 C B = 3 x 10 18, 335 scfm
- 11
= 2.61 x 10 Ci cc Assuming the 78 cpm for the background counting rate and the 107 cpm for the concentration 1x10-11Ci/cc (Cs-137), the count rate for point B is 279 cpm.
Point C (RM-8997 P&ID 25203-26029)
The airflow at this sampling point is 10,170 scfm. The ventilation exhaust system at this point services the charging pump areas, the degasifier areas, and the engineered safety features areas. The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 235 scfm.
Assuming the value of 3x10-9 microcuries per cubic centimeter exists in this area, the radioactive concentration CC, at this point is:
235 scfm Ci cc ---------------------------------
-9 C C = 3 x 10 10, 170 scfm
- 11
= 6.93 x 10 Ci cc Assuming the 78 cpm for the background count rate and the 107 cpm for the concentration 1 x 10-11Ci/cc (Cs-137), the count rate for the concentration at point C is 742 cpm.
Point D (RM-8434 A&B P&ID 25203-26029)
The air flow at this sampling point is 17,920 scfm. The ventilation system at this point services the equipment laydown area, the waste gas decay system areas, the aerated waste gas system area, and the RBCCW heat exchanger area. The airflow for the compartment having the least flow exhausting into this section of the ventilation exhaust system is 800 scfm. Assuming the value of 3x10-9Ci/cc exists in this area, the radioactive concentration, CD, at this point D is:
11.E-4 Rev. 35
-9 800 scfm -
C D = 3 x 10 Ci cc --------------------------------
17, 920 scfm
- 10
= 1.34 x 10 Ci cc Assuming the 78 cpm for the background count rate and 107 cpm for the concentration 1 x 10-11Ci/cc (Cs-137), the count rate for the concentration at point D is 1434 cpm.
The calculations and analysis for the four sampling points shown above indicate a count rate ranging from 279 counts per minute to 1656 counts per minute. Since these count rates are greater than three times the background rate, as specified by the manufacturer, the concentration shown in the above calculations can be detected and alarmed.
The above analysis is conservative in that the value used is based on an exposure to the concentrations specified for forty hours in any period of seven consecutive days. However, since the normal occupancy for these areas is significantly less than this 40-hour period, the exposure of personnel to airborne radioactivity will be considerably less than that stated in the above calculations.
In conclusion, the above analysis clearly demonstrates the capability of the airborne radiation monitoring system to detect and measure appropriate concentrations in the radwaste ventilation exhaust system and thus to comply with the requirements for personnel safety stated in 10 CFR 20.103 and 10 CFR 20.203(d).
11.E-5 Rev. 35