ML19283B659

Calculation of Release of Radioactive Matls in Gaseous Effluents from Nuclear Powered Merchant Ships (NMS-GEFF Code)

ML19283B659
Person / Time
Issue date:	07/30/1978
From:	Bangart R, Cardile F, Jay Collins Office of Nuclear Reactor Regulation
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References
NUREG-0466, NUREG-466, NUDOCS 7903060168
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{{#Wiki_filter:. NUREG-0466 CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM NUCLEAR-POWERED MERCHANT SHIPS (NMS-GEFF CODE? F. P. Cardile i R. L. Bangart J. T. Collic.s

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= I 2. = => r P i Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $6.00; Microfiche $3.00 The price of this document for requesters outside of the N>rth American Continent can be obtained from the National Technical Information Service. d LM

NUREG-0466 CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM NUCLEAR-POWERED MERCHANT SHIPS (NMS-GEFF CODE) F. P. Cardile R. L. Bangart J. T. Collins Manuscript Completed: June 1978 Date Published: July 1978 Division of Site Safety and Environmental Analysis Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

TABLE OF CONTENTS Page CHAPTER 1. INTRODUCTION 1-1 CHAPTER 2. NMS-GEFF CODE 2.1 Discussion 2-1 2.2 Nuclear Steam Supply and Auxiliary System Design 2-1 l

2. 3 Physical Description of Ship 2-6 2.4 Gaseous Sources and Ef fluent Pathways 2-7 2.4.1 Discussion 2-7 2.4.2 Sources and Effluent Pathways 2-7

2. 5 Definitions 2-11

2. 6 Instructions for Completing NMS-GEFF Code Input Data Cards 2-14 2.6.1 Parameters Included in the NMS-GEFF Code 2-14 2.6.2 Parameters Reauired for the NMS-GEFF Code 2-16 CHAPTER 3.

PRINCIPAL PARAMETERS USED IN NUCLEAR-POWERED MERCHANT SHIP SOURCE TERM CALCULATIONS AND THEIR BASES 3.1 Introduction 3-1 3.2 Principal Parameters and their Bases 3-1 E 3.2.1 Thermal Power Level 3-1 3.2.2 Plant Capacity Factor 3-2 3.2.3 Radionuclide Concentrations in the Primary and Secondary Coolant 3-2 3.2.4 Leakage Rate to Reactor Containment 3-19 3.2.5 Auxiliary Compartment Leakage 3-20 3.2.6 Engine Compartment (Turbine Area) Leakage 3-20 3.2.7 Main Condenser Evacuation System Partition Factor for Iodine 3-21 3.2.8 Steam Generator Blowdown Tank Vent 3-22 3.2.9 Reactor Containment Purge Frequency 3-22 3.2.10 Reactor Containment Internal Cleanup System 3-23 3.2.11 Decontamination Efficiencies for Charcoal i Adsorbers and HEPA Filters 3-24 3.2.12 Holdup Times for Charcoal Delay Systems 3-26 3.2.13 Radioactive Particulate Releases for Gaseous Effluents 3-27 3.2.14 Tritium Relear.es 3-29 3.2.15 Carbon-14 Release, 3-30 3.2.16 Argon-41 Releases 3-31 3.2.17 Guidelines for Roundin; Off Numerical Values 3-32 i

TABLE OF CONTENTS (continued) Pace CHAPTER 4. INPUT FORitAT, SA!1PLE PROBLEft, AND FORTRAN LISTING 0F THE NiiS-GEFF CODE 4.1 Introduction 4-1 4.2 Input Data Sheet 4-1 4.3 Sample Problem - Input and Output 4-5 4.4 Listing of N!!S-GEFF Code 4-9 4.4.1 Introduction 4-9 4.4.2 FORTRAN Program Listina 4-9 REFERENCES R-1 ii

LIST OF TABLES Table Pace 2-1 Decontamination Factors for Demineralizers 2.16 2-2 Removal Fractions for Charcoa' Adscrbers for Radioiodine Removal 2-25 3-1 Numerical Values - Concentrations in ?rincipal Fluid Streams of the Reference Nuclear-Power ed Merchant Ship with U-Tube Steam Generators 3-3 3-2 Numerical Values - Concentrations in Principal Fluid Streams of the Reference Nuclear-Powered Merchant Ship with Once-Through Steam Generators 3-4 3-3 Parameters Used to Describe the Referen_ Nuclear-Powered Merchant Ship with U-Tube Steam Generators 3-5 3-4 Parameters Used ti Describe the Reference Nuclear-Powered Merchant Ship wita Once-Through Stean Generators 3-6 3-5 Values Used in Determining Adjustnent Factors for N> : lear-Powered Merchant Ships 3-7 3-6 Adjustment Factors for Nuclear-Powered Merchant Ships with U-Tube Steam Generators 3-8 3-7 Adjustment Factors for Nuclear-Powered Merchant Ships with Gnce-Through Steam Generators 3-9 3-8 Comparison of System Parameters for a Sample NMS Design with that of the Nominal Valces of Table 3-3 3-14 3-9 Assigned Decontamination Efficiencies and Removal Fractions for Charcoal Adsorbers for Radioiodine Removal 3-25 3-10 Particulate Release Rate in Gaseous Effluents 3-28 iii

LIST OF FIGURES Figure Page 2-1 HSSS and Auxiliary Systems for a Nuclear-Powered Merchant Ship with U-Tube Steam Generators 2-3 2-2 NSSS and Auxiliary Systems for a Nuclear-Powered Merchant Ship wi th Once-Through Steam Generators 2-4 2-3 Gaseous Effluent Systems for a Nuclear-Powered Merchant Ship 2-8 2-4 Steam Generator Blowdown Treatment liethods 2-19 3-1 Removal Paths for Nuclear-Powered Merchant Ships with U-Tube Stean Generators 3-10 3-2 Removal Paths for Nuclear-Powered Merchant Ships with Once-Through Steam Generators 3-11 iv

CHAPTER 1. INTRODUCTION The Intergovernmental Maritime Consultative Organization (IMC0) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. Ar important aspect of these guidelines is the deter-mination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors. Since proposed design criteria for these merchant ships prohibit the routine release of liquid waste, a liquid source term has not been included here. The NRC staff has assumed that all liquid waste will be solidified and transported by the merchant ships to the home port for disposal in shallow land burial. In preparation of this report, the NRC staff has made extensive use of the experience gained in land-based nuclear power plants that was presented in NUREG-0017, " Calculation of Releases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors (PWR-GALE Code)."(1) Chapter 2 of this report provides instructions for using the NMS-GEFF Code, j It describes the parameters incorporated in the Code, the input data required, and a step-by-step procedure for completing the input data cards. Chapter 3 provides parameters for an assessment of reactor and radwaste 1-1

system performance for normal operation including anticipated operational occurrences. Items such as primary coola't radionuclide concentrations, expected leakage rates, iodine partition factors, and other parameters necessary to make a realistic assessment of the capabilities of the gaseous radwaste treatment systems used in nuclear-powered me, chant ships are delineated along with their bases. Chapter 4 contains a Fortran IV listing of the NMS-GEFF Code, a form for entering the input data, and a sample calculatien. 1-2

CHAPTER 2. NMS-GEFF CODE 2.1 DISCUSSION The NMS-GEFF (Nuclear-Powered Merchant Ship - Gasecus Effluents) Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous effluents from nuclear-powered merchant ships (NMS). The calculations are based on nuclear stean supply system, auxiliary system and radwaste systen desians applicable to the NMS. The calculations are also based on data and parameters applicable to the NMS. In this report the term " reference case NMS" is used, which applies to system and parameters con-sidered to be representative of HMS designs based on infomation obtained in preparation of this report. Parameters for this reference case NMS are referred to as " nominal" parameters. 2.2 NUCLEAR STEAM SUPPLY AND AUXILIARY SYSTEM DESIGNS At the present time, the type of nuclear steam supply system (NSSS), pro-posed for use in nuclear-powered merchant ships, has been a pressurized water reactor (PWR) with either recirculating U-tube steam generators or once-through steam generators. NSSS's using both of these types of steam generators are considered in this rcport. Descriptions of the NSSS and auxiliary systems for a proposed NMS are contained in the Preliminary Safety Analysis Report (PSAR), for the Competitive Nuclear Merchant Ship Program, NRC Project P-426(2) and in the Safety Analysis Study of the Nuclear Pro-pulsion Systen for Large High-Speed Merchant Ships, NRC Project P-433 These documents were used in establishing the NSSS system design and parameters used in this report. 2-1

A brief description of these systems is provided in the following paracraphs and in Fiaures 2-1 and 2-2. In a NHS using a pressurized water reactor, primary coolant water circu-lates through the reactor core where it removes the heat from the fuel el enents. In the steam generators, heat from the pressurized primary ccolant water is transferred to the secondary coolant water to form steam. The stean expands through the turbine and is then condensed and returned to the steam generators. The primary coolant uater flows back to the The principal mechanisms that affect the concentrations of reactor cora. radioactive materials in the primary coolant are: (1) fission product leakage to the coolant from defects in the fuel cladding and fission pro-duct generation in tramp uranium, (2) corrosion products activated in the core, (3) radioactivity removed in the primary coolant treatment system, and (4) activity renoved because of primary coolant leakage. A side stream of the primary coolant system is continuously passed through filters and demineralizers in the primary coolant treatment system. The purpose of this is to provide for purification of the primary coolant, for maintenance of primary coolant inventory, and for primary coolant chemistry control. In land-based pressurized wrter reactor plants, an additional side strean is taken off the primary coolant treatment system. This side stream acts as a bleed operation to maintain boron concentra ion in the primary coolant at the desired l^ vel, and is referred to here as the boron bleed systen. The boron is used in the pr' mary coolant to control the reactivity in the core. Doron concentration in the coolant is high at the beginning ?-2

PRIMARY COOLANT [ SECONDARY COOLANT ( R E. h (EXTR ACTION STE AM) TURBINES REACTOR CORE CIRCULATING U-TU BE MAIN m STEAM (STEAM GENERATOR CONDENSER GENERATORS BLOWDOWN) REACTOR COOLANT y y k PUMPS i A V \\ 7 4 s '? b A D FEEDWATER CONDENSATE MAIN y HEATERS DEMINERAllZERS CONDENSATE PRIM ARY COOLANT PUMP PUR8F C TION TREATMENT SYSTEM TO BORON BLEED m^ SYSTEM (if Used) PURiflCATION FILTERS AND DEMINERALIZERS FIGURE 2-1 NSSS and Auxiliary Systems For a Nuc! car Powered Merchant Ship

k R S E F. S N N I N B A ED R M U N R T O ESS C N NP I EM AM DU T NP O C S R E E T Z AI S L T NA N ER A DE L NN O O O I M C C E Y D R A R D ES ip TR h N y AE S O s C WT E D A mtn e a E t S E sh E H c yr F N Se O 2 yM S I 2r ad J O MT R K Elie H A i r O C N nxe .GMT FI A D Uuw E U C AA I T E )d IdGAo V N OER Ee P R Ls F r RT E O n U BU a HS N a e P f T E NI Sl c G O( S u S RM N N OE a TM B T r S o NE A T OY F S V TS L Y A O S O S T N D R C N O E YE I N Z RM TAI i RT AT AS L CR A ONS MA I TAP I E F E R CLM R R I TN e AOU P RL I T M T EOP UIFE N RC P D A L OO C y N Y RAm M I j R R P O E T R C yRA OC( E r [

of fuel 'ife, but is reduced by the boron bleed system during the fuel life as burnup of the fuel increases and reactivity control requirements de-crease. In reference 2, it was indicated that reactivity in the core of an NMS during the fuel cycle life is controlled through the use of control rod assemblies and burnable poison rods rather than through the use of a boron bleed system. The use of burnable poison rods is considered to be the reference case method for reactivity control foi the NHS in the NMS-GEFF Code. The NMS-GEFF Code also includes provisionr to calculate the effects of the use of a boron bleed system. The' principal mechanisms that affect the concentration of radioactive materials in the secondary coolant are (1) leakage of primary coolant into the secondary coolant in the steam generator (this is the only source of radioactivity in the secondary coolant), (2) radioactivity removed in secondary coolant treatment systems, and (3) radioactivity removed because of secondary coolant leakace. In a NMS using a recirculating U-tube stean generator, the nonvolatile radionuclides leaking from the primary coolant concentrate in the liquid phase in the steam generator. The degree of concentration is controlled by the steam generator blowdown rate and condensate deaineralizer flow rate. In a NMS using a once-through steam generator, since there is no liquid reservoir in a once-through steam generator, the primary coolant leakage boil; to stgam when it enters the secondary side of th? -team generator. Secondary coolant purity is maintained by a condensate demineralizer system 2-5

and there is no steam generator blowdown. The concentration of radioactivity in tF: secondary coolant is controlled by the condensate demineralizer flow i vate. 2.3 PHYSICAL DESCRIPTION OF SHIP This section provides a brief physical description of the reference case M45 so as to clearly delineice the location of the principal pieces of equipment and the source of the gaseous effluents. The following are the principal areas of the nuclear-powered merchant ship considered in our evaluation of gaseous ef fluents: 1. Reactor containment - the principal items located in the reactor containment are: a. reactor vessel, b. reactor coolant pump, c. steam generators, d. pressurizer, and e. main stean lines. 2. Auxiliaries compartment - the principal itens located in the auxiliaries compartment are: a. primary coolant treatment system, b. liquid radwaste system, c. gaseous radwaste system, d. solid radwaste system, e. residual heat removal system, f. chemical laboratory, and g. component cooling water system. 3. Engine Compartment - the principal items located in the enaine compart-ment are: a. high and low pressure turbines, b. main condenser, c. feedwater heaters, d. condensate system, and e. condensate cleanup system. 2-6

j 2.4 GASEOUS SOURCES AND EFFLUENi PATHWAYS 2.4.1 DISCUSSION I i The average quantity of radioactive material released to the environment l from the NMS during normal operation, including anticipated operational occurrences, is called the " source term" since it is the source or initial number used in calculating the environmental impact of radioactive releases. The calculations performed by the NMS-GEFF Code are based on (1) nominal primary and secondary coolant concentrations derived from the American l Nuclear Society Standard N237 corrected to the NSSS parameters of the ^ NMS, (2) the release and transport mechanisms that result in the appearance of radioactive material in gaseous waste streams, and (3) ship specific design features used to reduce the quantities of radioactive material 1 i ultimately released to the environment. Much of the data used in the cal-culations is based on NUREG-0017f1) which is based on data from operating l land-based reactors, field and laboratory tests, and engineering estimates of certain parameters. 3 ?j 2.4.2 SOURCES AND EFFLUENT PATHWAYS The following sources and effluent systems, shown in Figure 2-3, are con-sidered to be pathways for releases of radioactive materials (noble gases, particulates, iodine, carbon and tritium) in gaseous effluents to the environment during normal operation, including anticipated operational j occurrences. i s i 2-7 I N a I I

PRIMARY SYSTEM WASTE GAS PROCESSING SYSTEM y k k OOO WASTE GAS (j j (j STORAGE TANKS u u ,r >- TO ENVIRONMENT AUXILIARIES COMPARTMENT VENTILATION > TO ENVIHONMENT RFACTOR CONTAINMENT [ PURGE > TO ENVIRONMENT ENGINE COMPARTMENT (TURBINE AREA) VENTILATION F TO ENVIRONMENT M AIN CONDENSER EVACUATION SYSTEM > TO ENVIRONMENT STEAM GENERATOR BLOWDOWN TANK VENT (FOR RECIRCULATING U-TUBE STEAM GENERATOR PLANT ONLY) > TO ENVIRONMENT FIGURE 2-3 Gaseous Effluent Systems For a Nuclear Powered Merchant Ship

_.= 2.4.2.1 Primary System Waste Gas Processing System Radioactive gases are removed from the primary coolant by degassification. This process may take place in the venting of a purification tank in the primary coolant treatment system (PCTS) or in a gas stripper, if one is l used, in the PCTS. These gases may be collected in pressurized waste gas storage tanks and held for radioactive decay prior to release to the environment. This calculational model considers inputs to the waste gas processing system from both continuous stripping of the primary coolant during nomal operation and from degassing the primary coolant for two cold shutdowns per year. 2.4.2.2 Reactor Containment Purge Because of leakage through valve stems and pump shaft seals, some coolant escapes into the reactor containment. A portion of the leakaae evaporates, thus contributing to the gaseous source term, and a fraction remains as liquid. The relative amount of leakage entering the gaseous and liquid phases is dependent upon the temperature and pressure at the point where the leakage occurs. Most of the noble gases enter the gas phase, whereas iodine partitions into both phases. The NMS-GEFF Code calculates the release rates of noble gases and iodine to the containment atmosphere based on coolant leakage rates to the containment. Iodine partitioning is considered to be a function of the temperature of the coolant that is released. Particulate release rates are based on measurements at operating land based pressurized water reactors. 2-9 ~

\\ 2.4.2.3 Auxiliaries Compartment Ventilation Because of leakage through valve stems and pump shaft seals, some coolant escapes into the auxiliaries comparunent. A portion of the leakage evapo-rates thus contributing to the gaseous source term, and a fraction remains as liquid. The relative amount of leakage entering the gaseous and liquid phases is dependent upon t% temperature and pressure at the point where the leakage occurs. Most of the noble gases enter the gas phase, whereas iodine partitions into both phases. The NMS-GEFF Code calculates the release rates of noble gases and iodine to the auxiliaries compartment atmosphere based on coolant leakage rates to the auxiliaries compartment. Iodine partitioning is considered to be a function of the temperature of the coolant that is released. Particulate release rates are based on measure-ments at operating land-based pressurized water reactors. 2.4.2.4 Aain Condenser Evacuation System Radioactivity will be present in the secondary coolant system as a result of primary to secondary leakage as discussed in Section 2.2 above. The main condenser evacuation system pulls a vacuum on the main condenser resulting in a potential effluent pathway for radioactivity in the secondary coolant. 2.4.2.5 Engine Compartment (Turbine Area) Ventilation Radioactivity will be present in the secondary coolant system as a result of primary to secondary lpakage as discussed in Section 2.2 above. Because of leakage through valve stem packings and pumps in the secondary coolant system, sone ' secondary coolant escapes into the engine compartment atmosphere. 2-10

r, i-i 2.4.2.6 Steam Generator Blowdown Tank Vent This release path is only applicable to those NMS with retirculating U-tube [ i steam generators, since those with once-through steam generators do not have i steam generator blowdown systems. For plants equipped with steam generator blowdown systems, the NMS-GEFF Code determines the amount of iodine which i has the potential for release through the ir, vent, i 2.5 DEFINITIONS The following definitions apply to terms used in this report: Activation Gases: The gases (including oxygen, nitrogen, and argon) that become radioactive as a result of irradiation in the core. h i Decontamination Factor (DF): The ratio of the initial amount of a nuclide in a stream (specified in terms of concentration of activity of radioactive

L materials) to the final amount of that nuclide in a stream following treat-L ment by a given process.

l P Effective Full Power Days: The number of d ys a plant would have to operate (R at 100% licensed power to produce the integrated thermal power output during a calendar year, i.e., Effective Full Power Days = Integrated Themal Power ii = Licensed Power Level P total where E P is the ith power level, in MWt; 5 P is the licensed power level, in MWt; and l total T is the time of operating at power level P, in days. j j L 2-11 h F d.._D F _a

Fission Product: A nuclide produced either by fission or by subsequent radioactive decay or neutron activation of the nuclides formed in the fission process. Fuel Defect: An imperfection in the jacket of nuclear fuel rods, measured by the comparison of iodine-131 concentration in the primary coolant to the calculated (expected) concentration of iodine-131 ir the primary coolant for a 1% cladding defect level for a specified reactor system (coolant volumes, cleanup rates) minus the calculated iodine-131 concentration from a tramp uranium as determined from the ceasured ratio of iodine-131 to iodine-133 in the primary coolant. .[ Gaseous Effluent Stream: Processed gaseous wastes containing radioactive materials resulting from the operation of a nuclear reactor. j Partition Coefficient (PC): The ratio of the concentration of a nuclide in the gas phase to the concentration of a nuclide in the liquid phase when the liquid and gas are at equilibriun. Plant Capacity Facter: The ratio of the average net power to the rated power capaci ty. Primary Coolant: The fluid circulated through the reactor to remove heat. The normal radionuclide concentrations for the primary coolant are based on I4I the values in the American Nuclear Society Standard N237 , which is based on a review of available data and mathematical models. In order to obtain the nominal primary coolant concentrations for' the reference case NMS, the values in N237 were corrected to the nuclear-powered merchant ship NSSS parameters. 2-12 s

I R P Y Provisions are made in the Nf15-GEFF Code for adjusting the primary coolant concentrations should a particular merchant ship under design be designed to parameters that are different from the nominal values given in Table 3-3 or Table 3-4. The radionuclides are divided into the following categories: 1. Noble gases 2. Iodine 3. Cesiun 4 Other nuclides (as listed in Tables 3-1 and 3-2 of Chapter 3 of this document). Radioactive Noble Gases: The radioactive isotopes of helium, neon, argon, krypton, xenon, and radon, which are characterized by their chemical in-ac ti vi ty. The radioactive isotopes of krypton, xenon, and arcon are the key elements considered in dose calculations. Radioactive Release Rate: The averaoe quantity of radioactive material released to the environnent from a nuclear-powered merchant ship durino nonnal operation, including anticipated operational occurrences. Removal Fraction (RF): The inverse of the decontamination factor, i.e., the ratio of the final amount of a nuclide in a stream to the initial amount of that nuclide in a stream following treatnent by a given process. Secondary Coolant: The coolant converted to steam by the primary coolant in a heat exchanger (steam generator) to power the turbine. The nominal radionuclide concen-trations for the secondary coolant are based on the values in the American I4I Nuclear Society Standard N237 , which is based on a review of available data and mathematical models. In order to obtain the nominal secondary coolant concentrations for the reference case Nf15, the values in N237 were corrected to the nuclear-powered merchant ship NSSS parameters. Provisions 2-13

are made in the NMS-GEFF Code for adjusting the secondary coolant concentra-tions should a particular merchant ship under design be designed to para-meters that are different from the nominal values given in Table 3-3 and Table 3-4. Source Tern: The calculated average quantity of radioactive material released to the environment from a nuclear-powered merchant ship during normal operation, including anticipated operational occurrences. The source tern is the isotopic distribution of radioactive materials used in i evaluating the inpact of radioactive releases on the environment. I Steam Generator Blowdown: Liquid removed from a steam generator in order j to maintain proper water chemistry. l Tramp Uranium: The uranium present on the cladding of a fuel rod. j i 2.6 INSTRUCTIONS FOR COMPLETING NMS-GEFF CODE INPUT DATA CARDS r f 2.6.1 PARAMETERS INCLUDED IN THE NMS-GEFF CODE The carameters listed below are built into the NMS-GEFF Code since it was considered that they would be applicable to all nuclear-powered merchant ship source term calculations and do not require entry on the input data cards. 2.6.1.1 The Plant Capacity Factor 0.9 (330 effective full power days per year). f I 2.6.1.2 Radionuclide Concentrations ir the Primary Coolant, Secondary Coolant, and Main Steam See Section 3.2.3 of Chapter 3 of this document. 2.6.1.3 Leakage to the Reactor Containnent } 1%/ day of primary coolant noble gas inventory and 0.001%/ day of primary coolant iodine inventory. 2-14

2.6.1.4 Primary System Volumes Degassed Per Year Two cociant volumes per year for cold shutdowns plus volumes degassed due to continuous stripping. 2.6.1.5 Auxiliary Compartnent Leakage 160 lb/ day, partition f actor of 0.0075 for radiciodine. 2.6.1.6 Stean Generator Partition Factors (PF) Once-through Steam Generator P F_ Iodine 1.0 Nonvolatiles 1.0 Recirculating Il-Tube Steam Generator Iodine 0.01 Nonvolatiles 0.001 2.6.1.7 Main Condenser / Air Ejector Partition Factor for Radiciodine Partition factor of 0.15 for volatile iodine species and zero for non-volatile species. 2.6.1.8 Reactor Containment Internal Cleanuo System For systems using an internal cleanup system, the NMS-GEFF Code calculates the iodine concentration in the reactor containment atmosphere based on 16 hours of system operation, an iodine decontamination factor of 10 for the charcoal adsorbers, a particulate DF of 100 for HEPA filters, and an internal mixing efficiency of 70%. 2-15

2.6.1.9 Decontamination Factors for Demineralizers Table 2-1 Demineralizer Decontamination Factors Other Demineralizer !odine Cesium Nuclides Condensate System 10 2 10 Primary Coolant Treat-ment System Mixed Bed (Li 80 ) 10 2 10 3 3 Cation 1 10 10 Blowdown Treatment 10 10 10 2.6.1.10 Steam Leakage Rate to the Engine Compartment 1700 lb/hr at main steam activity. Partition factor of 1.0 for iodine. 2.6.2 PARAMETERS REQUIRED TO BE ENTERED ON INPUT DATA CARDS FOR THE NMS-GEFF CODE Complete the cards in the sections below from design information for the particular nuclear-powered merchant ship being evaluated and from the source term parameters specified below and discussed in Chapter 3 of this report. 2.6.2.1 Card 1: Name of Nuclear-Powered Merchant Ship Enter in spaces 33-60 the name of the nuclear-powered merchant ship. 2.6.2.2 Card 2: Thermal Power Leval Enter in spaces 73-80 the maximum thermal power level (in MWt) evaluated for safety considerations for the nuclear-powered merchant ship. 2-16

2.6.2.3 Card 3: Mass of Coolant in Primary System 3 Enter in spaces 73-80 the mass of coolant (in 10 lb) in the primary system i j at operating temperature and pressure. 5 2.6.2.4 Card 4: Primary System Letdown Rate Enter in spaces 73-80 the average letdown rate igal/ min) from the primary system to the primary coolant treatment system. + 2.6.2.5 Card 5: Letdown Cation Demineralizer Flow Rate Enter in spaces 73-80 the annual average flow rate (gal / min) through cation i demineralizers in the primary coolant treatment system for the control of cesium in the primary coolant. The average flow rate is determined by ~ multiplying the average letdown rate (value entered on Card 4) by the fraction of time the cation demineralizers are in service to obtain the i average cation demineralizer flow rate. 4 2.6.2.6 Card 6: Number of Steam Generators 1 Enter in spaces 73-80 the number of steam generators. 1 2.6.2.7 Card 7: Total Steam Flow Enter in spaces 73-80 the total steam flow (in 10 lb/hr) for all steam i generators. 2.6.2.8 Card 8: Mass of Liquid in Each Steam Generator 3 Enter in spaces 73-80 the mass of liquid (in 10 lb) in each steam generator. / 2.6.2.9 Card 9: Steam Generator Blowdown Rate f Enter in spaces 73-80 the steam geneiator blowdown rate in thousands of i lb/hr. For a once-through steam generator, leave spaces 73-80 blank. l 2-17

i 2.6.2.10 Card 10: Steam Generator Treatment Decontamination Factor i Enter in spaces 73-80 the decontamination factor (DF) that the iodine in s the steam cenerator blowdown undergoes in its treatment. Potential blowdown treatment methods are shown in Figure 2-4. This DF is deter-mined in the following way: 1. If the steam generator blowdown is recycled directly to the condensate system demineralizers without prior treatrent in the blowdown systen, i enter a DF=10.0 in spaces 73-80 which corresponds to the DF for iodine through a condensate demineralizer in Table 2-1. 2. If the steam generator blowdown is treated in blowdown system demin-eralizers pri'or to routino to the condensate system (no treatment by [L condensate demineralizers) enter a DF=100.0 in spaces 73-80. This DF l corresponds to the DF for iodine through a blowdown demineralizer in i Table 2-1. 3. If the steam generator blowdown is recycled to the condensate system l demineralizers af ter treatment in the blowdown system demineralizers, { enter a DF=1000.0 in spaces 73-80, which corresponds to the DF for iodine through a blowdown demineralizer and a condensate demineralizer combined in Table 2-1. l 4. If there is no steam generator blowdown treatment or if there is a once-through steam generator, enter a DF=1.0 in spaces 73-80. L i 2.6.2.11 Card 11: Fraction of Feedwater Through The Condensate h Demineralizers i Enter in spaces 73-80 the fraction of feedwater to the steam generator pro- ? cessed through the condensate demineralizers. If condensate demineralizers ( are not used, enter zeros in spaces 73-80. t 2-18 I'

l R CASE 1 STEAM l GENERATOR CONDENSATE MAIN DEMIN. CONDENSER BLOWDOWN CASE 2 STEAM MAIN I GENERATOR CONDENSER BLOWDOWN DEMIN. CASE 3 STEAM l GENERATOR CONDENSATE MAIN DEMIN. CONDENSER BLOWDOWN DEMIN. Fig. 2-4 Potential Steam Generator Blowdown Treatment Methods 2-19

'Wh I 2.6.2.12 Card 12: Boron Bleed Rate Enter in spaces 73-80 the rate (in gpm) of flow of bleed for boron control if this means is used to control reactivity. If boron bleed is not used to control reactivity, leave spaces 73-80 blank. 2.6.2.13 Card 13: Burnable Poison Rods for Boron Control Card 13 identifies whether burnab!e poison rods are used for reactivity control. 1. Enter 1 in space 80 if burnable poison rodr are used for reactivity control. 2. Enter 0 in space 80 if burnable poison rods are not used for reactivity control. 2.6.2'.14 Card 14: Method for Stripping Primary Coolant Noble gases are stripped from the primary coolant and ro'uted to the gaseous radwaste systen by one of the following methods: 1. Enter 0 in space 69 if there is continuous purging of the purifica-tion tank in the primary coolant treatment system. Leave spaces 76-80 blank. 2. Enter 1 in space 69 if there is intermittent degassification of the primary coolant letdown flow in a gas stripper. In addition, fill in spaces 76-80 in the following manner: a. If the gas stripper is in the boron bleed system, leave spaces 76-80 blank. 2-20

b. If there is no boron bleed system, but a gas stripper has been pro-vided for intermittent degassification of the primary coolant letdown flow, enter the average flow rate in gal / min through the gas stripper in spaces 76-80. The average flow rate in gal / min is detennined by multiplying the average letdown rate (value entered on Card 4) by the fraction of time the letdown is passed through the gas stripper. 3. Enter 2 in space 69 if there is contim ous degassification of the full primary coolant letdown flow in a gas stripper. Leave spaces 76-80 blank. The total amount of fission gases routed to the gaseous rad-waste system in the ship is calculated in the NMS-GEFF Code. 2.6.2.15 Cards 15-17: Holdup Time for Fission Gases Stripped from Primary Coolant The holdup time for gases stripped from the primary coolant is hand cal-culated. The calculations are based on the fcllowino parameters: 1. Pressurized Storage Tanks a. One storage tank is held in reserve for back-to-back shutdowns, one tank is in the process of filling, and the remainder are used for storage. The NMS-GEFF Code will calculate the effective holdup time for filling and add it to the holdup time for storage. tions are based on a waste gas input flow rate of Calculg/ day (50,000 f t /yr) and a storage tant pressure 70% of b. 3 140 ft the design value, c. If the calculated holdup time exceeds 90 days, assume the remaining gases are released after 90 days. The holdup time (T ) and fill time (T ) are calculated as follows: h f PV T = f 140 T = PV(n-2) h 140 2-21

where n is the number of tanks; n-2 is the correction to subtract the tank being filled and the tank held in reserve; P is the storage pressure, in atmospheres; T is the time reauired to fill one tank, in days: f T is the holdup time, in days; h V is the volume of each tank, in ft (STP); and 140 is the average waste cas flow rate to the storace tank, in ft / day (STP) per reactor. Enter on Card 15 the holdup time, in days, for Xe in spaces 73-80. Enter on Card 16 the holdup time, in days, for Kr in spaces 73-80. Enter on Card 17 the fill time, in days, in spaces 73-80. 2. Charcoal Delay Systems Charcoal delay system holdup times are based on the following equation: T = 0.011 P1K /F where F is the system flow rate, in f t / min; 3 K is the dynamic adsorption coefficient, in cm /a; f1 is the mas,s of charcoal adsorber, in thousands of counds; and T is the holdup time, in days. 2-22

The dynanic adsorption coefficient for Xe and Kr are based on the systen design noted below. DYHAriIC ADSORP', ION COEFFICIENT (cm /g) Operatina 77 F Operatina 77"F Operating 0*F Dew Point 45 F Dew Point 0"F Dew Point -20 F Kr 18.5 25 105 Xe 330.0 440 2410 Enter on Card 15 the hcidup tire, in days, for Xe in spaces 73-80. Enter on Card 16 the holdup tine, in days, for Kr in spaces 73-80. Leave Card 17 blank. Cover Gas Recycle Sjsten a. For this systen or other systems designed to hold gases indefinitely, the calculations are based on a 90-day holdup time. Enter on Card 15 the ho aup time (90 days) for Xe in space: 73-80. Enter on Card 16 cne holdup time (90 days) for Kr in spaces 73-80. Enter on Card 17 the fill time (0 days) in spaces 73-80. 2.6.2.16 Card 18: Primary System Waste Gas Processinc Systen Particulate Release Card 18 identifies the treatment provided for particulate removal from the primary system waste gas processing system effluent. If the effluent is treated through HEPA filters, enter a release fraction of 0.01 in spaces 73-80. If no treatment is provided, enter 1.0. 2-23

2.6.2.!7 Card 19: Auxiliaries Compartment Gaseous Releases Card 19 identifies the treatment provided for fodine and particulate re-moval from the auxiliaries compartment ventilation system effluent. The entries on the input data Card 19 indicate the release fraction of air-borne iodine and radioactive particulates applicable to the effluent. 1. If ventilation exhaust air is treated through charcoal adsoroers enter the appropriate release fraction in spaces 49-53 corresponding to the depth of charcoal used as indicated in Table 2-2. 2. If ventilation exhaust air is treated through HEPA filters, enter release fraction of 0.01 in spaces 76-80. d 3. If no treatment is provided for the ventilation exhaust, enter a 1.0 in spaces 49-53 and 76-80. 2.6.2.18 Card 20: Reactor Containment Free Volume 6 Enter the reactor containment volume (in 10 ft ) in spaces 73-80. 2.6.2.19 Card 21: Reactor Containment Internal Cleanup System 3 3 Enter the flow rate (in 10 f t / min) through the internal cleanup system (charcoal adsarbers) in spaces 73-80. The airborne iodine concentration calculations are based on the following parameters: 1. A primary coolant leakage rate of 0.001%/ day of the primary coolant iodine inventory. 2. Operation of the cleanup systen for 16 hours prior to purging. 3. A DF of 10 (removal fraction of 0.1) for the charcoal adsorber, DF of 100 (removal fractirn of 0.01) for the HEPA filters, and a mixing efficiency of 70%. 2-24 -i. E

TABLE 2-2 ASSIGNED REMOVAL FRACTIONS FOR CHARC0AL ADSORBERS FOR RADIOI0 DINE REMOVAL Release Fraction a Activated Carbon Bed Depth For Iodine 2 inches. Air filtration system designed to operate inside reactor containment. 0.1 2 inches. Air filtration system designed to operate outside the reactor containment and 0.3 relative humidity is controlled at 70%. 4 inches. Air filtration system designed to operate outside the reactor containment and 0.1 relative humidity is controlled at 70%. 6 inches. Air filtration system designed to operate outside the reactor containment and 0.01 relative humidity.is controlled to 70%. aMultiple beds, e.g., two 2-inch beds in series, should be treated as a single bed of aggregate depth. 2-25

4. Continuous leakage of primary coolant durino the operation of the in-ternal cleanup systen. 2.6.2.20 Card 22: Reactor Containment - Number of Periodic Purges Enter the total number of periodic purges in spaces 73-80. 2.6.2.21 Card 23: Reactor Containment Gaseous Releases Durina Periodic Puroes Card 23 identifies the treatment provided for iodine and particulate removal from the reactor containment periodic purging effluent. The entries on the input data Card 23 indicate the release fraction of airborne iodine and radioactive particulates applicable to the effluent. 1. If the purge exhaust air is treated through charcoal adsorbers, enter the appropriate release fraction in spaces 49-53 corresponding to the depth of charcoal used as indicated in Table 2-2, 2. If the purce exhaust air is treated through HEPA filters, enter a release fraction of 0.01 in spaces 76-80. 3. If no treatment is provided for the purge exhaust, enter a 1.0 in spaces 49-53 and 76-80. 2.6.2.22 Card 24: Reactor Containment Continuous Purge Rate Enter the rate (in cfm) at which the reactor containment is continuously purged, if the containment is continuously purged during power operation, in spaces 73-80. 2.6.2.23 Card 25: Reactor Containment Continuous Purge Treatment Card 25 ident,fies the treatment provided for iodine and particlate removal from the reactor containment continuous purge effluent. The entries on the 2-26 5

i l r i input data Card 25 indicate the release fraction of airborne iodine and radioactive particulates applicable to the effluent. 1. If continuous purge exhaust air is treated through charcoal adsorbers enter the appropriate release fractinn in spaces 49-53, corresponding to the depth of charcoal used as indicated in Table 2-2. 2. If continuous purqe exhaust air is treated through HEPA filters, enter l a release fraction of 0.01 in spaces 76-80. 3. If no treatment is provided for the continuous purge exhaust, enter f a 1.0 in spaces 49-53 and 76-80. a 2.6.2.24 Card 26: Steam Generator Blowdown Tank Vent System Iodine Releases Card 26 identifies the treatment provided for iodine removal from the steam I generator blowdown tank vent effluent. The entries on the input data Card 26 1 _i indicate the release fraction of airborne iodine applicable to the effluent. 1. Enter 0.05 in spaces 73-80 if the gases from the steam generator blow-g 3 down tank vent are released without treatment. 4 2. Enter a zero in spaces 73-80 if the cases from the steam generator blowdown tank are vented throuch a condenser prior to release. 3. Enter a zero in spaces 73-80 if the steam generator blowdown tank is vented to the main condenser air ejector. 4. Enter a zero in spaces 73-80 for a once-through steam generator system. 2.6.2.25 Card 27: Fraction of Iodine Released from the Main Condenser Evacuation System Card 27 identifies the treatment provided for iodine removal from the main condenser evacuation system effluent. The entries on the input data Card 27 I indicates the release fraction of airborne iodine applicable to the effluent. 2-27

e 1. If the main condenser evacuation system effluent is treated through charcoal adsorbers, enter the appropriate release fraction in spaces 73-80 corresponding to the depth of charcoal used as ind'cated in Table 2-2. ^ 2. If no treatment is provided for the main condenser evacuation system effluent, enter a 1.0 in spaces 73-80. B E-28 s

CHAPTER 3. PRINCIPAL PARAMETERS USED IN NUCLEAR-POWERED MERCHANT SHIPS ( SOURCE TERM CALCULATIONS AND THEIR BASES i

3.1 INTRODUCTION

The prin-ipal parameters used in the source term calculations have been i j compiled in this chapter so as to indicate the bases for certain of the nominal parameters. Operating data for nuclear-powered merchant ships i at the size (300 Mwt) considered in this report do not exist.

However, a large amount of data are available on reactor operating experience, and laboratory and field tests for land-based pressurized water reactors.

These data are presented and discussed in detail in NUREG-0017{1) The data Fave been used and where possible have been adjusted to apply to the calculations of source terms for nuclear-powered merchant ships. ] In addition, the discussions of references 2 and 3 provide design informa-k tion which was proposed for nuclear-powered merchant ships in the 300 MWt l range. Based on the use of information in References 1, 2 and 3, a listing -r of parameters and their bases was prepared which provides a realistic 2 assessment of reactor and radwaste system operation. = l

3. 2 PRINCIPAL PARAMETERS AND THEIR RASES i

I. 3.2.1 THERMAL POWER LEVEL 3.2.1.1 Parameter 9 The maximum thermal power level (in MWt) evaluated for safety consideration. A 3.2.1.2 Bases The power level used in the NMS-GEFF Code calculation is the maximum power level evaluated for safety considerations, since past experience indicates that the reactor will eventually operate at maximum power. 4

s 3-1

_i

~ -_L. - e .~ ? .-:-n J: 1 3.2.2 PLANT CAPACITY FACTOR

s 3.2.2.1 Parameter i

Plant capacity factor of 90%, i.e., 330 effective full power days per year. 3.2.2.2 Bases ? ] The source tern calculations are based on a plant capacity factor of 90%. - 3 Reference (2) provides a discussion of a typical travel itinerary for a ( - ) nuclear-powered merchant ship. Based on that discussion, it is assumed that a typical nuclear-powered merchant ship will operate at 100% power approximately 90% of the time, i.e., a 90% plant capacity factor. 3.2.3 RADIONUCLIDE CONCENTRATIONS IN THE PRIMARY AND SECONDARY COOLANT 3.2.3.1 Parameter t Tables 3-1 and 3-2 list the expected radionuclide concentrations in the primary and secondary coolant for nuclear-powered merchant ships (NMS) f using recirculating U-tube and once-throuch stean generators, respectively, with nominal desian parameters listed in Tables 3-3 and 3-4. Should the design parameters of a particular NHS be different than those listed in f.- Tables 3-3 or 3-4, adjust the concentrations in Tables 3-1 or 3-2 usina Tables 3-5 through 3-7 and Figures 3-1 and 3-2, as appropriate. = - 3.2.3.2 Bases 8 The radionuclide concentrations, adjustment factors, and procedures for effecting adjustments given in T ^1es 3-1 through 3-7 are based on the i values and' methods given by the American Nuclear Society Standard N237I4} I 3-2 ~ n 9 .S ,-4 '-----'---A C- -'i---sg+-- - - =

• g

-jj' ' ' ' '-'~7 a k. y e 1

TABLE 3-1 NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS OF THE REFERENCE NUCLEAR-POWERED MERCHANT SHIP WITH U-TUBE STEAM GENERATORS (uC1/g) ISOTOPE PRIMARY COOLANT ** SECONDARY COOLANT

• WATER ***

STEAM + Noble Gases Kr-83m 5.1(-3)++ Nil 1.6(-8) Kr-85m 2.6(-2) Nil 8.3(-8) Kr-85 2.1(-3) Nil 6.6(-9) Kr-87 1.5(-2) Nil 4.6(-8) Kr-88 4.8(-2) Nil 1.5(-7) Kr-89 1.3(-3) Nil 4.1(-9) Xe-131m 4.8(-3) Nil 1.6(-8) Xe-133m 2.6(-2) Nil 8.4(-8) Xe-133 1.3(+0) Nil 4.2(-6) Xe-135m 3.3(-3) Nil 1.1(-8) Xe-135 7.4(-2) Nil 2.4(-7) Xe-137 2.3(-3) Nil 7.4(-9) Xe-138 1.1(-2) Nil 3.5(-8) Iodines I-131 6.8(-2) 2.1(-5) 2.1(-7) 1-133 9.6(-2) 2.7(-5) 2.7(-7) Cesium Cs-134 6.4(-3) 7.2(-6) 7.2(-9) Cs-137 4.6(-3) 5.2(-6) 5.2(-9) Other Nuclides Mn-54 7.8(-5) 6.4(-8) 6.4(-11) Fe-59 2.5(-4) 1.9(-7) 1.9(-10) Co-58 4.0(-3) 2.6(-6) 2.6(-9) Co-60 5.0(-4) 2.9(-7) 2.9(-10) Sr-89 8.8(-5) 6.4(-8) 6.4(-11) Sr-90 2.5(-6) 1.3(-9) 1.3(-12) =dasec on a primary-to-secondary leak of 100 lb/ day.

• • The concentrations given are for primary coolant entering the letdown line.
    • The concentrations given are for water in a steam generator.

+The concentrations given are for steam leaving a steam generator. ++5.1(-3) = 5.1 x 10-3 3-3

TABLE 3-2 NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS OF THE REFERENCE NUCLEAR-POWERED MERCHANT SHIP WITH ONCE-THROUGH STEAM GENERATORS (uCi/g) ISOTOPE PRIMARY COOLANT

• SECONDARY C0OLANT**

Noble Gases Kr-83m 5.1(-3) 1.6(-8) Kr-85m 2.6(-2) 8.3(-8) Kr-85 2.1(-3) 6.6(-9) Kr-87 1.5(-2) 4.6(-8) Kr-88 4.8(-2) 1.5(-7) Kr-89 1.3(-3) 4.1(-9) Xe-131m 4.8(-3) 1.6(-8) Xe-133m 2.6(-2) 8.4(-8) Xe-133 1.3(+0) 4.2(-6) Xe-135m 3.3(-3) 1.1(-8) Xe-135 7.4(-2) 2.4(-7) Xe-137 2.3(-3) 7.4(-9) Xe-138 1.1(-2) 3.5(-8) Iodine I-131 6.8(-2) 3.8(-7) I-133 9.6(-2) 5.2(-7) Cesium Cs-134 6.4(-3) 5.9(-8) Cs-137 4.6(-3) 4.4(-8) Other Nuclides Mn-54 7.8(-5) 5.8(-10) Fe-59 2.5(-4) 1.5(-9) Co-58 4.0(-3) 2.3(-8) Co-60 5.0(-4) 2.6(-9) Sr-89 8.8(-5) 5.8(-10) Sr-90 2.5(-6) 1.5(-11) "ine concentrauons given are primary coolant entering the letdown line.

• • Based on primary-to-secondary leakage of 100 lb/ day.

The concen-trations given are for steam leaving a steam generator. 3-4

TABLE 3-3 PARAMETERS USED TO DESCRIBE THE REFERENCE NUCLEAR-POWERED MERCHANT SHIP WITH U-TUBE STEAM GENERATORS PARAMETER SYMBOL UNITS NOMINAL VALUE Thermal power P fuft 300 Steam flow rate FS lb/hr 1.3(6) Weight of water in primary WP lb 1.9(5) coolant system Weight of water in all WS lb 4(4) steam generators Primary coolant letdown flow FD lb/hr 1.3(4) (purification) Primary coolant letdown flow (yearly average for boron FB lb/hr 0 control) Steam generator blowdown flow FBD lb/hr 6.5(3) (total) Fraction of radioactivity in blowdown steam that is not t1BD 0.9 returned to the secondary coolant system

• Flow through the primary coolant treatment system FA lb/hr 1.3(3) cation demineralizer Ratio of condensate demineralizer flow rate to the total steam NC 0.65 flow rate Ratio of the total amount of noble gases routed to gaseous radwaste from the primary coolant treatment system to Y

0.25 the total amount of noble gases routed from the primary coolant system to the primary coolant treatment system

• NBD = 1.0 - h where DFBD is the decontamination factor (DF) that the iodine in the steam generator blowdown undergoes in its treatment.

(See Section 2.6.2.10). 3-5

TABLE 3-4 PARAMETERS USED TO DESCRIBE THE REFERENCE NUCLEAR-POWERED MERCHANT SHIP WITH ONCE-THROUGH STEAM GENERATORS PARAMETER SYMBOL UNITS NOMINAL VALUE Thermal power P MWt 300 Steam flow rate FS lb/hr 1.3(6) Weight of water in primary WP lb 1.9(5) coolant system Weight of water in all WS lb steam generators Primary coolant letdown flow FD lb/hr 1.3(4) (purification) Primary coolant letdown flow (yearly average for boron FB lb/hr 0 control) Flow through the primary coolant. treatment system FA lb/hr 1.3(3) cation demineralizer Ratio of condensate demin-eralizer flow rate to the NC 0.65 total steam flow rate Ratio of the total amount of noble gases routed to gaseous radwaste from the primary coolant treatment system to Y 0.25 the total amount routed from the primary coolant system to the primary coolant treatment system.

• The secondary coolant inventory is not of importance in a once-through steam generator plant because decay is not an important removal mechanism:

WS therefore cancels from the adjustment factors of Table 3-7. 3-6

TABLE 3-5 VALUES USED IN DETERMINING ADJUSTMENT FACTORS FOR NUCLEAR-POWERED MEROiANT SHIPS SYMBOL DESCRIPTION ELEMENT CLASS NOBLE OTHER GASES IODINE CESIUM NUCLIDES NA Fraction of material removed in passing 0.0 0.0 0.9 0.9* through the cation demineralizer NB Fraction of material re-moved in passing through 0.0 0.9 0.5 0.9* the purification demineralizer R Removal rate - primary 0.017 0.06 0.037 0.06* coolant (Hr 1)** NS Ratio of concentration in steam to that in water in the steam generator U-tube steam generator + 0.01 0.001 0.001 Once-through steam generator + 1.0 1.0 1.0 NX Fraction of activity removed in passing through the con-0.0 0.9 0.5 0.9 densate demineralizers r Removal rate - secondary coolant (Hr 1)* U-tube steam generator + 0.34 0.09 0.16 Once-through steam generator + (a) ( b) ( a) FL Primary-to-secondary leakage 100 100 100 100 (1b/ day)

• These represent effective removal terms and include mechanisms such as plateout. Plateout would be applicable to corrosion product nuclides.
  • These values of R apply to the reference case NMS whose parameters are given in Tables 3-3 and 3-4 and have been used in developing Tables 3-6 and 3-7.

For HMS not included in Tables 3-3 and 3-4, the appropriate value for R may be determined by the following equations: R = FB + FD)(Y) for noble gases R5 (FD)(NB) + {1 - N3)(FB + (FA)(NA)) for Iodine, Cesium and other nuclides. 5 5 (a )- 7.6 x 10 /WS (b)- 4.23 x 10 fg3 3-7

TABLE 3-5 (continued) +fioble gases are rapidly transported out of the water in the steam generator and swept out of the vessel in the steam; therefore, the concentration in the water is negligible and the concentration in the steam is approximately equal to the ratio of the release rate to the steam generator and the steam flow rate. These noble gases are removed from the sy. stem at the main condenser. ++These values of r apply to the reference case ftMS whose parameters are given in Tables 3-3 and 3-4 and have been used in developing Tables 3-6 and 3-7. For NMS not included in Tables 3-3 and 3-4, the appropriate value for r may be determined by the following equation: r=(FBD)(flBD)+(fjS)(FS)(NC)(flX)forlodine, Cesium,andother W3 nuclides. 3-8

TABLE 3-6 ADJUSTMENT FACTORS FOR NUCLEAR-POWERED MERCHANT SHIPS WITH U-TUBE STEAM GENERATORS SECONDARY COOLANT ELEMENT CLASS PRIMARY COOLANT (f)* WATER STEAM 6 Noble gases 633P 0.017 + A ** 1.3 x 10 Y R+A-FS 4 4 Iodine 633P 0.06 +A 4 x 10 0.34 + A 4 x 10 0.34 + A.f 7 T R+L WS r+X WS r+A 4 4 Cesium 633P 0.037 + A 4 x 10 0.09 + A 4 x 10 0.09 + A 7 f T R + A. W5 r+A WS r+K 4 4 Other Nuclides 633P 0.06 + A 4 x 10 0.16 + A 4 x 10 0.16 + 1 Y R + A. WS r + A. f WS r + A. 7

• f 1s the primary coolant adjustment factor and is used in the secondary coolant adjustment factors.

is the isotopic decay constant (hr-1), TABLE 3-7 ADJUSTMENT FACTORS FOR NUCLEAR-POWERED MERCHANT SHIPS WITH ONCE-THROUGH STEAM GENERATORS NUCLIDE PRIMARY COOLANT (f)* SECONDARY COOLANT 6 633P 0.017 + A 1.3 x 10 f Noble gases T R+A FS 5 Iodine 633P 0.06 + 1 7.6 x 10 Y R+A rWS 5 Cesium 633P 0.037 +A 4.2 x 10 Y R+A rWS 5 Other nuclides 633P 0.06 + A. 7.6 x 10 Y R+A rWS

• f is the primary coolant adjustment factor and is used in the secondary coolant adjustment factors.

3-9

LEAKAGE d FS NS RE^ T GENERA ORS GENERA ORS ESSEL CON ENSER PRIMARY SECONDARY FS(1-NC) SIDE SIDE / g FEEDWATER k y NC-FS FD BLOWDOWN FBD rs CONDENSATE DEMINERALIZER NX Y' FA PU RIFICATION CATION ^ DEMINERAllZER ~ DEMINERALIZER SYMBOLS ARE DEF.lNED IN TABLES 3-3 AND 3 5 NB FB y > TO BORON BLEED SYSTEM Y PURIFICATION FIGURE 3-1 Removal Paths For Nuclear Powered Merchant Ship With U-Tube Steam Generators

LEAKAGE FS NS STEAM STEAM REACTOR STEAM VESSEL GENERATORS GENERATORS MAIN CONDENSER PRIMARY SECONDARY FS(1-NC) SIDE SIDE l 3 FEEDWATER k y NC-FS FD CONDENSATE DEMINER AllZER NX Y FA PURIFIC ATION CATION ~ DEMINERAllZER ~ DEMINERAllZER SYMBOLS ARE DEFINED IN TABLES 3-4 AND 3-5 NB FB ~

• TO BORON BLEED SYSTEM y

Y PUR F ATION FIGURE 3 2 Removal Paths For Nuclear Powered Merchant With Once-Through Steam Generators

and by NUREG-0017. (1) The data in Tables 3-1 through 3-7 were obtained by adjusting the corresponding data in N237 and NUREG-0017 to the NSSS parameters of a reference case nuclear-powered merchant ship (NMS). The values in Tables 3-1 and 3-2 provide a set of reference case radionuclide concentrations in the primary and secondary coolant systems for NMS designs with the reference case system parameters specified in Tables 3-3 and 3-4. The values in Tables 3-1 and 3-2 are considered to be representative of radionuclide concentrations in an NMS over its lifetime based on currently available data and it.odels. The values in Tables 3-3 and 3-4 were judged to be representative of typical NHS design based on references 1, 2 and 3. It is recognizeo that the nominal values selected and used in Tables 3-3 and 3-4 will not be applicable in all cases. For that reason, a means of adjustino the concentrations to the actual design parameters of a particular nuclear-powered merchant ship has been provided in Tables 3-5 through 3-7. The adjustment factors in Tables 3-5 through 3-7 are based on the following expression: w( A R)k where C is the specific activity (in uCi/g), k is a conversion factor, 454 g/lb, R is the removal rate of the isotope from the system due to demineralization, leakage, etc. (hr-1), s is the rate of release to and/or production of the isotope in the system (in uCi/hr), I w is the fluid weight (in Ib), and I lis the decay constant (hr-1). 3-12

The following sample calculations illustrate the method by which the NMS-GEFF Code will adjust the radionuclide concentrations in Tables 3-1 and 3-2. The computer output of these sample calculations is shown in the sample case in Section 4.3. As indicated in Tables 3-6 and 3-7, adjust-ment factors will be calculated for noble gases, iodine, cesium, and other nuclides. Table 3-8 lists the parameters of a sample nuclear-powered merchant ship design using a U-tube steam generator. Table 3-8 also compares these parameters with the nominal values shown in Table 3-3. As can be seen from Table 3-8, certain of the parameters for the sample NMS design are different from those for the nominal case. Therefore, the primary coolant activity is recalculated using the actual design value for all parametgrs and the methods described below. 1. Noble Gases (Xe-133 is used as an example) Using the equation for noble gases in Table 3-6, the adjustment factor, f, is calculated as follows: f= 633 0.017 + 1 (1) WP R + A. where the terms in the ecuation are defined in Tables 3-3 and 3-5. In calculating f, the variable R is calculated first by using the equation given in Table 3-5 for noble gases R = FB + (FD)(Y) (2) WP where the tenns of the equation are as defined in Tables 3-3 and 3-5. 3-13

TABLE 3-8 COMPARISON OF SYSTEM PARAMETERS FOR A SAMPLE t#is DESIGN WITH THAT OF Tr1E fl0MINAL VALUES OF TABLE 3-3 SAMPLE NOMINAL VALUE I PARAMETER VALUE (TABLE 3-3) Thermal poter level, in MWt 330 300 6 6 Steam fiow rate, in lb/hr 1.6 x 10 1.3 x 10 5 5 Mass of primary coolant, in lb 2.0 x 10 1.9 x 10 4 4 Water weight in all steam generators, 6.0 x 10 4.0 x 10 in lb 4 4 Primary coolant letdown, in lb/hr 1.8 x 10 1.3 x 10 Boron bleed rate - yearly average, 140 0 lb/hr Steam generator blowdown flow, in lb/hr 4000 6500 Fraction of blowdown activity not 0.9 0.9 returned to secondary steam Cation demineralizer flow, lb/hr 0.0 1300 Condensate demineralizer flow fraction 0.65 0.65 Y (see definition in Table 3-3) 0.25 0.25 3-14

Use the parameters given above in Table 3-8 and the noble gas parameters given in Table 3-5 and substitute in Equation (2) above. 4 R = 140 + (1.8 x 10 x 0.25) = 0.023 5 2 x 10 Use the value of R in Equation (1) above. -3 f= 633 x (330) 0.017 + 5.5 x 10 = 0.82 5 -3 2 x 10 0.023 + 5.5 x 10 The adjusted Xe-133 primary coolant concentration = (adjustment factor) x (nominal Xe-133 concentration) = 0.82 x 1.3 uCi/g = 1.1 uCi/g as shown in the computer output for Xe-133 in Section 4.3. 2. Iodin _e (I-131 is used as an example) Using the ecuation for iodine in Table 3-6, the adjustment factor, f, is calculated as follows: f= 633 0.06 + h (3) WP R + h. where the terms in the equations are defined in Tables 3-3 and 3-5. In calculating f, the variable R is calculated first by u,ing the equation given in Table 3-5. R = (FD)(NB) + (1 - NB)(FB + (FA)(NA)) (4) WP where the terms in the equation are as defined in Tables 3-3 and 3-5. Use the parameters given in Table 3-8 above and the iodine parameters given in Table 3-5 and substitute in Equation (4) above. 3-15

4 R = (1.8 x 10 x 0.9) + (1 - 0.9)(140 + (0.0)(0.0)) = 0.081 5 2 x 10 Use the value of R in Equation (3) above. -3 f = 633(330) 0.06 + 3.6 x 10 = 0.8 5 -3 2 x 10 0.081 + 3.6 x 10 The adjusted I-131 concentration = (adjustment factor) x (noninal I-131 concentration) = 0.8 x 0.068 uti/g = 0.055 uCi/g as shown in the computer output for I-131 in Section 4.3. 3. Cesium (Cs-137 is used as an example) Using the equation for Cesium in Table 3-6, the adjustment factor, f, is calculated as f.ollows: f = 633P 0.037 +1 (5) WP R+A where the tems in the equation are as defined in Tables 3-3 and 3-5. In calculating f, the variable R is calculated first by using Equation (4) above. The cesium parameters given in Table 3-5 and the parameters given in Table 3-8 are used in the equation. R = (1.8 x 10 x 0.5) + (0,5)(140 + (0.0)(0.9)) 0.045 = 5 2 x 10 Use the value of R in Equation (5) above. f = 633(330) 0.037 + 2.6 x 10-6 = 0.86 5 2 x 10 0.045 + 2.6 x 10-6 3-16

The adjusted Cs-137 concentration = (adjustment factor) x (nominal Cs-137 concentration) -3 = 0.86 x 4.6 x 10- uCi/g = 4 x 10 uCi/g as shown in the computer output fo* Cs-137 in Section 4.3. 4. Other tjuclides (Sr-89 is used as an example) Using the equation for other nuclides in Table 3-6, the adjustment factor, f, is calculated as follows: f = 633 0.06 + A (6) WP R+X where the terms in the equation are as defined in Tables 3-3 and 3-5. In calculating f, the variable R is calculated first by using Equation (4) above. The parameters for other nuclides given in Table 3-5 and the para-meters' given in Table 3-8 are used in the equation. R = (1.8 x 10 )(0.9) + (1 - 0.9)(140 + (0.0)(0.9)) = 0.081 5 2 x 10 Use the value of R in Equation (6) above. f = 633(330) 0.06 + 5.7 x 10-4 = 0.81 5 2 x 10 0.081 + 5.7 x 10-4 The adjusted concentration of Sr-89 = (adjustment factor) x (nominal Sr-89 concentration) 5 5 = 0.81 x 8.8,x 10 uCi/9 = 7 x 10 uCi/q as shown in the computer output for Sr-89 in Section 4.3. 3-17

A similar method is used in the NMS-GEFF Code to adjust secondary coolant concentrations for nuclear-powered merchant sbf ps with parameters different than those specified in Tables 3-3 and 3-4. The primary coolant concentrations listed in Tables 3-1 and 3-2 are based on 0.12% fuel cladding defects and the escape and release rate coefficients discussed in Section 2.2.3 of NUREG-0017. In that document, a value of 0.12% is considered to be the weighted average of the fuel performance data at land-based operating pressurized water reactors and is considered to be representative of expected operation. In addition, Section 2.2.3 of NUREG-0017 presents escape rate coefficients used in conjunction with 0.12% fuel cladding defects to calculate the rate of entry of fission . products into the primary coolant. The secondary coolant concentrations are based on 100 lb/ day primary-to-secondary leakage. The primary to secondary leakage rate experience for operating lanu based PWRs, is given in Section 2.2.3 of NUREG-0017. For recirculating U-tube steam generators, carryover due to mechanical entrainment is based on 0.10% moisture in the steam. This value is based on the average of data for moisture carryover at several operating land-based PWRs that use recirculating U-tube steam generators as presented in Section 2.2.3 of NUREG-0017. It is assumed that 1% of the iodine is carried-over with the steam in recirculating U-tube steam generators. 3-18

n n-i-L'M-'--li.- For once-through steam generators, it is assumed that 100% of both non-volatile and volatile species are carried over with the steam since this type of steam generator has no liquid reservoir and 100% of the feed is i converted to steam. It is also cssumed that 5% of the iodine entering the steam generator from the primary system is in a volatile form, based on iodine species data + gathered in the primary coolant for four operating land-based PWRs as pre-sented in NUREG-0017. The volatile iodine will behave similarly to a noble gas at steam generator operating temperatures and have a partition factor of 0.15 at main condenser operating temperatures. The bases for the main l condenser partition factor are discussed in Section 3.2.7 of this report. 3.2.4 LEAKAGE RATE TO REACTOR CONTAIHMENT l 3.2.4.1 Parameter Daily leakage rate of 1% of the noble gas inventory and 0.001% of the iodine 1 inventory in the primary coolant is released to the reactor containment a tnosphere. 3.2.4.2 Bases A The daily reactor coolant leakage rates to the reactor containment are = based on measurements made at operating land based PWRs and presented in 5 Section 2.2.4 of NUREG-0017. These measurements are based on Xe-133 and i 1-131 concentrations in the containment atmosphere at operating land-2 _j based pressurized water reactors. Using the concentration data, the con-1i tainment volumes and the Xe-133 and I-131 primary coolant concentrations, I g 3-19 __i.

-i----i==i

.--..i-.m.mmmimanni -n

respectively, it was determined that 1%/ day of the noble gas inventory in the primary coolant is released to the reactor :ontainment atmosphere, and that 0.001%/ day of the iodine inventory in the primary coolant is leaked to the reactor containment atmosphere. 3.2.5 AUXILIARIES COMPARTMENT LEAKAGE 3.2.5.1 Parameter l 160 lb/ day primary coolant with an iodine partition factor of 0.vJ75. 3.2.5.2 Bases The source term calculation is based on an assumed primary coolant leakage rate of 160 lb/ day (20 gal / day). In the absence of available data, this value is based on engineering judgment and is consistent with values pro-posed in Environmental Reports for land-based PWRs. As indicated in Section 2.2.5 of NUREG-0017, the partition factor of 0.0075 is based on the fact that 5% (0.05) of the iodine in primary coolant is assumed to be in the volatile species (see Section 3.2.3.2) and on laboratory data which indicated a partition coefficient of 0.15 for volatile iodines. Applying the partition coefficient of 0.15 to the volatile fraction (0.05) of primary coolant iodine, a partition factor of 0.0075 is obtained. 3.2.6 ENGINE COMPARTMENT LEAKAGE (TURBINE AREA) 3.2.6.1 Parameter Leakage rate of 1700 lb/hr and an iodine partition coefficient of 1.0 1 fo* all iodine species. 3-20

3.2.6.2 Bases It is assumed that steam will leak to the engine compartment atmosphere at a rate of 1700 lb/hr. The leakage is considered to be from many sources, each too small to be detected individually, but which, taken collectively, total 1700 lb/hr. The most significant leakage pathway is considered to be leakage through valve stem packings. The partition factor for iodine is assumed to be

1. 0.

The leakage is considered to occur as a vapor, i.e. as steam or hot liquid flashing to steam, which causes the iodine to remain airborne as a gas or an aerosol. 3.2.7 NAIN CONDENSER EVACUATION SYSTEM PARTITI0'N FACTOR FOR IODINE 3.2.7.1 Parameter Iodine partition factor of 0.15 for volatile species and zero for nonvolatile species. 3.2.7.2 Bases Based on laboratory data presented in Section 2.2.7 of MUREG-0017, an iodine partition factor of 0.15 is assumed for volatile iodine species in the main condenser air evacuation system. Since nonvolatile species will be removed by mechanical entrainment in the offgas stream and since the fraction of secondary coolant lost as entrained moisture through the air ejecto s is 1 negligible, a partition factor of zero is used for nonvolatile iodine species. 4 3-21

3.2.8 STEAfl GENERATOR BLOWDOWN TANK VENT i i 3.2.8.1 Parameters j 1. Iodine partition factor of 0.05 if the tank is vented directly to the atmosphere, i.e., steam is not condensed prior to venting. ~ 2. Iodine partition factor of zero if the blowdown tank is vented through 4 a condenser (tank vent condenser or main condenser) or if the blowdown is cooled below 212 F, i.e., flashing is not used for heat removal. [ I 3.2.8.2 Bases Based on data presented in Section 2.2.8 of NUREG-0017, an iodine partition factor of 0.05 is assumed for iodine released to the atmosphere when the blowdown flashes to steam in a blowdown flash tank. If provisions are made to prevent flashing (cooling blowdown below 212 F) such as by passing the l I blowdown flow through a heat exchanger for cooling, or if the flashed steam leaving the blowdown flash tank is condensed, a partition factor of zero s is used. 3.2.9 CONTAINf1ENT PURGE FREQUENCY ~ 3.2.9.1 Parameter j Releases are based on 4 purges per year for maintenance in the reactor con-tainment and for refueling. In addition, releases are based on either [ continuous or additional periodic purges of the containment as indicated by the NMS designer. If there is a continuous Furge system for the reactor j containment, the N?tS designer shall specify the size of this system in cfm. ^{ If there is a periodic release the number of purges shall be specified. The number of periodic purges at full power is assumed to be 20. i 3-22

3.2.9.2 Bases It is assumed that the reactor containment is purged four times /yr for refueling and maintenance in both the cold shutdown condition and the hot standby condition. In addition, experience at operating land-based PWRs presented in Section 2.2.10 of NUREG-0017, has indicated the need to vent the reactor containment frecuently during full power operation to control the containment pressure, temperature, humidity and airborne activity l evel s. For these reasons, the source term calculations should include the effect of these releases, either continuous or periodic, during full power operation. Therefore, the releases should consider a continuous containment ventilation rate specified by the designer along with the 4 purges /yr in-dicated above', or the releases should consider frequent periodic purges of the containment at full power along with the 4 purges /yr indicated abovE. A freauency of 20 purges /yr during power operation is considered to be representative of the operating experience accumulated at land-based pressurized water reactors. 3.2.10 REACTOR CONTAINMENT INTERNAL CLEANUP SYSTEM 3.2.10.1 Parameter Assume the internal cleanup systen will operate for 16 hours prior to purging with an iodine DF of J0 (removal fraction of 0.1) for charcoal adsorbers, a particulate DF of 100 (removal fraction of 0.01) for HEPA filters, and a mixing efficiency of 70%. 3.2.10.2 Bases Internal cleanup systems may be used to

• educe airborne iodine concentra-tions in the reactor containment atmosphere prior to purging.

Such systems 3-23

normally recirculate containment air through HEPA filters and charcoal adsorbers to effect iodine and particulate removal. For source term cal-culations, it is assumed that the cleanup systems are operated for 16 hours prior to purging. It is considered that charcoal adscrbers provide a DF of 10 for iodine, that HEPA filters provide a DF of 100 for particulates, and that the recycle mixing efficiency is 70%. The system operation time of 16 hours considers that two shifts will elapse following a decision to enter ( thc containment. The time period of two shifts is a reasonable amount of time for pre-entry preparations. The 70% mixing efficiency is based on containment iodine concentration data taken at a land-based pressurized water reactors before and af ter a cleanup as shown in Section 2.2.11 of NUREG-0017, 3.2.11 DECONTArilNATION EFFICIENCIES FOR CHARC0AL ADSORBERS AND HEPA FILTERS 3.2.11.1 Parameter For charcoal adsorbers use a decontamination efficiency for iodine removal corresponding to the appropriate charcoal bed depth listed in Table 3-9. The release fraction corresponding to each decontamination efficiency is also listed in Table 3-9. These release fractions are to be entered on Cards 19, 23, 25 and 27 as appropriate. For HEPA filters use a decontamination effi-ciency of 99% (release fraction of 0.01) for particulate removal. 3.2.11.2 Bases The iodine removal efficiencies given in Table 3-9 are based on data given in USNRC Regulatory Guide 1.140,(5) "Desic, Testing and Maintenance Criteria } o 3-24 mienmiiminis-imi-umminimimiiemisii-smummieisin -i -umi-mm --mm sin i-uminimummmmmi-i-minims -in-un isnamni-

TABLE 3-9 ASSIGNED DECONTAMINATION EFFICIENCIES AND REMOVAL FRACTIONS FOR CHARCOAL ADSORBERS FOR RADI0 IODINE REMOVAL ASSIGNED ACTIVATED CARBON a ACTIVATED CARBON DECONTAMINATION RELEASE FRACTION BED DEPTH FOR RADI0 IODINE

• FOR IODINE **

2 inches. Air filtration system designed 90% 0.1 to operate inside reactor containment 2 inches. Air filtration system designed 70% 0.3 to operate outside the reactor containment and relative humidity is controlled at 70% 4 inches. Air filtration system designed 90% 0.1 to operate outside the reactor containment and relative humidity is controlled at 70% 6 inches. Air-filtration system designed 99% 0.01 to operate outside the reactor containment and re'lative humidity is controlled to 70%.

• From USHRC Regulatory Guide 1.140, Table 2.
  • Release fraction (RF) = 1.0 Decontamination Efficiency 100.

aMultiple beds, e.g., two 2-inch beds in series, should be treated as a single bed of aggregate depth. !i 3-25

for Nomal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants." In order to use the decontamination efficiencies of Table 3-9, the adsorbers and filters should be designed and tested to the criteria of USNRC Regulatory Guide 1.140. 3.2.12 HOLDUP tit!ES FOR CHARC0AL DELAY SYSTEtiS 3.2.12.1 Parameter T = 0.011f1K/T where F is the system flow rate, in f t / min; K is the dynamic adsorption coefficient, in cm /g; M is the mass of charcoal adsorber, in thousands of pounds; and T is the holdup time, in days. 3 DYNAMIC ADSORPTION COEFFICIENT (cm fg) Operating 77 F Operating 77 F Operating 0 F Dew Point 45 F Dew Point 0 F Dew Point -20 F Kr 18.5 25 105 Xe 330.0 440 2410 3.2.12.2 Bases Charcoal delay systems are evaluated by using the preceeding equation and dynamic adsorption coefficients. T = MK/F is a standard eauation for the calculation of delay time in charcoal adsorption systems. The dynamic adsorption coefficients (K values) for Xe and Kr are dependent on operating l 3-26 1

temperature and moisture content in the charcoal, as indicated by the values in the above parameter. The K values used represent a composite of data from operating reactor charcoal delay systems and reports concerning charcoal adsorption systems presented in Section 2.2.13 of NUREG-0017. The coefficient 0.011 adjusts the units and is calculated as follows: 3 3 -5 3 3 T (days) = f1(10 lb)K(cn /g)(454 g/lb)(3.53 x 10 ft /cm ) F(f t / min)(1440 min / day) f T = 0.011 3.2.13 RADI0 ACTIVE PARTICULATE RELEASES FOR GASEOUS EFFLUENTS 3.2.13.1 Parameter Use the radioactive particulate release rate for gaseous effluents niven in Table 3-10. 3.2.13.2 Bases Section 2.2.15 of NUREG-0017 presents a discussion of particulate releases from pressurized water reactors. This data in NUREG-0017 is based on operating experience at a number of PWRs. The data in Table 3-10 is obtained in the following manner: P C gg3 NMS = 9 j C PWR where P is the particulate release from a nuclear powered merchant ship gpq3 for an isotope, i, as given in Table 3-10, Ci/yr. l 3-27

TABLE 3-10 PARTICULATE RELEASE RATE Ifl GASE0US EFFLUEf4TS Ci/yr per reactor PRIMARY SYSTEM REACTOR AUXILIARIES WASTE GAS fiUCLIDE CONTAlflMEllT COMPARTMErlT PROCESSIf4G SYSTEM Mn-54 0.0055 0.0045 0.011 Fe-59 0.0019 0.0015 0.00038 Co-58 0.019 0.015 0.0038 Co-60 0.0085 0.0068 0.0017 Sr-89 0.00043 0.00033 0.00008 Sr-90 0.00008 0.00005 0.00002 Cs-134 0.0055 0.0045 0.0011 Cs-137 0.0095 0.0075 0.0019 1 3-28

i P is the particulate release from a PWR for an isotope, i, as ppg I given in Table 2-17 of NUREG-0017, Ci/yr. C is the primary coolant concentration for the isotope 1, for g793 the nuclear-powered merchant ship as given in Table 3-1 or 3-2, uCi/gm. C is the primary coolant concentration for the isotope, i, for PWR g a PWR as given in Table 2-2 or 2-3 of NUREG-0017, uCi/gm. 3.2.14 TRITIUt1 RELEASES 3.2.14.1 Parameter The tritium produced in the reactor and available for release through the gaseous pathway is 120 Ci/yr for a 30011Wt nuclear-powered merchant ship which does not use burnable poison rods for reactivity control (continual boron bleed is used ir tead). For a 300 ifWt nuclear-powered merchant ship which uses burnable poison rods the tritium release is 600 Ci/yr. 3.2.14.2 Bases y The tritium released in a land-based PWR has been found to be approximately 0.4 Ci/yr/t1Wt based on operating experience as discussed in Section 2.2.19 f of NUREG-0017. For a 300 f1Wt Nf15 this would be 120 Ci/yr. This tritium is considered to be produced primarily by ternary fission and neutron r 3-29 -i i _a T i

p s f ~ 1 ' ,.s

r... _ -

--- u 1_ n - - ~ - - ~.~~- ~<~- -"- ~ - ..h a c.! activation of boron-10 in the coolant In estimating the tritium release from a nuclear-powered merchant ship which does not use burnable s. poison rods we have used the parameters of Section 2.2.19 of NUREG-0017 to calculate tritium produced by ternary fission and boron-10 activation. [ r k In Section 2.2 above, it was indicated that burnable poison rods may be used for reactivity control in nuclear-powered merchant ships and that this is considered as the reference case method of reactivity control. j Activation of the boron in the burnable poison rods forms lithium-7 which

• I.

i-in turn produces tritium as a result of neutron activation. Based on data presented in reference 2, we estimate that the tritium produced by lithium ~ activation in the burnable poison rods is approximately 480 Ci/yr. Since Section 1.0 indicates that there will not be any liquid discharges 7 from the ship except in its home port, we have assumed that, in order to control tritium levels in the coolant, all the tritium produced is b released to the gaseous pathway either throuah system leakage or the use 3 of an atmospheric evaporator. g 3.2.15 CARBON-14 RELEASES 3.2.15.1 Parameter The annual ouantity of carbon-14 released from a nuclear-powered merchant ship is 1 Ci/yr. It is assumed that the carbon-14 will form volatile compounds ( that will be collected in the waste gas processing system through degassing of - {,, J the primary coolant and will be released to the environment via the plant vent. a ~ 3-30 I , j .,,---,9.....g ._.-._...-.;-----.----y..- +

1 3.2.15.2 Bases The principal source of carbon-14 is the thermal neutron reaction with oxygen-17 in the primary water coolant. As indicated in Section 2.2.26 of NUREG-0017, the carbon-14 production is a function the mass of water in the reactor core, the thermal neutron flux, the thermal neutron cross section for oxygen-17, the number of atoms of oxygen-17 in water the 3 specific activity of carbon-14 and the irradiation time. Of these variables only the mass of water in the reactor core and the thermal neutron flux are dependent on the specific reactor size. The reactor core of the 300 MWt NMS is somewhat smaller than the 3400 MWt nressurized water reactor analyzed in NUREG-0017. Based on reactor core and thermal neutron flux data presented in Reference 2, we estimate the release of carbon-14 from an NMS to be approximately 1 Ci/yr. 3.2.16 ARGON-41 RELEASES 3.2.16.1 Parameter The annual cuantity of argon-41 released from a nuclear-powered merchant ship is 10 Ci/yr. The argon-41 is released to the environment via the reactor containment vent when the containment is vented or purged. 3.2.16.2 Bases Argon-41 is formed by neutron activation of stable naturally occurring argon-40 in the reactor containment air surrounding the reactor vessel. The argon-41 is released to the environment when the containment is vented or 3-31

purged. Section 2.2.27 of NUREG-0017 presents data on releases of argon-41 from operating land-based reactors. Based on the average of that data, we estimate the annual release of argon-41 to be 10 Ci/yr. 3.2.17 GUIDELINES FOR ROUNDING OFF NUtiERICAL VALUES In calculating the estimated annual release of radioactive materials in gaseous wastes, round off all numerical values to two significant figures. 3-32

CHAPTER 4. INPUT FORF%T, SAMPLE PROBLEM, AND FORTRAN LISTING OF THE NMS-GEFF CODE

4.1 INTRODUCTION

This chapter contains additional information for using the NMS-GEFF Code. Chapter 2 of this document describes the entries required on input data cards. Section 4.2 contains a sample input data sheet to orient the user in making the entries descibed in Chapter 2. Section 4.3 of this chapter contains a listing of the input data cards for a sample problem and the resultant output for that sample problem. Section 4.4 contains a discussion of the nuclear data library used and a FORTRAN listing of the NMS-GEFF Code. 4.2 INPUT DATA SHEET The following pages show (1) the form in which data should be entered on input data sheets and (2) a sample completed sheet. 4-1

NAME U.S.NLICLEAR REGULATORY COMMISSION DRTE SUBMITTED ADP TRANSCRIPTION SHEET DATE NEEDED PROGRAM 00000000011111111112222222222333333333344444444445555555555 GGGGrGGGGG 777777???78 1 2 3 4 5 0 7 8 2 Q ! 2 3 1 5 6 ? ? ? 0 1 1 3 4 5 6 [ 9 3 0 1 2 1 4 'i E Z 1 ? Q l C 3 4 5 C ? 3 1 0 1 1 1 1 5 E ? 9 ? 0 1 2 3 1 5 E ?. 9 2 0 1 2 3 4 5 R ? 9 1 0 1 C A R D 1 N A N E N A H E O F S H I P [ ] 2 C A R D 2 P 0 W T H T H E P M A L P O W E R L E V E L (.M r G A. W A T_ T.S J_ ] 3 C A R D 3 P C V 0 L M A 5 (- 0 F P R I M A R f C 0 0 L A N T ( T H0 U 5 A N U S t 8 S } L ] 4 C A R D A _ L E LQ R N E F l F AR Y _ 1 1 5 L C ti L E I O Q W4 R 4 [ [ ( _G P h J .) r!a i 5 a t i z r R r t o u r_G P r 3 t c = Ro s C a r i R t r T o o w N C A T ro N D e n i N C_A R Q _ 6_ A D L.f.h h H M E. 1 R_ 0 f _ _SJ 11s F G.L NL L A3D 1 5 _ __[ ) b I E ] 7 C A P o 7 T o s T r t T O T A L s T E A n r t o w t i t t iO N t B t / HP 3 1 D 1 ft_ E e _C L. SL E A M G E N ER A T 0R { T H_ 0 V 5 A _N D L 8 S J _ _( O C A R D 9 B L W D W N S T E A fl G E N E RI A T 0 R B L 0 W D OW N ( I HC U S A N D L B S / H R ) I ) 0fu 3 C A.R D a W u l M A S $ 0 f L 1 ] 10 cA R D t o__ 3.l t A r G t N 1 p A t Q a B_ t p W D 0w N D E C O N r A C T 0 9 r 0 R i 0 n i N t _ { _ j 9 A P D 1 1 F F C D M C ON E EN s A T F DE M r N E P A t i Z' E R F E 0 W f A A C l i O N l l Akk []j A 12 c A 8 0 1 2 s s L o n B o r 0 F B L E E D 8 A T [ lG P DO [ c a p o i e U P A e t t P 0 r S O N 9 0 0 5 r 0 R R t A C T i v i T T C O N T R 0 t ? i i r v E S o i r i n 0 13 y , _4 , J _ g g 2 (,]_ f..L 0 W _ b _ ) C_b.8 0 U I .T F 0 0 0 F S ] R. I [ E ! N G _ f R [ M 4. 8 Y _( Q Q LA N [ 1 3 g U 1 0 l 1 15 c A R D i s T A U i H 0 t t: U P T i s E r 0 2 x c N e N. r o n ys i 16 c A R. a l e t A u z n o t t u i' . t i n t _r D a K R Y E I n n._ ( B A Y. S l _ __( _ ) b l 17 C A R D 1 7 T A U r i t t T i e r O I D C C AY T A N V 5 f 9A Y % i 18 C A R D 1 8 GA S W A S T E 5 V S T E H P AR T I C U L A T E R E L E A S E F R A C T I O N [ j Ch l 13 C A R D 1 9 A U X I L i A R I Et C M P 1 I OD I N E R E t f R A C ] P A h h i C h L A T E [R L R A 20 C A. R D 2 0 C Q N V 0 L. R ( A C I Q R C O R J A I N M E N I V 0 L U M t _ [ M I L L I O N_ C U _f I, ) ( _ ) I 21 C A R D 2 1 C F M R E A C I O P C N T M T A T M0SP H E P F C L E AN U P R p I t f I H 0 U s AN D r F y ) l r E U P Q E $ / I C A R [ ) 22 C A R D 2 2 R f A C I D R E N J,N I P E R I D3 L C Y Q L_ [ U R G E I f J P A D T i C Ui A T E P ft r PA C (_ d 23 c A n D 2 3 P E R i O D i C PU P c t i o n i N r R e i r P A C I d 2N C A R D 2 4 _ R E A C 1 0 R C h l H I L a ).L b_ R A I E ( CEM)__ 1 3 11 0 U.S PU R L L N E R E L F RA C ] A R T i C U t A T E P E t F W A C 25 C A R 0 2 5 C 0 N T I N U 0 0 S P R G I O D NRC M M 75)

M, = s ~ e --- x- - .- + si t H ~ NAME

j. S. \\iC _ E2 R R EGi _2 ~ORY CO* * ;: SS :: C \\'

D DATE SUBMITTED RDP T RAiqSC R1PTION SHEET DATE NEEDED PROGRAM 000000000111111111122222222223333333333444444444455555555556666G6666G77777777778 s 12 3 '4 51712 21114 Sj712 2 L P liiG If13 D if,.ll1E 7112 L2 3'4 S 6 L12'Q l'214I5lGf731 I 1 2 3 4 5 6 7e 9 0 1 2 3 4 5 G 78 90 1 CAR D 2 6 F V N F R AC T 1 0 N 1 00 IN E R ELE AS E D F R 0M BL 0 WD O W N T ANK V E N T !( 2 C A R 0 _ 2 7 f E J P F R A C T I O N I O 3 [ N E R Ek C A 5 E D _M A I N C 0 N D E N E V AC 5 Y S T E N [ ] 3 T N 'k. S g A 7 8 9 10 ,11 f.. a,12 14 15 16 17 ' [ 20 ^~ 21 22 24 25 + ' NRC - 53 (175) 9 y e. '. ' ff! 'O 'w ei = W 3 ~ ,4 ' + -{ s , W t.

CARO 1 wA*E NAME UF SHIP HERCHANT SHIP SAPPLE CASE CAWD 7 PodTH THEWHAL PUaFR LEVFL (>EGA=ATTS) 330 CARO 3 putVDL HASS OF l'EI"4WY CDULANT (THOUSANUS LHS) 700 CARD a OEn1FL PR1HAWY SYSTEF LETDuaN WATE.(Gue) 3e. CAWo % CHFLp LETucvN CAT!OH UEM}NERALJZE4 F(Um (GPP) D. CARD 6 GEN NU*uER OF STEAM r.E NE W A TOWS 7 C A R') T TOSTFL TOTAL STEAM FLua (MILLIUN LWS/HW) 1.A CAWD 8

• LI
• ASS OF LIGUID IN EACH STEA* G E'<t u a T op (THiiUSAND LHS) 13, CARD o MLwDwN STEA* fot NE R A THW HLowonaN (THuus Le/Hu) 4n Caen to STEAh GENEWATUN HLnauo=N DEcuN F acitie FnW IUDINE 1.ernt CARO 11 rFCDM CON 0ENSA1E L'E"INFWALIZFo FLua FkAC11tw 0.AS CAgn 17 S9 LOW Houoa HLEED RATF (GPO) 14n.

CARO 13 HuwNAHLE PutSON 0003 FDM WEACTIVITY CONTwoL7 1 IF YES.O IF NO CARD in HEtwto FOW STw1ppING PWIPAWY COULANT. INPUT ost U4 2 Floa CAWD IS TAUT HULouP TIME F0W yF.hov (OAY31 16 CAWD 16 Tad 2 HoLuup f1*E FOW =WYPTON (DAYS) th, CARD IT TA03 FILL TIME OF DECAY TANKS (DAYS) 9.9 CAQ" 1m GAS >ASTE SYST[M PARTICULATE WELEASF. FWACTluN 1 CA' to Aux!LIAWIES CPpf !UUIN6 wtt FRAC 1. P ARTICt'L A T ES Wtl FW AC 01 CAWD 2n CDNvDL kEACTOW CNTni voi tiMt (MILLloN CU.FT.) n.no CAWD pj kEACTUW CNT"Y ATMUSPHEWE CLEANJP RATE (THOUS.CFM) 0.n CAWD 77 WEACTDk CHT"T.PfpIODIC PPWGE = PUWGES/YFAG pn. CARO 73 PEWIDDIC 5'UWGE tootht 4tt FWAC 1 PAWIICULaitS WFL FRAC ,n1 CacD po WEACTUW Lnfuf CouTINoous FUW(.F WATE (CFM) CAWD 7% CUN T I' UQUSPodGE 100!NE ktL FWAC P ARTICUL A TES REL F W AC I A C T !!: It't'I N E p5 L E A SF D F W ' ' *' HLU Dit>N YAkM vtNT 0.ns CAWD p p. FVN CAWD 77 FEJ F s A C T I U*i Itin t hE ptLEASED FWo" HAlW CONU EVAC SYSTt" t. e -Ch

i

4. 3 SAMPLE PROBLEM - INPUT AND OUTPUT The following pages show printouts of the input and output for a sample problem using the NMS-GEFF Code.

4-5

wtWLpaNI $rflp.%sMPLt C a $g Nisu t aw $ftAM 3HFPLY SYSit > W 4Wamt it h b intwwaL FontN LtwfL (athemaffS) I S O. 0 s,o o n LaPALIls Fallow 0 wa 9433 of phi maw r CisOL au f (1HouSahv3 Lns) 2dd.00000 PtwLtNI F ut t mIIM L L abOING bt t t L I S .1/uou PPIMakt S T 3 f t +4 L t 1 Dt)=s. kaft (bpm) $0.0n000 ttlunaN Cat!b4 pt >1ht hat lit M PLUa (BPM) 0.00000 f.UMMt h DF bita= bththalbWh 2.00004 IDial sit aN PLUa (MlltIUN LNS/Hk) 3.eun00 4a33 of iluulu IN taLH bit ak htNtwalow (THousAND LuS) 30.nn000 3it 44 bt ht N 4 IslW MlunhuaN ( IMUti$ lb/MN ) 4.00000 f>aClluN 06 !UUlht Ih ULUal'Uph hul htfukNLD fu StLONUAWT COOLa>1 90000 C U'.0 t h 3 4 I t UtMINikaLidth FLua FWaCIlON 6S400 howou ett to wait (Gru) seu.uon00 b45tnUS aASit 3Y3f t p PahaatitWS ahD INPuf3 fHtNt

• LHNTINUUUS SINIPPihG i+ Puh! FICA 110N TANn PRI

.# C ul tL A N T Sik]> PING haft (bkM) e HDLwwP flht FUM sfhuN (baV3) 16.00000 etoLUUP flPt F ild skvPfuN (Va13) 16.00004 FILL Tiet he DtCAV T&Nn3 (DAYS) 5,50000 MWimahv Ch0LANI LtAn IU aus!LIAWT CMPT (L b/U A T )

160, aualtlaNitS CMP 1 ttaa tuolhi Pawfit!UN taCTOW 00f50 hab aaSTL SYSILM PaWitCULaft NELLASL FNACTION I u3000 aus t L lawlt S L orP ARIMt hi IU0tht WittaSt FNACTION 1.00000 s

PaWIICulait wtLta8E FM4Cf14M 01000 e wtaClow CNIhf VULUME (MILL ION CU.Ff.) 04000 Ch PatuutNLY UF PMIMaMV COOLANI OtbaSS!hG (flat!/yW1 2 PMlmapy 10 StCHNuaMY Ltan Watt (Lu/UAV) 100 6NatfluN IuulNL HypaSSING LUNutNSalt OtMINta;L12tW .55000 luu l %E PawtIT!hN FaCTUW ( b a S /L l uu l t'l IN SitaM GestHa10w 03 FWtuatN07 UF wtalTUw (hfMT Pt NlOUIC PUNGt L Ilkt S/ VW 3 24 ktatlod CNimt 81W10DIC PUHbt luulNt wtLtASE PHACTION .50000 Paw 11(utait wt L t 431 F wa'.f!VN 01000 INtMt l$ NHf 4 Nt&CIUd CNIMI CONI!NUUUS PVWGt 3f t a4 Llan 10 tNGINt LMPI !LnS/"") !?00.00000 twatfline leiplNE wittaStD tWUa blunohmN Taha VENI .05000 twaLilom lublNt vF LL &Stu F Nom MalN LOND tv&C SYSTEM I.00000

"talnaNI 3HjP 3a.p[g (agg wabEHUS kttla$E wait CU=Its FEh vtah PdlMaNT StLohoany has sialvejNb aata vt N T IL al!um Cuota%l CoulaNT Mtv.vuoh magh Coho infat (alC wot t /bwl t mIC Wot t /6= 3autoomh (v.ilhouuS >tatfuw ausillasy E =G l kt vt h1 U666a3 tvat Syst. ""I U. O. 1.Ut+0! 0 O. g. g. 1.Cf*04 MW.el" 5.2 Set.ul

1. 3 39t us v.

D. U. O. U. O. O. O. RN.M5M 2.432t*02 6.62Mt.08 W. g, g, g. G. O. C. e, mm.e5 8.nS3t.03 e.221t 99 c. 3.ut+08 9 O. c. D. c. 1.0E 08 MM*a7 8.S$2t*W2 3.s6 7t un u. C. O. O. u. C. O. o,

• • 982t*Vt

1. d

• 7 t 0 7 v.

O. C. 3.0E+00 p. 9. O. 1.ut*00 MW.a9 1.358t.03 3.ae0E.oe 9 0. O. G. u. C. u. c. ut.13tm 3.svat.0 3 8.oS6t.us 0 2.eg+01 c. U. O. O. O. 2.ug,gt 't.133M 2.2not.ul 6.uu5t.ve G. 1.0E+09 2.0t*0u u. v. O. C. 3.bE*09 ut 131 l.0bsteu0 2.eS 7t u6 2.et+08 1.7E*05 1.5E*02 2.6E*01 o. O. 1.6E*01 5.9f*03 e sa at.135m 3.etit.93 9.320t-09 9 U. O. O. G. O. 9. 6. At.l35 1.299t W2 1.92 8 t 0 7 9 O. O. 2.0E*00 U. 9. 1.0E+0g 3.ct,00 It 137 2.eudt.03 6.2e dt u9 e. o. O. c. O. 6 O. O. atollo I. le f t.02

2. 9 65t.o e 0

O. O. O. u. O. O. c. Iolat hubLt hast3 2.ut+01

1. l li S.u6Ft.ul 1.yalt.uS U.

O. 9.6t.0e 9.PE.03 1.11 03 1.2t.02 6.54 03 3.ce.02 3 353

8. 3 t St.u d 2.333t.u5 u,

o. 2.st.u. I,gt.02 1.et.03 1.7L.u2 9.3E.03 e.3f.02 C.l* 0 l.0E*00 c. 9. O. O. O. l.UE*00 "*3 C. O. 6.6t*03 6.6Ee01 c. O. O. l.3f*02 0.0 apptaw3No In trt tantt IhulC att s wittast is Less inan 1.c L3/vp pow hostr Ga3 aho Caweom-la. LESS Thah u.uuun CI/vd Fuw 100thf. AND LtS$ THaN f( UP total Fow Thltlum

athCnakt snip saastt Cast a l w ellk ht Pah1]CULalt ut((aSt halt (UWif$ Ptb ttak

rwimah, ahta s t % I l t a l lige t uot a r.1 matti 683 S M I C >t iL I / La )

JVSit M BlaClow suultlawy ItalaL mm.5m e.252t.b5 8.et.oe e.at.05 3.6t.05 9.6t.04 Pt.5g 2.ouet.os 3.lt=0e 1.5t=c5 I.2t.05 3.St=0e LO.5n 1.2 tut.95 1.ot.05 8.5t.68 1.2t.08 1.3t.05 Cu.60 e.svet.ue I.et.03

6. e t.05 5.et.uS 3.St.03 SW.he 7.065t =v5 6.et=o5 3.5t.o*

2.6t.ot 7.et.05 hw.90

2. colt.06 I.6t.05 6.et.07 4.ct.07 8.71 05 L S.l ia 5.522t.0 5 9.5t.pe e.71 05 5.96 05 1.01 03 L5 167 3.968t.93 I.6t=0) 6.2t.05 6.5E-05 1.et.03 I

CD

I 4.4 LISTING OF NMS-GEFF CODE 4.

4.1 INTRODUCTION

Calculation of the releases of radioactive materials in gaseous effluents using the Nits-GEFF Code requires the computer program which is available in card deck form from the Effluent Treatment Systems Brancii, USNRC, (301)492-7775. The card deck consists of a CDC FORTRAN source deck and a sample input deck. The prcgiam reads the input from logical unit 50 and writes the output to logical unit 51. 4.4.2 F0P'i.1N PROGRN1 LISTItlG The remainder of this chapter provides the program listing for the NMS-GEFF Code. 4-9

I 4 a t T eC T S OC C C COOC C OeOCC COOOOOOCOOeOOOOO OOOeOCOCOOOCOOC C O C CCC C C = C C C 2 t C C S C O C C C C OOC C COOO COOCOC COC OOCCC OOOO CCCCO OCC CC C C OO C 4h E@ C=%*T# SNE @ C =N*e#eh EPO -Nms# e>CPO= Nee #eb CPO =NM WPON E @ O-NM C C C C==========NNNNNNNNNNem-ememmemesseesee3 3#@########goed O C S* C = t t = C LC OO C O C OO OO OOC OO OOOOOOOOO OOOOC OCO 3OO COD OOC C O O DOO C C C C CC G CC CCOS 0 3C OOCOOG OOC OC OOCCOOCOOOOOOOCOC C OOC OOOCCCOOO C Ie e e en e

• =

C = 0 e e e o a = N w e a @eek 3-w E k m e e CC CC N e= C# e

• I

%N N &w E 9 e 9 0 g 3 = e 9* e ow E wwww a s e = T w 2 e o o

• O e

ON C c e E w: 9 2 w me o e e a w t#me e m m CJ M C O 2 9 =# 4 O*=# S w

• e 2

dy C e o e e e e e w 2 em a s. = e ome w em eO fw FO* N N3 = ehp= e e * = EN e o e e o N mm E Ca & ot = E I= N Om e eame a= a e .O 2

• == **

w e se e OOC C w e C

==

• • w a

w .o 2 C A 2 C 4

1. m=

m e e m 5 0 0 9 e p N ma A E E s wwww w e w

1. 3 F C Nw e Nw C

E e os e NN>r ew e og e e e O SC C M 3 @ N& . m wM : w C 2 e o wu * =J = S* O S e @#F= w Q& W e w es era m T e F c e em h e e e e m e - 2 e

• A JOS* es E

O 2

• e#

5 %==PS m &w = As .,2 E6NC * = w ee-unCMe e wq e o e *%# ee w C Em 3o =

1. ** wn#

D= = %t a 2

• =2=N

1. S @ eSh c em e a

C 7 = 7 se 4e 6 =>CC e emC C C C ep em e w w w g we O PS e# maw - Iwe

• F3>

og e oww wwe e-p 2 e t e=Cc =ce e g a 3 of g C e F 0 eNww Om> ( m. M,N ,J ew ee - N1

• 2

.a#4 e-no e s-e e. c e. C e s. =. e e Fm C &= w ? ** J w e 2 e a Fa mm o e m eO#s=e e# e e C.* e o c2 E c=c e = Zaas weerNr O v-2 7%*e m.E t o 4L e 2 e.% # es20% eg e e e e egeN e er N>t e4 J2. V.

• t%*w=

= -C w Ow me we o e e e o e e e w/ Ca& W O2 So Iw e e awNedakc se a _ e e-m F r= eNw Tem e ... so e- **2 8 m Ww2#e e e # en a>a e en o e so a ee & C T== ? J&

• w.0=&

.e e em

== e O=C @ W OccecreE R e 1 e e: w ZF O ewm ? ok 8 8 e =>UC o% 9 es t ee ea so 4 e = m wee w >4._

• Na2 eM*

e I ww e t 2 e newh wwmww e o e 4 2 wh*C e 2 wm P E %w S

• • C#=
• e m

M*= mE g eS #Owao m eD e C T e*=7 m = 2 e =.Ne e W-E C w 2=4w owh CNgkr ok e% W PJ e eN w e O m. a C C ee-C we s bww ?N mww 0* CeNOdeemCOk O se

• se O.C e e o e e e o e og== m e

eg g g e>N

• e=

e .geNw a. & w 3 eg ee ur w; 1 2 6..Omae N-e e

1. owcz

• = eeaw

. e e a c;wwCe a N= 2. ssN.ede== m# e =m o =m a

==2 w e ae e= o e e e o e e y2= o e sa D w &1 e J >w e-morwerg Op= e e w e m e2

• e 30 2

e

• a*

wT# =&

• ua e

3 a 4C e*eC CC C== e se e>E#m ea w== u> e eC e o aa eao et%e= Oe>N em em e e e e e oce

-==

N2

wew

or=

2 2

NN%* o#*tNew F G e F12 O

• T C=wwwwwp ee
• w emNNe w e m*w Ce
• Ow C Cw w w T
• =E w=e9 0 m

% 8 me or e o ed NC C O eme O e owe e owN

1. ee 3 7 #

wTN*==eTN Tww =*%C 2 = 22C e4CE =eep#Nm o e% s &a sm ocNOe sem C= e 4 *E Ms30w N& lawe a M & ;* = C L w e- *

• w#
• et e24 w*& O=e e

e= e C#C e 8 oceceC o@ ** 2

• seme *
• 2
  • = em si e m 2Cw 20 s6=

wwwCa=2C e et y C e 22.2deE Om

1& 2waww e

2 e2wc ew ow% c%=> 2e e g =* emm e se om o e&&= e m a.e e= e m.* w a e m em m raet t e=p re e e e oc g 4 g en N N > ee e e e e o e e* e==== o *

• wwwe=

we

== me esc =N2

==m w mm u em o ouE nc*e = w 1Lhe2 2 22222 2 *2220 = es eC &was>C eNeerSh %me44=edec w 2& 5 : 2 0 ! ! ! O 0 * *_ C 2 4

• e=CNNew we Tr2N=*===NNwNNwew e==>

wC

=======- E=== E= 0 Fra e ma em amcwwwww wwwww ww w Oww = we 2ePeedemewn# @ t 0 E O== opm sC &&%>2>====>==>>>>>>==** cm C22222220a222 ew awe e a e = = ************weem e we ewee4 I K K ITT II F T TZ rK oF E T awsJ1wewwmwwm Mwwww2w e2 memw M Ee e I t sI I I & F I JI I Ip s we=2=CEw e> >>>>>2322 2 222222222=222 3 Om*

==

M==== e 4 EOeo se se 4 4 eeeOCCOOOOOCOOCCOh 000 212 2wCCCCCCCC =&CCCC* e Ce =wC eE CE= O CCCCCammwewwwesmemw>mam =Ne N =N -Nee - eh t O-Nee # eke p O=we fe@@@fefffeede 444 WW WW 4-10

4-q m 9 4 p

5 a

• .s L [ + 4,

g. .,.,s. r(, 4. m s. y n.y a s* 4 .s.., g s . g ye _ v i

• a a.

e 1 + a s P w P r I r. 'Y CCococooCC.C J C SC C oC C CoC CC ocoOCC OcoceeC C CC oCCCocooccoceCooC C CoCC C C CCC t e ~ C 3 C C C = CC C C C C C C oC C C CC SCC CO 3C CoceC C oC C C ococcC ooc etch t ec-NeoctmceC =Na scens@ o=Newcome@ o=NeopenE @ c= Nee # 4*C pC= 4 st eC th hh h hhh hhmE EE E C CE CE E9 @ @ @ @ #PPP@ CocoC occoc==========NN a 3. c f o C C C =C oco3CO O cco3G Ocococcocooceo====================-- t C T3 C C CcCecCocoC CoococcoccooococcooooooooCocococcococcccCO W X a e e2 = w o E m = W w2= e 4 E C e a m' m s N = 2 w Me = h mem e e 2 y = g =

== = = mMae# N = Z= 2 e'=et O 2 a t w 2erk e = 2 = c0 N = L e.a P e M -e e w eso >C w w J 2==* e e a w = >wCwm S. % = 1 T = 2 =& r= .e a a m meaa ,t tw2 e e e=** em = O2aw = me ; 2 O21= oC e w > a:

• • x C

c I A a =w e > Rw1 m => C Fe w Immm 2 e.

he

4 .COm N

• w

= w 4>24= = I g? _ * *2 1 a= e =C C e

== w w e c

==

• C=

2 tw= = 1 wm M w . so a Mut. C2 O

• I
• 2m 2

ww>

• 9 m

w= 2

O2=

F e 2 w Cw=O+= w = Wa r wC -weC

*C

=2 = 4

=== m6 de 9 2>=. N m2 o = = = = = => a h i i =FC=2* F *1 = eC a C w 20 = wm u 2R

==mv ( % 7

• • =

2 -ww w 2 C w 4 m: C e 3= 0 = C >& deem O + A" eg : 4 3r Ow2I e m Z> = C+ =C 2 e2%w 2 Cs w 7 2 cmw: = 4 % e O2 M 20 w C.% w = .J = w. = a

== =m.i 2 w=C 1 2 = = 2 0 me w =-C a w. 2 e = = e zh e = w .oe n > = 22 w + m => acueee C a f

== ==? m2 e w 1 00T w ww F 204 m C w a & 2 > = C 2 de

• • ema.

T = re%% Cw e 3

==2&2C m = >C e %w = w&&weg =

• 2% #

J e

• =

= w m2 7 2& Os w = e me w a e .jaeA C. w &O e = 2 2! 2

• J e

=&Im e. e

• 1 m w

m m w new Cw m = 2 e em>r = -w a m ee 4 e a a

• w we m

wrCw2e = 2M C===== 2 x i a >. e Fr 1

• • C.w

e.. e

• w a

w w l wa w w ze w e o am C w

==.

ec 7 > w r >rra = wem Zw es = emas== w e amm* CE e w c e r#e w am u w 2; Fw a wC=

• a q w& w w w l

=Ce

• 6

? M =Cv h*wo 2m e w se >. z

• ae e=c

=

• a e

-= = 2 *Ta w C 1 2Cp we==ew& s tw a e2 2 C e Z A 0 1 es w 2=2 e 4 *wwe K !=

==Mm e =4 i e ww w m>&2em me = M* g ..ay

=

c== ew e mum =m s : 7mt== w. =Cw wave w >>te ZE w Or

• C e

ICeest owea e > em : 2CZ= 2 = Om = + 2 e e neww > ww ow= s C eC#2= 2 1 -w==w meze o u N

• C=

e.a et em> 2 memete

==>

== .eeg = w e m 2e = see r me= 6 a wree ser c em m 2 met w m eC% we r 5 0 e=W F me==C O =>1 F E. N. o =: e = =e a o. N = = 3 m >%=c => 04 2& e Is w C 00 Om e 2 2mC **> 2 w wwCIAE== =o3 s mmNst O 2mwrw eew2 & = = ; ee

===ew w a e. e e o c e e O

• C

=C 3 2e S Im aw = me G A t t e. e e* e= > =* w= = T Cm C2 2 = wI m =>we C se wwCC v m o meu

• CE 2w*Cw= e2 m m.e e

a 2 2 m = 2 a=1ww1 2 O ec 1 2.== men e e=L es

• Im om

=Ia=2 Cm . 2e ra: t CX w =. mo w =. C e 4 > > E..z e W e.e.m e.e g m e r..e 2e SE=

=

wt>

• • w

= w2 e d.e c n.- --m.e e ea .. e. C m m

e. g.

= a

w. = o e= a =. c =.

a. m a amm >

==m ea=Neae

==e e a .== =>m

== e.. a m om emen om au

===mmum leem a mm um am#m e em e=4=Ne>No@ # **#Ceme

• • W ere 4

em samme@ CN4% Zeeder===E eed des 4 eN=am e = = =.= >w =w a r. c e w ow e w- -(

e

==N-4 -g=

===w-e-u -=====wwwe--- --e-Chwcww

ew-w&Oww owe

wwwwQw wwww wwww=w e w =Fm => 2cOww cw awcw ow = 2 ccurewcrC#C#we

m-N=ww-==>=wm>>>s>mmm=>>=www e-mw e see e Zweeme ag e emeeeeeenweeermeee eeg weeeCem eeCwwwhwwwewwwow 3 C I et e t r>Em 7 I T T T ET T E 12

• EE
• EE Z TmI I
• 1E F=FwT TWCwww=C=e*C=4*

I = = = e = w = * = m. m ame.e2mosmammatema2=2w3M**mCOCwO*;Onwwm23*2 R e2NR w>23 om 32 22mmeg

• w w.2 0 0 m :ms.kwawsmawwwwwwwwwwwmCC Os 000 CCO

CC OC=C

=C &2w2 m !=0 N e e e ~ aces &ommoea kom Neer 4> eo== N e#= a-E 2 33 5 : 222& E ww wwwwwewewa= a*- w = =N = = = = = o e c o ooos===NNN Nem emme me mee e se o e o @799@@@@@@@ O 4 4_11 . l - ' k[" .[.,i ------'k [ga '-] ',- '-,,., y' S a, -'w'7"* 4# + - f a , p. g ,3

9 T S TC C C C EOC C C C C CC 9CCCOOOOOOCPOOC CC C OOOOO O '1 O C CC C SOCOOOOOOOO 2 2 C C O C C C C C C 3 CC OC OCC C C OcOC C C CC OO -aC C C CC C C OC CCC OOC CC CCC OCC OCC %*W # dh E P

==N*T/ d>K@ C-NeW#ehCPC=P A W# eh t 9 0 = N #+ @ #dh E @ @ =N*W#4her %%%%%%%%**********4 5 SeeW WWW W##8

1. 1. 1. 1. 1. 1. 1. .4444@---m m --=======m--mm S eeddkkhh hhhkkk

- --======----- C G C C OOS G OOOC CC - -------m---mmm-- -m--=wm COOCOC OOCOOOOccccOOCCOOSCOOCOOOOO9COCOCOOOOO d m e 3 m C 9 m. E o m b e 1 e O = C e C 4 -Mg e O CC-C C TE C% C -m 1 -m-10 3 2 4 m = C e a e GE SS C s-i .m m #.

=

=Ia b I CWC -a F h00-e C =Ce O*Oe a e 3. m* I E m E awa. O I e &=>Z T mh E FueC=. e e C wk E R e e %= ONC*C E w; 3 wWOC* S OsOa 2 malm 1 > 00e R e en ewe

• 6 C ou ed om os
• CaCJL Z

s emosO O e agC. C =*E oO

• L
• C w

= ** => e m ;

• O e4 oC

=O 1 COe % m eC oC3 J 2 a m G C 202 22 O2 C2 0$CRCG 71 2=C C2: 2 1 e d

• * *=

=Ia M a MM O4 02 O e E R C40J 03 OG O A C e a J O20=a: 5 J-Naa C 0.ex 0 s O.e 0.mt O.e. e==e1 .E 1 t a-e

• =

E e-2 =m e owe aa ea=aO aOE OA 3

• ae 4

Ae & .e

• = Cw W

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= REFERENCES 1. NUREG-0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1976. 2. United States Department of Conmerce, Maritime Administration, Preliminary Safety Analysis Report for Competitive Nuclear Merchant Ship Progran, P-426, April 1974. 3. Westinghouse Electr'.c Corporation, Safety Analysis Study for Nuclear Propulsion System for Large High-Speed fierchant Ships, P-433, July 1968. 4. American Nuclear Society ( ANS) 18.1 Working Group, American National Standards Source Term Specifications ANSI N237-1976, " Radioactive . Materials in Principal Fluid Streams of Light-Water-Cooled Nuclear Power Plants," flay 1976. S'. Regulatory Guide 1.140, " Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust Systen Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," tiarch 1978. 6. NUREG-0002, Vol. 3, " Final Generic Environmental Statement on the Use of Recycle Plutonium in lif xed 0xide Fuel for Light Water Reactors," U.S. Nuclear Regulatory Commission, August 1976. e R-1

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