ML20070L333

Rev 1 to 6610-PLN-4200.02, TMI Emergency Dose Calculation Manual

ML20070L333
Person / Time
Site:	Crane
Issue date:	03/15/1991
From:	Shawn Williams GENERAL PUBLIC UTILITIES CORP.
To:
References
6610-PLN-4200.0, NUDOCS 9103190407
Download: ML20070L333 (109)

Text

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THI-1 EMERGENCY DOSE CALCULATION MANUAL (EDCM)

INSTRUCTION MEMO Document Control Desk office Nuc. Reactor Regulation U.S. Nuclear Regulatory Commision Washington, D.C.

20555 2 INFO RETURN TO: Betty Nash Procedure 01stribution Control Unit 2 Admin. Bldg.

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Please update your Copy of the TMI-1 Emergency Dose Calculation Manual (EOCH) as j

instructed below.

Also, please sign the acknowledgement at the bottom of this memo and return to Betty Nash as-shown above, REMOVE INSERT 6,6 /O ~T1 N ~ Alc'A00 02 L G /n-9L Al-42 an. n2 Qu&, n

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TMI Emergency Dose Calculation Manual (IDcM) 1 Applicability / Scope The EDCM is applicable to all qualified Roepns'ible Office Radiological Assessment Coordinators. This manual provides the methods used to perform dose projections during emergencise 6610 This document is within QA plan scope X

Yes No Effective Dake Safety Reviews Required X

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e Nuclear Tm Radiologieel contrnis Department 6610-PLN-4200.02 Title Revisicn No.

TMI Rmeroency Dose Calculetion Manu al (EDCM) 1 Table of contents Section g

1.0 PURPOSE 4.0 2.0 APPLICA3ILITY/ SCOPE 5.0 3.0 DEFINITIONS 6.0 4.0 PRER2QUISITES 15.0 5.0 PROCEDURE lti. 0 i

5.1 Source Term calculationo 38.0 5.2 selection of Release Pathways and Charactorintics 14.0 5.3 calculatito of NRC Damage Class and Isotopic Percentages 23.0 i

5.4 Radiation Monitoring System (RMS) Scurce Term Calculation 35 0

/"'A 5.5 Post Accident Samples fource Torm Calculation 40.0 5.6 Contingency calculations Source Term Generation 42.0

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I 5.7 Decay Scheme Calculation 41 0 5.8 Noble Gas to Iodito Ratio Calculations 49.0 5.9 Effluent Release Flow Rates 52.0 l

5.10 Two-Phase Steam Flow Determination 65.0

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I 5.11 Source Term Filtration 66.0 d

5.12 Meteorology Inquiry 68.0 i

5.13 Dispersion Model 69.0 j

l 5.14 offsite Air Sample Analysis 75.0 1

l 5.15 Liquid Release Calculation 70,0 5.16 Protective Action Recommendation Logic 84.0 5.17 Dose Prcjection Model overview THI-l 90.0 5.18 TMI-2 Source Torm Calculation 91.0

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6.0 PESPONSIBI' ITIES 101.0

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___TMI Emeroeney Dese calculation Manual (EDCM) 0 1.0 (URPOSE The purpose of this manual is to provide a document that describes the assumptions and methodolory used in the current TM1-1 and TMI-2 Radiological Assessment coordinator (kAC) programs. This includes ca'.~ulating projected on-site and of f-site doses f rom releases of radioactive material to the environment in accident conditions upon implementation of the Beergency Plan. As such, this document deterLbes nethods of projecting off-site doses during emergencies or for training purposes. Indications of releases may result from Radiation Monitoring System (RMS) readings, on-site or of f-site sample terul's, or contingency calculations, if RMS and sample results are not available. Tneke dose projections are performed by computer using the current version of the 'th1-1 or TM1-2 RAC programs. The Radiological Assessment Coordinator is 19sponuible for implementing the dose projoe?. ion process for THI-1 and THI-2.

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t TMt Emergency Dopo calculation,, Manual (EDCW) 0 l

2.0 APPLICABILITY / SCOPE i

The EDCH is applicable to all qualified TMI Radiological Assessment Cooruinator I

I (R8.0) personnel invo1**d in the projection of on-site and off-site doses during an emergency. This manual provides the nethode.used in performance of dose projectione during energency situations where radioactive material-has been or is j

predicted to be released to the environment.

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i 3.0 DrrINITIONS 3.1 BUILDING WAKE EFFECTS - When an atmospheric release occurs at, near, or below the top of a building (or any structure) the dispersion of the release is affected by the wake effect of the building. Air flow over and around the structure from the prevailing wind tends to drive the release i

down to the ground on the downwind side of the structure. This has two i

effects:

it increases on-site concentrations dramatically, while slightly l

reducing concentrations downwind for a short distance. Far downwind concentrations are affected very little by building wake, muill'ing wake effects are most noticeable for ground level or low flow stack releasee

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euch as the rondenser off-gas exhaust. Normal plant ventilition usually j

has a high enough flow that building wake does not affect the plane eignificantly. Building wake is accounted for as part of the split wake release modeling.

3.2

" CHI over V" (X/Q) - is the dispersion of a gaseous release in the l

environnant calculated by the split wake dispetsion model. Normal units of X/O are sec/ cubic meter. X/Q is usea to determine environmental atmospheric concentrations by multiplying the source term represented by Q.

Thus dispersion, X/Q (sec/ cubic meter) times source term, Q (pci/seo) yields environmental conenntration X (pci/ cubic meter). X/Q is a function of many parameters including wind speed, delta T (change in temperature s

with height), release point height, building site, and release velocity, among others. The release model takes all these into acccunt when

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calculating atmospheric dispersion.

I 3.3 CONTAINMENT h!R CONTROL ENVELOPE (CACE) - This facility provides a l

contairment outside of the TMI-2 equipment hatch. This containment uses-two AMS-3 radiation monitors to monitor releases, at 2000 CrH, from thio facility when the facility is in use.

3.4 CONTAINMENT ATHOSPRERIC POST-ACCIDENT SAMPLING SYSTEN (CATPASS) - Post i

accident sampling system capable of providing sample (s) following an l

l accident condition, coincident with a blackout, with limited personnel exposure. The campling system, located in a post-accident accessible area, 4

stovides the capability for obtaining samplee of the Reactor Building steasphere, within one hour after the decision has been made to acquire the onople(s). The samples (s) are then used for radiological and hydrogen

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• aalysis. These results will provide an indication of the extent of core damage and provide good data for the Reactor Building source tGrm if a Reactor Building release is possible.

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TH! Emergency Dose calculation Manual,1EDCN) 1 3.6 CONTINGENCY C;LCULATION - A source term calculation per!ntmed in the absence of sufficient effluent radiation monitoring system readings or post accident senple data. It l' a mathematical calculation based upon the most repreaentative physical meu l of actual accident plant conditions.

3.6 CORE DAMAGE - (Note: This delinition to be used in lieu of -

defective / failed fuel.) A set of core classifications used to address the requirements of the NRC NUREG 0737 Criterion 2(a) upon implementation of the Emergency Plan. Based upon RCS pressure and incore thetisocoup14 readings, an assessment is made of the degree of cladding failure, fuel overheat, and fuel melt.

t 3.7 DEFICTIVE FUEL / FAILED TUEL - See definition of cero damage.

3.8 DOSE AATE CONVERSION FACTOR (DRCF) - A parameter calculated by the letho18 and models of internal dosimetry, which indicates the committe4 dose equivalent (to the whole body or an organ) per unit activity ?nhaled or irvissted. Thia parameter is specific to the radionuclide and the done pathway. Dose conversion factors are connonly tabulated in units of 3

mrom/hr per curio /m inhaled or ingested.

3.9 ELEVATED RELEASE - An airborne ef fluent plume which is well above ary ouilding wake effects so as to be essentially unentrained is termed an

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elevated release. The source of the plume may be elevated either b; virtue of the physical height of the source above the ground altaation and buildings or by a combination of the physical height and the jet plune rise, semi infinite modeling of elevated releases generally will not produce any rignificant ground level concentrations within the first 14w hundred gards of the source. Semi infinite modeling of elevated releas*E generally have less dose consequence to the public due to the greater downwind distance to the ground concentration maximum compared to grot.nd releases. Elevated releases as used in the EDCH actually means "not at ground" in the split wake plume model. Other definitions of "e)*=at9d" with respect to plumes abound in literature.

3.10 EMER32NCY ACTION LEVEL (EAL) - Predetermined conditions or values, including radiation dose rates; specific levels of airborne wat #;,orse; or surface-depostted contamination; events such as natural disastere or fires; or specific instrument indicators which, when reached or exceeded, require implemuntation of the Emergency Plan.

3.11 EMERGENCY DIRECTOR (ED) - Designated on wite individual having the rasponsibility and authority to implement the Emergency Plan, and who will coordinate ef forts to limit consequences of, and bring under control, the emergency.

3.12 EMERGENCY DOSE CALCULATION MANUAL (EDCM) - This controlled dose calculation manual is the documentation describing the content and calculational methods of the Radiological Assessment Coordinator (RAC) program.

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TMI Emeroency Dees calculation Manual (20cM) 1 I

3.13 EMERGFNCY OPERATIONS FACILITY (EOF) - The Emergency Operations Facilities serve as the primary locations for management of the Corporation's overall emergency response. These facilities are equipped for and staffed by the Emergency Support Organisation to coordinate emergency response with of f-site support agencies and to assess the environmental impact of th; emergency. The EOF participates in accident assessment and transmits appropriate data and recommended protective actions to Fedaral, 8 tate and Local agencies.

3.14 EMERGENCY PLANNING 20NE (EPf) - There are two Emergency Planning fonse.

The first is an area, approximately 10 miles in radius around the site, for which emergency planning consideration of the plume exposure pathway has been given in order to assure that prompt and effective actions can be taken to protect the public and property in the event of'an accident. This is called the Plunie Expceure Pathway EPf.

The second is &a area approximately 50 miles in radius around the site, for which emergency planning consideration of the ingestion exposure pathway has been given.

This is called the Ingestion Exposure Pathway EPE.

3 15 ENTRAI! MENT = When a release is treated as a wake split release an entrainment factor is applied to specify how much of the release is to be considered elevated and how much is to be considered a ground release.

Entrainnent factor is related to the building wake offset. The entrainment factor is computed on a case by case basis and is dependent on both the g

j stack exit velocity and the wind speed. At low wind speeds and high exit velocities, building effects are lowest and the entrainement factor selects for elevated release. At high wind speeds and/or low exit velocities the building effect is hignost and the entrainment factor results in a ground level release. Intermediate conditions cause entrainment factors which will split the release between ground and elevated. The general forin for the application of the entrainment factor (Ef) ist E/Q(splitwake)=E/Q(ground)*Ef + E/Q(elevated)*(1-Ef).

As can be seen from the formula, when tho entrainment factor is one, the release is entirely ground and when the entrainment factor is aero, the release is entirely elevated. When 0 < Ef < 1 then the release is split.

3.16 EPICOR II - Radioactive Liquid Waste Processing Facility located on the east side of TWI-2.

This f acility is used to process TMI-2 radioactive i

waste. An Eberline PING radiation monitor is located on the ventilation eystem of this facility. The ventilation system everage flow rate is 9000 CFM.

3.17 EXCLUSION AREA (EA) - As defined in 10CFR100.3; "that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property f rom the area".

At TMI this is an area with a 2000 ft. radius from the point equidistant between the centers of the THI-1 and TMI-2 reactor buildings.

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THI Emeroency Dose caleu_lation Manual (EDCM) 1 3.18 EXIT VELOCITY AND PLUME RISE - Atmospheric dispersion and ground concentrations are in part dependent on release height. Higher release heights will cause lower maximum concentrations at gtound and will cause that maximum to occur further downwind than would a lower release height.

The effective height of a stack is not only dependent on its physical height, but also on whether the plume rises or not. At high linear flow rates (exit velocity), the release plume behaves much like a geyser and rises in a jet flow above the stack. The height to which the jet flow rises becomes the offective stack height.

3.19 FINITE PLUME HODEL - Atmospheric dispersion and dose assessment model which l

is based on the assumption that the horisontal and vertical dim 6nsions of an offluent plume are not necessarily large compared to the distance that gamma rays can travel in air. It is more realistic than the semi-infinite plume model because it considers the finite dimensions of the plume, the radiation build-up f actor, and the air attenuation of the ganana rays coming from the c1 cud. This model can estimate the dose to a receptor who is not submerged in the radioactive cloud.

It is particularly useful in evaluating dosas from an elevated plume or when the recepter is n6ar the offluent source.

3.20 FUEL MANDLING BUILDING ENGINEERED SAFETY FEATUkE VENTILATION SYSTEM - The ruel Handling Building ESF Ventilation Systee, is being added to THI-1 in accordance with a ccamitment M the NRC.

This connitment has been included in the NRC THI-1 restart report. The Fuel Handling Building E8F Ventilation system is installed to contain, confine, control, mitigate, monitor and record radiation release resulting from a TM;-1 postulated spent fuel accident in the Fuel Handling Building as descr Wed in FSAR, Section 14.2.2.1, Update 1, 7/82. Normal operation of the Fuel Handling Building ESF Ventilation System will be during THI-1 spent fuel movemente in the Fuel Handling Building. The system design shall in11ude adequate air filtration and exhaust capacity to ensure that no unccntrolled radihactive release to atmosphere occurs. The system shall include strauent radiation monitoring capa,bility.

3.21 GAUSSIAN PLUME EQUATION - An equation which takes input parameters of plume height, eigen-Y, sigma-E, and wind speed, which explicitly calculates the straight line Gaussian Plume Dispersion. The caussian Plume equation actually averages short t6rm variations to produce a mean effective plume, so short term measurements of the plume may not be duplicated by the caussian Plume Model.

3.22 GPOUND RELEASE - An airborne offluent plume which contacts the ground essentially at the point of release either from a source actually located at the ground elevation or from a source well above the ground elevation which has significant building wake effects to cause the plume to be entrained in the wake and driven to the ground elevation is termed a ground level release. Ground level releases are treated differently than elevated releases in that the X/Q calculation results in significantly higher concentrations at the ground elevation near the release point. Ground releases also have generally lower X/Qs all the way downwind.

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TMI Emeroeney Dese Calculation Manual (EDCM) 1 3.23 KYDROGEN PURGE SYSTEM - Post-accident containmer.: purge system is designed to maintain the hydrogen concentration of the post-accident containment atmosphere below the lower flammability limit. The system does this by introducing outside air into the Reactor Suilding, which allows the displaced containment atmosphere to be dischargaJ in a controlled manner into the normsl Reactor Building exhaust duct. In the flow path three release rates exist which can be additive to give flow from 5 to 1250 CFM.

3.24 INTERIM SOLID WASTE STAGING FACILITY (ISWSF/ PAIL *T SKED) and the THI-2 PAINT SNED OR RADWASTE MATERIAL STORAGE FACILITY (RMSF) - These facilities have no ventilation system or radiation monitor, but have the potential to release radioactive material to the environment.

3.25 LOW POPULATION EONE (LPE) - As enfined in 10CTR100.3 "the area Lmmediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measuaes could be taken in their behalf in the event of a serious accident.

3.26 METEOROLOGICAL INFORMATION AND DOSE ASSESSMENT SYSTEM (MIDAS) - This is the acronym for the computer program that can be used by the Environmental Asessement Command Center (EACC) to project off-site doses for routine

-s effluents and releases during emergencies. The MIDAS program runs on a

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main frame computer. Some features of MIDAS that are not in the RAC

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program are ingestion pathway doces, liquid and gas population doses, dose projections at any desired point of interest, and sector dose integration.

3.27 NRC DAMAGE CLASS - A method of estLmatirig the extent of core damage per NUREG-0737 Criterion 2 (a) under accident conditions requiring Lmplementation of the Emergency Plan. The initial estLmate of the degree of reactor core damage is derived from the calculated radionuclide concentrations that are measure?. on water samples taken from the water inventory of the primary system. The assessment is performed utilizing a matrix that consists of ten (10) possible damage categories ranging from "no damage" to " major clad damage plus fuel melting".

I 3.28 OFF-CENTERLINE DOSE CALCULATIONS - Dose calculations that Are calculated at i

various distancos from tha calet. lated plume conter11ne (0, 50,100,150, 250, and 500 meters). These calculations are performed at 28 distances from the plant.

3.29 OFFSITE AIR SAMPLE ANALYSIS SYSTEM - An air sampling and analysis system l

specifically designed for iodine air sampling and thyroid dose assessment.

The system consists of an air puup unit which draws air through a canister containing a highly efficient iodine adsorbing material and a Geiger Mueller S*nctor for canister evaluation.

3.30 PARTITION FACTOR - (Condenser), see NUREG-0017 l

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3.31 POST-ACCIDENT SAKPLING $YSTAM (PASS) - System used for acquiring a pressurized liquid sampin of the RCS during emergency conditions. The post-accident reactor coolant sampling system provides a means of obtaining a representative sample of reactor coolant, inclucing dissolved gases, reactor coolant bleed tank contents and reactor containment sump contents, within one hour ef ter th4 decision to acquire the sample, without excessive operator exposure or compromtse of interfacing safety-related systems.

3.32 PLANT SNUTDOWN RADICIODINE SPIKING -

Radiciodine spiking occurs at < 15% reactor power, and is caused by the fuel gap activity being washed out of the fuel assemblies at low temperatures and pressures.

3.33 PROTECTIVE ACTION 001DE (PAG) - Projected radiological dose or dose commitment values to individuals of the general population and to emergency

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workers that warrant protective action before or after a release.of radioactive material. Protective actions would be warranted provided the reduction in individual dose expected to be achieved by carrying out the protective action is not offset by excessive risks to individual safety in taking the protective action. The protective action guide does not include the dose that has unavoidtbly occurred prior to the assessment.

3.34 PROTECTIVE ACTION RECOMMENDATION (PAR) - Those actions taken during or af ter an emergency situation that are intended to minimise or eliminate the

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hazard to the health and safety of the general public and/or on-site personnel.

3.35 RADIATION INSTRUMENT SHOP - Facility used to repair / calibrate / maintain instrumentation used by Radiological and Environmental Controls Departments. Tnis f acility includes calibration sources that may possibly be released in a worst case accident (e.g., fire).

'.'his f acility has a ventilation system, (rated at 4000 CrM), but no installed radiation monitor, or filtration.

3.36 RADIATION MONITORING SYSTEM (RMS) - Th1 RMS detects, indicates, e

annunciates, and records the racketion level at sels:ted locations inside and outside the plant to verify compliance with applicable Code of Federal Regulations (CFR) limits. The RMS consists of the following subsystems:

area monitoring, atmospheric monitoring, and liquid monitoring.

3.37 RADICIODINE PLATECUT - Iodines are very chemically reactive, being members of the halogen family. As auch, iodines have a very high probability of reacting with almost any othat material they come in contact with.

Radiciodine plateout is a generic term for the mechanisms by which radioactive iodines are removed from a waste stream by contact with materials not specifically designed or engineered for radioiodine removal.

Examples of potential radiciodine plateout reactions are the removal of iodine from gaseous wastes by adsorption onto interior surfaces of ductwork and piping and on any exposed surf aces of the roca or building originating the release.

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__ TMI Emergency Dose Calculation stanual (EDCM) 1 3.3B RADICIODINE PROCESSOR STATIONS (MAP-5) - System used for acquiring particulate and lodine samples from the Reactor Building Exhaust, Auxiliary and Puol Handling Building Exhaust or the condenser off-gas Exhaust during emergency conditions. The stations are controlled by solenoit valves which activate on high alarm indications on the low gas channels of the effluent stream. Flow is actuated through (3) parallel filter cartridges per station. The sampling times are adjustable on sach local control panel.

The filter cartridges tnet be removed manual'y for analysis.

3.39 RAD 10 LOGICAL ASSESSMENT COORDINATOR (RAC) - The RAC is responsible for all on-site radiological assessment activities. Initielly, the RAC is responsible for directing the on-site and off-site survey teams. The RAC is relieved of off-site radiologicel monitoring responsibilities by the Environmental Assessment Coordinator. The RAC performa dose projections, based upon source term estimates and provides information to the EAC.

Initially the Group Radiological Control Supervisor assumes the role of the RAC until relieved by the Initial Response Emergency organisation RAC, and RASE.

3.40 RADIOLOGICAL ASSESSMENT SUPPORT ENGINEER (RASE) - Individuals assigned to assist the RAC in performing dose calculations, source term calculations, and ovarall assessment and control of radiological hasards. Normally one RAC and one RASE are on duty at all times.

O 3.41 REACTOR COOLANT SYSTEM (RC8) - This system contains the necessary piping and components to provide sufficient water flow to cool the reactor. This system provides fer the transfer of thermal energy from the reactor core to the once through steam generators (OTso) to make steam, acts as a moderator for thermal fission, and provides a boundary to separate fission products from the atmosphere.

3.42 RELEASE DURATION - Release duration refers to the time interval during which radionuclides are released from the nuclear facility. Releases may be monitored, unmonitored, actual, or projected. The time interval used to estimate a release of unknown duration should reflect best eettmates of the plant technical staff. In the absence of other information, use two hours as the expected release duration. For purposen of determining whether to take a protective action on the basis of projected dose from an airborne plume, the projected cose should net include the dose that has already been received prior to the timo ths dose projection is done.

3.43 RELEASE RATE - This term refers to the rate at which radionuclides are relateed to the environment. Normally, it will be expressed in curies per second (CL/sec) or microcuries per second (pci/sec).

3.44 RESPITtATOR AND LAUNDRY MAIFTENANCE FACILITY (RLM) - This facility is used to process clean and maintain laundry and respiratore for THI-1 and THI-2.

This facility's 900 CFM ventilation system is monitored using a EbeOline PING radiation monitor. The RLM has KEPA filters installed in the ventilation system.

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Title Revision No.

TMI Kmargency Dose calculation Manual (EDcM) 1 3.45 RMS RESPONSE FACTOR - Parameter which is used to convert RMS monitor count rates to total microcurie per cubic centimeter of the assumed or measured radionuclide spectrum passirg by the monitor. This is different from a meter calibration factor which does the same thing for a single calibration nuclide. These factors are adjusted for changes in mixture decay.

3.46 SEMI-INFINITE PLUME MODEL - Dose assessment model which is based on the assumption that the dimensions of an ef fluent plume are large compared to the distance that gamma rays can travel in air. If the plume dimensions are larger than the gamma ray range, then the radius of the plume might just as well be infinite since radiation emitted from beyond a certain distance will not reach the receptor. The ground is considered to be an infinitely large flat plate and the receptor is nasumed to be standing at the center of a hemispherical cloud of infinite radius. 'The radioactive cloud is limited to the space above the ground plane. This is the origin of the name OEMI-INFINITE PLUME. The noble gas MPC's were developed on the basis of the semi-infinite plume model.

3.47 SIGMA-Y AND SICMA-E - Parameters of the Gaussian diffusion equation which determine horizontal and vertical diffusion. Sigma-Y and Sigma-E varies by stability class and distance from release point.

3.48 SOURCE TERM - A source term is the activity of an actual release or the

,r-Activity available for release. The common units for the source term are

,\\.

curies, curies /sec, or multiples thereof (e.g., microc. ries). The term

" Source Term" derives from the equations involved in doing dose calculations, since the equations contain many terms (a torc being mathematical nomenclature for a portion of an equation), tho " Source Term" is that portion of the equation which addressoa the activity released.

Although the term "3ource Term" is used loosely to mean almost any activity for airborne, liquids, and even dose rate cal alations in plant, strictly speaking " source Term" applies only to radioactive material actually released.

3.49 SPLIT WAKE RELEASE - Airborne releases, for purposes of assessing of f-site dispersion, must address the elevation of the release since wind speed changes with height, buildings affect dispersion for low releases and even wind direction can be different. Many release points are actually at a height where, given dif ferent conditions of release flow rete and meteorology, could either be most accurately described as ground or elevated releases, or some mixture between the two.

The purpose of treating a release as a split wake release ik to address this problem.

When a release paint is set up to be treated as a split wake release, the atmospheric dispersion is calculated based on a mixture of elevated and ground releases. Thus at high release flow rates the release may appear to be entirely an elevated release and at very low flow rates it may appear to be entirely a ground level release. In intermedi4te conditions, the model will " split" the release between ground and elevated as appropriate, so that a release might be 25% ground and 75% elevated from the same release point.

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xx, Radiologie41 Controls Department 6610-PLN-4200.02 Title Revision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 3.50 s! ABILITY class a Diesersion of an affluent plume '.n the atmosphere is a function of the amount of mixing occurring bet ~.+en the plume and the atmosphere around the plume. The amnunt of mAxing is related to what is referred to as the stability of the atmosphere. conditior4 which create good mixing are unstable and conditions which create pooter mixing are stable. Pasquill stability class is a breakdown of the reistive atmospheric stability into seven groups, denoted as A through 0, from most unstable to most stable. In the pasquill stability class system, acability is related to the relative change in temperature with height, delta T.

The more negative the change in temperature with increasing height, the more unstable the atmosphere. Tne RAC program uses sensors on the Meteorological tower at 33 feet and 150 feet to determine the delta T.

Once the delta T is determined, a stability class is selected based on the delta 7 and the atmospheric diegersion (X/Q) is calculated based on the selected stability class.

3.51 STATION VENT HPR 219 - This radiation monitor and release pathway is the main release point for THI-2.

All ventilation from the Reactor Building, Auxiliary building, and ruel Handling Building are routed to the station vent release point, at an avarage flow rate of 120,000 to 130,000 crH. The radiation monitor HPR"219 is the originally installed Victoreen Radiation Monitor using a particulate, iodine, and gaseous sampling and monitoring system, similar to the THI-1 radiation monitor RM-A2 with a moving

(

particulate filter.

3.52 STATION VENT HPR-219A - Eberline PINr unit installed in THI-2 to monitor the THI-2 Station Vent Stack. The r:,d out unit is located in the TMI-2 Turbine Building. This unit is used in conjunction with HPR-219 to monitor the main THI-2 release pathway.

3.53 TERRAIN FAC70R - The terrain factor is the terrain height in meters above plant grade. Terrain f setor varies with sector and dir' ince f rom the release point.

l 3.54 TWO PRASE RELEASE - Liquid and steam release from the main steam safety relief valves, rolicwino discharge to the environment the steam fraction is calculated assuming snare is no change in system entropy and that the 0780 wide range level instrument is indicating that the valve inlet fluid condition is either pure liquid or steam (greater than 600 inches as indicated on the PCL Panel, PI-950A and PI-962A).

3.55 WASTE RANDLING AND PACKAGING FACILITY (WHPF) - This facility is used to handle and package radioactive waste mainly f om THI-2.

This facility's ventilation is monitored by a PING /AMS-3 radiation monitor, and runs at 7100 crM.

3.56 WIND SPEED ADJUSTMENTS - Since wind speed varine with height and the wind speed sensors are not at the release height, an adjustment is made to extrapolate the measured wind speed to the wind speed at the release height. The adjustment amount is dependent on the stability class.

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Radiological controls Department 6610-PLN-4200.02 Title Revision No.

TN! Emergency Dose Calculation Manual (EDCM) 0 4.0 PREREQUIstTFS 4.1 The following are the prerequisites for performance of THI projected doses using the methods in the EDCM, and the current TMI-1 or THI-2 RAC Program.

4.1.1 The Emergency Plan is being implemented.

4.1.2 The RAC station is manned and functional.

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$UClSer rMr Radiological controls Der >artment 6610-PLN-4200.02

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Title aavision No.

THI Emeroency Dose calculaticn Manual (EDCN) 1 5.0 PROCEDURE This section of the EDCM is divided into the processes that are contained in the i

RAC computer programs for THI-1 and TMI-2.

Listed below is a table of contsnts for the procedure section of the EDCMs 5.1 THI-1 Source Term Calculations 5.2 Selection of Release Pathways and CLaracteristice 5.3 Calculation of NRC Damaga Class and Isotopic Percentages e

5.4 Radiation Monitoring System (RMS) Source Term Calculation

{

5.5 Post Accident Samples Source Term Calculation 5.6 contingency Calculations Source Term Generation 5.7 Decay Scheme calculation i

5.8 Noble cas to Iodine Ratio Calculations i

p 5.9 Effluent Release Flow Rates I

5.10 Two-Phase Steam Flow Determination 5.11 Source Term Filtration j

i n.12 Meteorology Inquiry 5.13 Dispersion Model 5.14 offsite Air Sample Analysis 5.15 Liquid Release Calculation 5.16 Protective Action Recommendation Logic 5.17 Dose Projection Model overview, TMI-1

+

5.18 TMI-2 Source Term Calculation Each part of this section explains what each process does and how it does it.

l To use the TMI-1 or THI-2 RAC program with an IBM or IBM compatible computer, perform the following steps:

1.

Turn on CRT by pulling out "ON" switch. Adjust brightness / contrast on monitor appropriately.

2.

Check that modem is on (for microcom modem - DTR and mail mode carrots on).

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16.0 2042c i

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"*=' r Radiological Controls Department 6610-PLN-4200.02

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Title Revision No.-

TMI Emergency Dese Calculation Manual (EDCM)

O f

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3.

Turn on printer (ON/CFF switch).

4.

Insert THI-1 or THI-2 Disk into the A disk drive (3.E" disks).

i a.

Turn on computer (right side ON/OFF switch) i l

b.

RAC program will load within 1 minute.

I NOTEi To reload or interrupt program with thr, computer ON ~ hit i

l the Ctri, Alt, Del keys at the same thee, with disk in l

the PC.

j 5.

For computers with RAC program on a hard drives i

a.

Turn on computer.

b.

Menu will appear, choose appropriate RAC Program, or l

I c.

If no menu, use directory RAC1 for THI-1 or RAC2 for THI-2.

l O

d.

Type RAC.

e.

The RAC program will load within 1 minute.

l I

6.

The program options are listed in the bottom line and the range of input allowed in the top lines on the CRT.

(

7.

When finished with each screen's input, push the appropriate function key (ex. F10, F4, F1) for the next function.

?

8.

Dose calculations normally take about 1-2 minut -

r4 will print when t

completed.

l 9.

When dones Remove all disks, turn off CRT. com,.

er, and printer. Store disks appropriately.

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I Title Revision No.

TMI Emergency Dose calculation Manual (EDCM) 0

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i 5.1 source Term Calculations - The source term portion of the TH1-1 dose f

assessment program is used to generate the quantity and radionuclide make l

up of the radioactive material released (or available for release) to the environment. Once the source term is measured or estimate 1, meteorological and dosimetry models are applied to the assessment. Some specific accident scenarios are used to calculate radionuclide release factors-and assess the

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accident consequences.

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Radiological controle Department 6610-PLN-4200.02,,,

Title Revision No.

i TMI Emeroeney Dose calculation Manual (EDCM) 1 5.2 selection of Release Pathways and characteristics - The TM1-1 CDeputer I

program will prompt for the release pathway and the release i

characteristics.

?

5.2.1 The following are the Release Pathways:

1.

OTso Tube Rupture Release Includes: via the condenser off-gas or directly to atmosphere.

2.

Reactor Building Release i

3.

station Vent Release Includes: Auxiliary Building, Fuel Handling Building (FHB) and ESF FHB.

5.2.2 The following are the Release characteristics i

1.

OTSO Tube Rupture via condenser of f-gas 2.

OTSG Tube Rupture directly to stmosphere vi4 the Main Steam Reliefs or Atmospheric Dump Valves 3.

LOCA in the Raactor Building 4.

Fuel Handling Accident in the Reactor Building l

5.

Fuel Handling Accident in the Fuel Handling Building, including ESF Fuel Handling Building Releases 6.

LOCA in the Auxiliary Building-7.

Waste Oas Tank Release 5.2.3 The following choices are then offered for the method to be used in the source term generations 1.

Use Post Accident Sample Result 2.

Use RMS 3.

Use contingency calculation NOTE:

The least accurate to the most accurate source term l

generation methodology is contingency calculations, RMS calculations, and post accident sample result calculations.

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Muclear TM, Radiological Controle bepartment 6610-PLN-4200.02 Title Revision No.

TMI Emeroency Dose Calculation Manual (EDCM) 1 5.2.4 The THI-1 RAC computer program accosunodates airborne releases from the following pathways (See Figure 5.2-1):

A.

The OTSO Tube Ruptures 1.

RM-A5 Condenser off-gas 2.

RM-A5 High-condenser off-gas 3.

RM-025 Condenser off-gas 4.

RM-026 Main 8 team Reliefs and Atmospheric Dump Valves 5.

RM-G27 Main Steam Reliefs and Atmospheric Dump Valves 6.

RM-A5 KAP-5 Samples 7.

Main Steam Release directly to the atmosphere 8.

Contingency Calculations without RMS or Post Accident samples B.

The Reactor Buildings f

1.

RM-A9 Reactor Building Purge 2.

RM-A9 High-Reactor Building Purge 3.

RM-G24 High High-Reactor Building Purge 4.

RM-A2 Reactor Building Atmosphere 5.

CATPASS Samples 6.

RM-A9 MAP-5 Samples 7.

Contingency Calculations without RMS or Post Accident Samples C.

The Station Vent 1.

RM-A4 Fuel Handling Building Exhaust 2.

RM-A6 Auxiliary Building Exhaust 3.

RM-A8 Station Vent (Auxiliary and Puel Handling Buildings) 4.

RM-AB High-Auxiliary Building Exhaust O

20.0 2042c

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i Nuclear Radiological Controls Department 6610-PLN-4200.02 Title Revision No.

TMI. Emergency Dose Calculation Manual (EDCN) 1 5.

M-A14 - Isr Fuel llandling Building Exhaust 6.

AM-A8 MAP-5 samplee 7.

Contingency calculations without AMs or Post Accident samples s

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Title Revision No.

TMI Eniergency Doso Calculation Manual (EDCM) o FIGURE 5.2-1 CT Af piotNillt 6 ADi&1100,lel10b ike

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6610-Pl.N-4 2 00. 02 Title Revision No.

I l

TMf _.Eneroency Dope Calculst tor, wtnual (RDCM) 1 i

t 6.3 calculation of NRc _pamage class and Isotopie Pereontages

• This calculation will determine the mix or percentages of the following fif teen radionuclices.

l 6.3.1 yen woble_ cases rive modteindines 1.

Kr-85m 1.

1-131 e

2.

Kr-86 2.

1-132 6

3.

Kr-87 3.

3-133 l

4.

Kr-88 4.

1-134 6.

Xe-131m B.

1-135 l

1 6.

Xe-133m 7.

Ko-133 8.

Ee-135m 9

Ke-135

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xe-138 5.3.2 The NRC Damage Class Determination The determination of t.ie KRC Damage Class is performed using various core temperature regions from Operations Procedure 1210-6 see Figuru ?.3-1.

The core temperatures used in this section of the program come from operations computer pt C4006, which is the average of the five highest incore thermocouples. The curves relating to saturation, and cladding failures are approximated by straight line equations. NRC damage l

classes 1 - 10 are based on the different pressure and i

I temperature regions of Figure 6.3-1.

l 6.3.2.1 Core Temperature Regions - Figure 5.3 1 The region to the left of curve C represents normal RCS activity, NRC Class-1. The region between curves C and D represents RC8 plus a percentage of gap &ctivity, NRC Class 2 - 4.

The region between Curve 0 and curve E represents RCs plus all gap activity plus a percentage of noble and volwtile fission product release from fuel grain boundaries, (CS, 3, Rb), NRC Class 5 - 7.

The

+

region between curve E and 2560er incore temperature represents ROS activity plus 100% of the gap activity and 100% of the in vessel melt release assuming NUREO-1228 release fractions, NRC.

Class 8 - 10.

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Radiolocacal controls Department 6610-PLN-4200.02

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Title Revision No.

YMI Emergency Dose Calculation Manuni (EDcM) 1 5.3.2.2 The curves represented in Figure 5.3-1 have the ic11owing

}

equations e

curve C: Temperature Temperature a 406.1 + (0.34027

• PRESS)-(0.0000538

• PRESS )

l 8

cuyve D Temperature e

3 Temperature = 690.92 + (0.37178

• PRESS)-(0.0000554

• PRESS )

Curve E Temperature Temperature = 1135.3 + (0.40018

• PRESS)-(0.0000567

• PRESS )

[

3 Where Temperature a Incore Thermocouple Temperature (F');

computer Point C4006.

l PRESS = RCS Indicated Pressure (Psig) 5.3.2.3 The matrix below shows the theory of fuel damage based on l

TDR-431.

NRC DAMAGE CLASS DEGREE OF HINOR INTERMEDIATE MAJOR DECRADATION

<10%

10 - 50%

>50%

No Fuel Damage (RCS)

-No Damage : Class 1 or Class 1A j

Cladding Failure (GAP) 2 3

4 Fuel overheat (Fuel Matrix) 6 7-(

Fuel Melt (Fuel Matrix) 8 9

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Radiological Controls Department 6610-PLN-4200.02

7tgg, Revision No.

,L iWI__tmergency Dose Calculation Manual (EDCM) 1 5.3.3 Caleviation of Radionuelide Mix Percentages Based on NRC Damage Classification. Once the determination of NRC Damage Class 1 - 10 has been determined from the Core Temperature Regions the various radionuclide mix percentages can be calculated based on the distribution of the RCS Activity, GAP Activity, and/or ruel Matrix Activity. The program models the verious combinations of activities for each NRC Damage Class as follows:

RCS CAP FUEI.

NRC DAKY,E ACTIVITY ACTIVITY MATRIX class FRACTION FRACTION FRACTION 1

1 0.0 0.0 2

1 0.1 0.0 3

1 0.5 0.0 4

1 1

0.0 5

1 1

0.1 6

1 1

0.5

[

7 1

1 i

s-8 1

1 1

9 1

1 1

10 1

1 1

5.3.3.1 Therefore, as an examples NRC Damage Class-6 would consist of 1004 RCS Activity, plus 100% of the GAP Activity, plus 50% of the ruel Matrix Activity.

t 5.3.3.2 THI-1 Normal RCS Activity - THI-1 normal RCS Activity (NRC Damage Class-1) and the rapid power transient RCS Activity (NRC Damage Class-1A) are listed in table 5.3.3.2.

The normal default PCS Activity is based on operational RCS activity from fuel cycle 6, on August 3,1990 at 0450. This normal RCS activity (NRC Damage class-1) can be modified in the RAC program, if a current RCS sample result is available. The modified RCS activity will then be stored for future use by the RAC program. If the RCS activity changes it may be changed by the user at any time. The three possible RCS activities available for NRC Damage Class one, ares a) current default RCS activity, b) stored RCS activity, or c) new RCS activity data.

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MUCIS8r rx, Radioloolcal controls Department 6610-PLN-4t00.02

_ Title Revision No.

TM! Emergency Dose Calculation Manual (EDCM) 1 5.3.3.2.1 plant Shutdown Radiciodine spiking represents the " spiking" of the radiciodines and noble gases. Plant Shutdown Radioiodine spiking occurs at < 15% Meactor Power, and is caused by the fuel gap activity being washed out of the fuel pins at transient temperatures and pressures.

The default " spiking" factore for radiciodines and noble gaeed ares Radioiodines times 50. fifM)

Noble gases times 2. (>4; I

The Tsc should be requestr4 to provide the actual " spiking" factors for the particular plant shutdown situation. The default factors are to be used if information is not available f',om the TSC.

This " spiking" of radioiodin6 and noble gas activities. represent an increase in RCS radioactivity due to a plant evolution and do not represent an indication of fuel damage; i.e.,

NRC Damage caisses 2-10.

spiking only occurs when fuel defects are present.

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Radiological Controls Departnent 6610-PLN-4200.02

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Title Parision No.

TH! Emergeney Does Calculation ManJai (EDcM) 1 l

5.3.3.3 Tables of NRC Damage Classes 1 and 1A l

l i

• THI-1 NORMAL RCS PLANT SHUTDOWN RADICIODINE 1

SPIKING - RC8 NRO DAMAGE DAMAGE CLASS 1 CLASS 1A pCi/et;**

Percent curiose*

pC1/cc Percent curies **

1-131 1.295-03 0.65 2.765-01 6.453-02 1 33 1.383+01 1-132 2.088-02 10.52 4.468+00 1.04E+00 21.40 2.23E+02 I-133 1.45E-02 7.34 3.115+00 7.255-01 14.92 1.55E+02 i

2-134 3.222-02 16.29 6.90E+00 1.61E+00 33.14 3.45E+02 I-135 2.423-02 12.24 5.188+00 1.21E+00 24.90 2.598+02 SUsroTAL 9.30E-02 47.05 19.92 4.65E+00 95.69 9.963+02

?

KR-85M 3.503-03 1.77 7.50E-01 7.00E-03 0.14 1.50E+00 l

KR-85 0.00E+00 0.00 0.00E+00 0.00E+00 0.00 0.00E+00 KR-87 6.818-03 3.45 1.46E+00 1.363-02 0.28 2.92E+00 f

KR-88 8.455-03 4.28 1.81E+00 1.695-02 0.35 3.62E+00

+

12-131H 0.00E+00 0.00 0.003400 0.00E+00 0.00 0.00E+00 i

XE-133M C.00E+00 0.00 0.00E+00 0.00E+00 0.00 0.00E+00 IE-133 3.09E-02 15.63 6.62E+00 6.18E-02 1.27 1.32E+01 0

XI-135M 7.91E-03 4.00 1.695+00 1.58E-02 0.33 3.398+00 i

l KE-135 2.71E-02 13.71 5.81E+10 5.422-02 1 12 1.16E+01 KE-138 2.00E-02 10.12 4.28E+00 4.00E-02 0.82 8.573+00 8UbT0TAL 1.05E-01 52.95 22.42 2.0iE-01 4.31 4.485+01 TOTAL 1.98E-01 100.00 4.23E+01 4.86E+00 100.00 1.04E+03 NOBLE CAS TO 1.13 1.13 1.13 0.05 0.0S 0.05 IODINE RATIO

• " NORMAL" RCS CONCENTRATION AND PERCENTAGES ARE FROM NORMAL CYCLE D RC8 OPERATIONAL DATA

• • THESE CURIE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 QALLONS l

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i Nuclear

,M, Radiological Controle Department 6610-PLN-4200.02 Title Reviolon No.

TNT Imeroency Dose calculation Manual (EDCM) 1 6.3.3.4 Gap Activity and Fuel Matrix Activity - Cap activity and ruel Matrix activity used in the program are determined from TDR-431.

These curie activities are based on irradiation of the entire core at full power, 2535 MW,,

for 930 days.

THI UNIT 1 CAP ruel Matrix Puel Act.

CAP Act.

Matrix Isotope curies curies Kr-85m 4.84E+04 0.45 2.13E+07 2.36 Kr-85 7.40E+04 0.70 8.595+04 0.01 Kr-87 2.63E+04 0.2$

3.90E+07 4.33 Kr-88 6.672+04 0.63 5.91E+07 6.56 Re-131m 7.96E+04 0.75 5.40E+05 0.06 Re-133m 9.30E+04 0.87 3.09E+06 0.34 Re-133 8.34E+06 78.31 1.28E+08 14.20 Ee-135m 2.72E+04 0.26 3.375+07 3.74 Re-135 3.45E+04 0.32 1.59E+07 1.76 Re-138 0.00E+00 0.00 0.00E+00 0.00 1-131 1.29E+06 12.11 6.37E+07 7.07 1-132 1.05E+0S 1.74 9.70E+07 10.76 l

I-133 2.79E+05 2.62 1.43E+08 15.86 I-134 1.74E+04 0.16 1.67E+08 18.53 1-135 8.83E+04 0.83 1.30E+0B 14.42 Sum 1.13E+07 100.0 Sum 9.02E+08 100.0 l

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Radiological Controls Department 6630-PLN-4200.02

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Title Revleion No.

THI Emergency Dose calculation Manual (EDCM) 1

[

5.3.3.5 The following tables represent the NRC Damago Class 2 - 10 mikes, percentages, curies, and concentrations used in the RAC code.

DAMAGE DA.4 AGE class 2 CLA15 3 pCi/ec'*

Percent curies pC1/ce**

Percent Curies I-131 6.03E+02 12.09 1.29E+05 3.01E+03 12.11 6.45E+05' I-132 8.66E+01 1.74 1.85E+04 4.32E+02 1.74 9.25E+04 I-133 1.31E+02 2.62 2.79E+04 6.52E+02 2.62 1.40E+05' I-334 8.43E+00 0.17 1.80E+03 4.10E+01 0.16 8.76.E+03 2-135 4.15E+01 0.83 8.88E+03 2.07E+02 0.83 4.42E+04 7

SUBTOTAL 8.70E+02 17.45 1.86t+05 4.35E+03 17.46 9.30E+05 KR-85M 2.28E+01 0.46 4.872+03 1.13E+02 0.45 2.42E+04 KR-85 3.50E+01 0.70 7.48E+03 1.75E+02 0.70 3.74E+04 KR-87 1.25E+01 0.25 2.67E+03 6.16E+01 0.25 1.32E+04

[

KR-88 3.15E+01 0.63 6.73E+03 1.56t+02 0.63 3.1t"'04 KT-131H 3.72E+01 0.75 7.962+03 1.86E+02 0.75 3.98E+04 KE-133M 4.36E+01 0.87 9.32E+03 2.17E+02 0.87 4.65E+04 KE-133 3.90E+03 78.29 8.35E+05 1.95E+04 78.30 4.173+0G KE-135M 1.28E+01 0.26 2.74E+03 6.37E+01 0.26 1.36t+04 KE-135 1.72E+01 0.34 3.668+03 8.17E+01 0.33 1.75E+04 I

KE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 SUBTOTAL 4.12E+03 82.55 8.81E+05 2.05E+04 82,54 4.40E+06

~~

TOTAL 4.99E+03 100.00 1.07E+06 2.49E+04 100.00 5.33E+d6 NOBLE GAS 4.73 4.73 4.73 4.73 4.73 4.73 i

TO 10 DINE i

RATIO P

i

• • THESE FCi/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF $6,595 GALLONT 5

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DAMAGE DAMAGE I

CLASS 4 C1. ASS 5 pCi/cc**

_ Percent Curies pC_1/cc**

Perecht Curles_

1-131 6.03E+03 12.11 1.29E+06 3.58t+04 7.60 7.662+06 I-132 8.65E+02 1.74 1.85E+0$

4.62E+04 9.81 9.99E+06 I-133 1.30E+03 2.62 2.79E+05 6.81E+04 14.46 1.46E+07 I-134 8.16E+01 0.16 1.75E+04 7.81E+04 16.59 1.67E-07 I-135 4.13E+02 0.83 8.83E+04 6.12E+04 12.99 1.31E+07 SUBTOTAL 8.698+03 17.46 1.86E+06 2.89E+05 61 44 6.195+07 l

1.02E+04 2.16 2.185+06 KR-85M 2.26E+02 0.45 4.84E+04 KR-85 3.60E+02 0.70 7.48E+04 3.90E+02 0.08 8.34E+04 KR-87 1.23E+02 0.25 2.63E+04 1.83E+04 3.90 3.93E+06 KR-88 3.12E+02 0.63 6.602404 2.79E+04 5.93 5.98E&O6 IE-131M 3.72E+02 0.75 7.965+04 6.24E+02 0.13 1.34E+05 II-123M 4.35E+02 0.87 9.30E+04 1.88E+03 0.40 4.02E+05 IE-133 3.90E+04 78.31 8.34E+06 9.88E+04 20.97 2.11E+07 IE-135M 1.27E+02 0.26 2.721+04 1 59E+04 3.37 3.40E+06 XE-135 1.62E+02 0.33 3.47E+04 7.59E+03 1,61 1.62E+06 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 SUBTOTAL 4.11E+04 82.54 8.79E+06 1.82E705 38.56

-3789E+07 TOTAL 4.98E+04 100.00 1.07E+07 4.71E+05 100.00 1.01E+08 NOBLE CAS 4.73 4.73 4.73 0.63 0.63 0.63 10 DINE RATIO t

• • THESE pCi/cc VALUES ARE BASED CN NORMAL R05 VOLUME OP 56,595 GAI.LONS 1

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,m Radiological controls Department 6610-PLN-4300.02 Title Revision No.

TNT. Emergency Dose Calculation Manual (EDC')

1 i

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DAMADE DAMADE CLASD 6 CLASS 7 sci /cc**

Percent curies pCi/cc88 Percent Curles E

.31E+0?

3.04E+05 7.13 6.50E+07 1

Y-131 1.55E+0L 7.18 s

I-132 2.20E+05 10.55 4.875+07 4.54E+05 10.66 9.72E+07 1-133 3.35E+05 15.56 7.18E+07 6.70E+05 15.71 1.43E+0B

[

1-134 3.90E+05 18.10 8.35E+07 7.80E+05 18.31 1.67R+08 i

I-135 3.04E+05 14.11 6.51E+07 6.08E+05 14.26 1.30E+08 t

SUBTOTAL 1.41E+06 65.50 3.02E+08 2.82E+06 66.07 L,03E+08

%R 85M 5.00E+04 2.32 1.075+07 9.985+04 2.34 2.13E+07

[

KR-85 5.50E+02 0.03 1.18E+05 7.51E+02 0.02 1.61E+05 KR-87 9.12E+04 4.23 1.95E+07 1.82E+05 4.28 3.90E+07-i KR-88 1.3RE+05 6.42 2.96E+07 2.765+05 6.49

,5.92E+07 XE-131H 1.63k+03 0.08 3.50E+05 2.90E+03 0.07 4.20E+05 XE 133H 7.65E403 0.36 1.64E+06 1.49E+04 0.35 3.18E+06 1E 133 3.38E+05 15.68 7.23E+07 6.37E+05 14.95 1.36E+08 XE-135M 1.892+04 3.66 1.69E+07 1.58E+05 3.70 3.37E+07 KE-135 3.73E+04 1.73 7.98E+06 7.45E+04 1.75 1.59E+07 4

XE-138 1.70E-01 0.00 3.64E+01 1.70E-61 0.00 3.64E+01 SUBTOTAL 7.44E+05 34.50 1.59E+08 1.E5E+06 33.43 3.10E+08 I

l TOTAL 2.16E+06 100.00 4.61E+08 4.26E+06 100.00 9.12E+08 y

NOBLE CAS 0.53 0.53 0.53 0.51 0.51 0.51 IODINE I

l RATIO I

'T12SE pCi/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 CALIANS l

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1 Nucleev T,1 Radiological controls Departnent 6610-PLN-4200.02 Title Revision No.

THI Emergency Dose calculation Manual (EDCM) 1 DAMA0E DAMADE class 8 CLASS 9 pCi/ce** _._ Percent Curtes pC1/ce**

Percent Curies I - 13.*

3.04E+05 7.13 6.bOE+07 3.04E+05 7.13 6.50E+07 I-132 4.54E+0S 10.66 9.72E*07 4.54E405 10.66 9.72E+07 I-133 6.70E+05 15.71 1.43E+08 5.70E+05 15.71 1.43E+08 I-134 7.80E+05 18.31 1.67E+08 7.803+05 18.31 1.67E+08 I-135 6.08E+05 14.26 1.30E+08 6.03E+05 14.26 1.30E+08 SUBTOTAL 2.82E+06 66.07 6.03E+08 2.02E+06 66.07 6.03E+08 KR-85M 9.98E+04 2.34 2.13E+07 9.98E+04 2.34 2.13E+07 KR -85 7.51E+02 0.02 1.61E+05 7.51E+02 0.02 1.61E+05 KR-87 1.02E+05 4.28 3.90E+07 1.82E+05 4.28 3.90E+07 KR-88 2.76E+05 6.49 5.92E+07 2.76t+05 6.49 5.92E+07 XE-131H 2.90E+03 0. '17 6.20E+05 2.90E+03 0.07 6.20E+05 KE-133M 1.49E+04 0.35 3.18E+06 1.49E+04 0.35 3.18E+06 KE-133 6.37t+05 14.95 1.36E+08 6.37E+05 14.95 1.36E+08 IE-135M 1.58E+05 3.70 3 37E+07 1.58E+05 3.70 3.37E+07 XE-135 7.4FE+04 1.75 1.598+07 7.45E+04 1.75 1.59E+07 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 SUBTOTAL 1 45E+06 33.93 3.10E+08 1.45E+06 33.93 3.10E+08 TOTAL

4. 2 6E+ 0-i 100.00 9.12E+29 4.26E+06 100.00 9.12E+08 NOBLE GA3 0.51 0.51 0.51 0.51 0.51 0.51 IODINE RATIO

• • THESE pC1/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS 1

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Title Revision No.

1 TMI Emergency Dose Calculation Manuel (EDCM) 1 1

i DAMAGE class 10 pCi/cc**

Pe rcerit

_yurien 1-131 3.04E+05 7.13

' 50k+07 l

T~~3*

4 54E+05 li 66 9.72E+07

- 4.'n 70E+05 15.71 1.43E+08

.80E+05 18.31 1.67E+08 6.08E+06 14.26

1. 30E + 08 l

$USIC. AL 2.82E+C6 66.07 6.03E+08 KR-8BM 9.9SE+04 2.34 2.13E+07 KR-85 7.51E+02 0.02 1.61E+05 KR-87 1.82E+05 4.28 3.90E+07 KR-88 2.76E+05 6.49 5.92E+07 XE-131H 2.90E+03 0.07 6.20E+05 KE-133M 1.49E+04 0.35 3.18E+06 XE-133 6.37E+05 14.95 1.36E+08

[

XE-135M 1.58E+05 3.70 3.37E+07 XE-135 7.45E+04 1.75 1.59E+07

' XE-138 1.70E-01 0.00 3.64E+01

[

k SUFTOTAL 1.45E+06 33.93 3.10E+C8

\\s_J' TOTAL 4.26E+06 100.00 9.12E+08 NOBLE GAS 0.51 0.51 0.51 i

IODINE RATIO

• • THESE pCi/cc VALUES ARE BASED ON NORMAL RCS VOLUN.C OF 56,595 GALLONS 5.3.3.6 Calculation of Radionuclide Mix Percentages - Once the computer has determined the total amount of combined activities or curies for a certain NRC Damage Class, thesa curies are then normalised j

to 1001, i.e.,

the percertage of each radionuclide in the total mix is calculated. These percentages are then displayed on the screen along with the NRO Damage Class,1 - 10.

Exceptions - The above percentages are replaced in cases where an assumed mix is more appropriate. These cases are 1.

Contingency calculations fort a.

Spent Fuel Accident in the Fuel Hand:,ing Building -

FSAR mix is assumed.

b.

Fuel Cask Accident in the Fuel Handling Butiding - FSAR-mix is assumed.

33.0 2042c r

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,,1 Radiological controls Department 6610-PLN-4200.02 Title Mvision No.

l TWI Emeroency Dose calculation Manual (EDCM) 1 c.

Spent Fuel Accident in the Reactor Building - FSAR mix is assumed, d.

Weete Gas Decay Tank - FSAR mix is assumed.

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Radiological controls Department 6610-PLN-4200.02 Title Revision No.

__ TMI Emergency Dose Calculation Manual (EDCM) 1 5.4 Radiation Monitoring System (RMS) Source Term Calculation - This section of the program allows the user to determine an ef fluent source term from readings on the TMI-l Radiation Monitoring System.

5.4.1 Only those RMS channels available for a selected release pathway-art offered to the user. Those are listed in Section 5.2.

To calculate a source term from a RMS ret. ding the following i

parameters are used:

1.

RMS READING:

CPM, mR/HR, OR CPM / MIN 2.

RMS CHANNEL EFFICIENCY RELATING To THE CALIBRATION NUCLIDE CPM /pCI/CC 3.

THE METER RESPONSE FACTOR 4.

THE NRC DAMAGE CLASF MIXTURE 5.

THE RELEASE FLOW RATE 5.4.2 In order to gather the above information the program will proceed in the following manner.

f\\

1.

Once a release pathway has been chosen the first option to

\\,,_ /

the user is whether or not to decay the mixture from the time of reactor shutdown. If "yes" is chosen the program decays the eventual mixture based on the time -input by the user.

2.

The appropriate radiation monitors for the pathway chosen are then displayed. The user then chooses the radiation l

monitor that le the most representative of the release in terms of "on scale" and the proper range.

3.

The next information required is the actual radiation monitor reading in counts per minute, mr/ hour,'or counts per minute per minute (based on a rise time).

4.

The next data needed is for the flow rate for the release pathway. The applicable-flow rates are discussed in the effluent flow rate section. Appropriate default values are also listed.

5.

Once the above data has been entered, the program will use the RMS reading, the particular radiation monitor's ef ficiency, the monitor response f actor, the nuclide fraction from the isotopic per:entage section of the program relating to -NRC damage class determination, radionuclide mix l

percentages, and the associated flow rate to the environment to calculate a source term. Source terms are identified for the noble gas source term and for the radiciodines.

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(

36.0 2042c l

N NI TMI Radiological Controls Department 6610-PLN-4200.02 Title Revision No.

_ TMI Emeroeney Dose Calculation Manual (EDCM) 1 5.4.3 The calculations are performed in the following fashion:

1.

First the total monitor response factor is calculated by multiplying the individual nuclide percentages from the NRC damage class :*etermination by the individual nuclide monitor response facters.

15 M=

E P

• In 1

total monitor response factor Where M =

individual nuclide percentages from NRC j

P =

damage class I, =

individual nuclide monitor response factors The I,'s for the various RMS detectors are listed as follows:

N x

37.0 2042c.

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Title Revision No.

THI Emergency Dose L_sleulation Manual (EDCM) 1 INDIVIDUAL MIXTURE RESPONSE FACTORS (I,)

}

Beta Scint. Detectors RM-A2Lo, RMA4Lo, Scintillation Ion GM Tubes RM-A6Lo, RM-A5Lo, Detectors Chamber RM-ASHi, RM-A8Hi(3)c RM-A8Lo, RM-A9Lo(4)

NUCLIDE RM-026 & RM-G27(1) RM-G24(2)

RM-A9Hi, RM-025 RM-A15Lo RM-A14Lo Kr-85m 70.7 212.2 2.35 1.92 Kr-85 1.0 1

0.011 1.98 Kr-87 356 324.39 3.59 9.12 Kr-88 1160 334.15 3.70 2.78 Xe-131m 9.01 4.88 0.054 0.0 Xe-133m 10.6 35.15 0.378 0.0 Xe-133 0.0 90.24 1

1.0

(<80 kev)

Xe-135m 193 195.12 2.16 0.0 b

Xes-13 5 111 221.95 2.54 2.59 l

Xe-138 1560 939.02 10.41 4.62 I-131 172 240 2.66 I-132 1030 747.8 8.286 I-133 274 219.51 2.432 I-134 1180 542.44 6.011 I-135m 706 341.46 3.784 (1)

E MeV* Dis Nuclide i

E MeV* Die cal Nuclide (calibration isotope is Kr-85, threshold set to exclude Xe-133 at 80 kev.)

(2)

It Probability Nuclide E% Probability Cal. Nuclide (calibration nuclide is Kr-85)

I (3)

Et Probability Nuclide It Probability Cal. Nuclide (calibration isotope is Xe-133; values for RM-A5Hi, except Xe-133, will be multiplied by 4)

(4)

E Beta decay probability

• Beta end-pt. energy nuclide E Beta decay probability

• Beta and-pt. onergy cal. nuclide (cal. isotope le Xe-133)

• Padiolodines filtered out prior to noble gas channel

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38,0 2042c i

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"""5" Radiological Controls Department 6610-PLN-4200.02

/

Title Revision No.

TMI Emergency nose calculation Manual (EDCM) 1 2.

The noble gas source term in pci/see is now calculated using the following equation and input data 1

• Ng/100) * (Flow)

• 472 Ngut = ( 1

• ACT

• _ Me H

Where Ngst =

Noble Gas source term in pCi/sec.

Total monitor response factor, unitless.

M

=

cpm, mR/hr, cpm / min reading from the monitor.

Act

=

Monitor sensitivity in cpm /pci/ce, mR/Hr/pci/cc, He

=

or epm / min /pi/cc.

Flow =

Flow rate in CFM cc/sec/CFM.

472

=

Sum of Noble gas percentages from the selected Ng

=

NRC classification.

3.

The radiciodine source term in pi/see is then calculated using the

..oble gas source term and the noble gas to iodine ratio, as discussed in Section 5.8.

I Riot = Ngst

• Ri

• 1 Ng Tfcf (if applicable)

Where Rist = Radiciodine source term-in pCi/sec.

Ri = The radioiodine to noble gas ratio.

Ng Tief =

Two phase steam factor - If this is a steam release with water; the water will-tend to keep the radiolodines in solution. -

l 4.

The noble gas and the radiolodine source terms are then multiplied by the individual isotopic percentages of the NRC damage class mixture to determine the pC1/sec of each of the 15 nuclides, 10 noble gas and 5 radioiodines.

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39.0 2042c

MUClear rx1

- Radiological controls Department 6610-PLN-4200.02 Revision No.

(

Title 1

TMI Emergency Dose Calculation Manual (EDCM) 5.5 Post Accident samples Source Term Calculation - One option of the RAC program is to use actual plant effluent sample results to devalcp the release source terms. This is in f act the preferable method for estimating release quantities if the sample results are available since the method eliminates some of the built in conservatisms of using monitor readings, or contingency calculations. Using sample results also eliminates errors in the source term when the actual release mixture is different from the assumed mix.

The routines which allow use of the post accident samples contained in the RAC programs provide the mena selectors to call the different subroutines for each type of post accident samole.

5.5.1 One menu selection is the sample station / method to be used. For the Reactor Building three options are presented: 1) CATPASS (Containment Atmosphsre Post Accident sample System),

2) Marinelli/prefilter (marinelli with a particulate and lodine filter upstream), or 3) MAP-5, Radiolodine Processor Station.

For the condenser off-gao or the Aux /FHB release pathways, only the Marinelli/prefilter and MAP-5 samples are available. A menu will appear on the coraputer screen which will list the two or three available sample methods and prompt with ' enter choice'.

When a choice - 1, 2, or 3 as appropriate - is entered the program will continue.

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5.5.2 The MAP-5 program will prompt for each of the identified

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radioiodine species in the silver zeolite or charcoal cartridge sample from the MAP-5 Processor Station. It will place the value for each of the five radioiodine species into one of the elements of the five element array in microcuries per ec.

It will than sum the resulta and print out the sum.

The user is provided the l

options to decay the mixture from time of shutdown and from time of the sample. NORMALLY THE DECAY CORRECTION WOULD NOT BE APPLIED FROM TIME OF SHUTDOWN SINCE THE ANALYSIS ITSELF ACCOUNTS FOR THAT.

If the release is from the Reactor Building, an option selection is provided to determine if the Reactor Building is isolated or not and if the release is proposed or in progress.

It then will adjust the noble gases. Since the MAP-5 only provides information en the radioiodines, the expected ratio between the iodines and noble gases is used to approximate the noble gas activity. Since the MAF-5 is downstream of the charcoal filters, if the charcoal filters are effective, the noble gas activity will be increased by a factor of ten to account for the filter reduction of the icdines. Based on the l

release rate in CFM the isotopic concentrations in pC1/cc are converted to a release rate in pCi/sec for the final source term.

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40.0 2042c

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Radiological Controls Department 6610-PLN-4200.02

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Title Revision No.

x.

TMI Emergency Dose Calculation Manual (EDCM) 1 5.5.3 The CATPASS progrta prompts for the 10 noble gas and the five radiciodine nuclides identified from air stupling. The noble gas and iodine activities in microcuries per cc are put into an array. Options are then providoc for decaying the mix from the j

sample to dose projection time. Since the CATPASS only applies to the contajnment, the options are again provided_to sele:t whether the containment is isolated or not and if the release is proposed or in progress. Calculations are made using the input activities to develop new isotopic percentages and the activities

)

are changed from pC1/cc to pCL/see based on the release rate defined to arrive at the final source term.

5.5.4 The marinelli program is called if the marinelli/profilter option is selected. This option is available for all three release pathweys. Since the production of the source term is based on the measured isotopics, as did the CATPAOS program, the marinelli program proceeds in an identical manner.

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Radiological Controls Department 6610-PLN-4200.02 k_,/

Title Revision No.

s TMI Emergoney Dose Calculation Manual (ELCM) 1 5.6 Contingency Calculations Source Term Generation - The contingency calculations attempt to determine a source term based upon a prioritized set of plant conditions. The user is guided through a set of questions in order to model the contingency calculation with the best obtainable information.

In this way, credible conservative assumptions, as defined in the FSAR default parameters, are replaced with real-time accident conditions as indicated by plant instrumentation. This will make the calculated source terms more realistic 5.6.1 In the contingency program the'previously determined release pathway is utilized to selects 1

Secondary Side Release 2.

Reactor Building Release 3.

Station ventilation Releases The " Secondary Side Release" includes accidents that result in release via: The condenser off-gas, the atmospheric dump valven, the main steam reliefs, and a main steam line rupture.

p.

The " Reactor Building Release" includes accidents that result ita

(

a releave from the Reactor Building vias the purge duct, when

's the purge valves are open, or design basis leakage, when the s

purge valves are closed.

The " Station Ventilation Releases" include accidents that result in a release from the Auxiliary Building, Puel Handling Building, or ESF Puel Handling Building.

5.6.2 A " Secondary Side Releaso" contingency.claculation is calculated by identifying four parameters:

1.

RCS Activity (D1) pC1/cc i

2.

Primary to Secondary Leakage (D2) gpm 3.

Transport Fraction (D3) 4.

Two Phase Release (Tfcf) 5.6.2.1 The "RCS Activity" le determined utilizing:

f 1.

RM-L1 High (A1) cpm, D1 = Al/22.2 pC1/cc 2.

RM-L1 Lo (A1) cpm, D1 = Al/1270 pCi/cc 3.

Most recent RCS sample, in pC1/cc, or

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Radiological Controle Lepartment 6610-PLN-4200.02 Title Revision No.

TM1 Emerconey Dose Calculation Manual (EDCM) 1 4.

Def ault to a RCS concentration dependent on the NRC Damage Class RCS Default NRC Damage Class pCi/cc 1

• 1.98E-01 lh

• 4.86E+00 (Using default spiking factors) 2 4.99E+03 3

2.49E+04 4

4.98E+04 5

4.71E+05 6

2.16E+06 7

4.26E+06 8

4.26E+06 9

4.26E+06 10 4.26E+06 Damage classen 1 and 1A are variable, based on RCS activity entered from sample data and radiciodine spiking factors m

used.

(G 8

5.6.2.2 The " Primary to Secondary Leakage" is determined utilizing 1.

RCS identified leakage [D2] gpn 2.

Default to 400 gpm for a double-ended tube shear [D2]

5.6.2.3 The " Transport Fraction" [D3] is a function of the release pathway.

[D3) is calculated by the equation: D3=Fr

• 0.0075 +

(1-Fr) where Fr = fraction of the releass via the condenser off-gas. For a release through th condenser off-gas the noble l

gas transport is 1.00, the radio.odine transport fraction is i

0.0075.

The radiciodine transport fraction is a product of: The

(

fraction of radiciodine entering the OTSG from the RCS that is a l

volatile iodine species (.05) and the partition factors for volatile iodine species in the main condenser (.15).

Non-volatile lodine species have a partition factor of zero in the condenser off-gas. For a release direct to atmosphere the l

noble gas and radiciodine transport-fractions use the fraction of steam released to total steam flow from the OTSG. An additional partition factor [Tfcf} is applicable for a two phase direct release. The resultant source terms in pC1/sec are calculated by:

Ngst = D1

• D2

• Ng/100

• 63.09 Rist = D1

• D2

• D3

• RI/100

• 63.09

• 1/Tfef.

D%

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i Radiological Controls Department 6610-PLN-4200.02

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Title Revision No.

j TMI Emergency Dose Calculation Manual (EDCM) 1 i

I S.6.3 A " Reactor Building Release" contingency calculation is calculated by one of two methods.

If the accident type is a LOCA then four parameters are identified 1.

RCS Activity (A2] pCi/cc 2.

RCS Leakage to RB (A3) gpm 3.

Transport Fraction, E4 = 0.1 or 1.0 depending on the Reactor Building Spray status 4.

Release Flow Rate CFM; E3 = flow

• 472 to convert CFM to ec/sec 5.6.3.1 The RCS Activity is determined utilizing:

1.

RM-L1 High Channel (A1) cpm; [A2) = [A1)/22.2 pC1/cc 2.

RM-L1 Low Channel (A1) cpm; [A2) = [A1)/1270 pCi/cc 3.

Representative RC4 marple results [ A2] in poi /cc 4.

Default to 6 U:S concentration dependent on the NRC Damaga g

Class:

Rr:S Default NRC Damage Class pC1/cg 1

• 1.98E.01 1A 84.86E+00 (Using default spiking facters) 2 4.99E+03 3

2.49E404 4

a.98E+04 5

4.71Ee03 6

2 162+C6 7

4.2(K+06 l

l 8

4.26E+06 l

9 4.26E+06 l

10 4.26E+06 Damage classes 1 and 1A are variable, based on RCS activity entered from sample data and radiciodine spiking factors used.

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Radiological Controle Department 6610-PLN-4200.02,

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Title F,avision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 l

I l

4.

Default mix according to core condition as previously l

identified. The noble gas and radiciodine released to the Reactor Building is calculated as follows:

El = A2

• A3

• Ng/100

• 3785/5.6E10 pCi E2 = A2

• A3

• Ri/100

• 3785/5.6E10 pCi t

5.

If the accident type is a Fuel Handling Accident in the Reactor Building, then the number of damaged fuel rods le identified by the " user" or an FSAR default condition is used.

El = 1.7

• Num rod /208 i

E2 = 0.05

• Num rod /208 5.6.3.2 RCS LEAKAGE - The RCS leakage to the Reactor Building is determined by requesting the " total gallons of RCS leakage into the RB".

The transport fraction is determined on the basis of the status of Reactor Building Spray. The Noble gas transport fraction is assumed to be 1.00.

Radiolodine concentration iR f-s reduced as a result of plateout of elemental iodine. The

(

transport fraction for instantaneous radioiodine plateout is 0.5.

x An additional adjustment of the radioiodine concentration in the Reactor Building is necessary when Reactor Building Spray is activated.

l 5.6.3.3 TRANSPORT FUNCTION - This fraction in used to reduce the f

radioiodines available for release in the Reactor Building, if the Reactor Building spray is utilized. This fraction of 0.1, will reduce the radiolodines by 90%.

5.6.3.4 RELEASE FLOW RATE - The release flow rate is determined via flow rate recorder FR-148 if the purge valves are open.

If the purge valves are closed the release flow rate is determined via the design basis RB leakrate adjusted for actual RB internal pressure as indicated on PT-291.

5.6.4 A " Station Ventilation Release" contingency calculation is calculated by utilizing the FSAR fuel assembly mix or WGDT mix.

1.

Waste Gas Release 2.

Fuel Handling Accident in the Fuel Handling Building.

l 3.

ESP Fuel Handling Release.

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Radiological controls Deptrtment 6610-PLN-4200.02 Title Revision No.

TMI Emeroency Dose calculation Manual (EDCM) 1 5.6.4.1 The extent of the accident is modeled bya 1.

Determining the number.of damaged fuel rods for a fuel handling accident in the Fuel Handling Building, or for the EST THB.

Ngst = 4.2E.

• Nam rods /56 pC1/Second Rist = 750

• Num rods /56 pC1/Second 2.

Using the worst case FSAR source term for WCDT's of 10,000 curies of noble gas and 5 curies of radiolodine, or using the typical WGDT FSAR mix of 1000 curies of noble gas and 5 curies of radiciodine.

3.

Determining the curies released for a Waste Gas Accident.

Typical source term based on a typical inventory of Ngst = 1.0E9/Dr pCL/second Riot = 1.0E5/Dr pC1/second p

or i

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FSAR worst cases Ngst = 1.0E10/Dr pCi/second Rist = $E6/Dr pCi/second Wheret Dr = duration of release.

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Title Revision No.

f THI Emergency Dose calculation Manual (EDCM) 1 5.7 Decay scheme Calculation - The user has the option to (1) decay the postulated mixture from the time of reactor shutdown to time of the dose 1

projection or (2) decay sample data from the time the sample is obtained to the time of dose projection. Subroutine (decay) only decays forward in time. This subroutine (decay) adjusts the individual nuclide percentages according to the conventional exponential decay equations A = A, exp (-it) 5.7.1 Fifteen isotopes are decayed according to the equation N(w) = I(w)

• EXP (-decay time

• f(w))

where:

I(w)

= postulated isotopic percentage decay time a user input time f(w)

= isotopic decay constants read from data flies r

5.7.1.1 The adjusted isotopic percentages N(w) are corrected for Xenon buildup due to iodine decay. For Xe-131m the equations are:

S1 = I(11) - N(11)

N(5) = 0.88

• S1 + N(5) l whore:

S1 = amount of I-131 decayed 0.88

• S1 = amount of Xe-131M buildup

[

5.7.1.2 For Xe-133M the equations are:

S1 = I(13) - N(13)

N(6) = 0.02

• S1 + N(6) where:

4 S1 = amount of I-133 decayed 0.02

• S1 = caount of Xe-133m buildup N(7) = 0.98

• S1 + N(7) calculates the amount of Xe-133 buildup from I-133 47.0 2042c l

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.4UCIS8r Radiological Controls Department 6610-PLN-4200.02 I

Title Revision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 5.7.1.3 For Xe-135m and Xe-135 the equations are 61

= I(15) - N(15)

N(8) = 0.3

• S1 + N(8)

N(9) = 0.7

• S1 + N(9) where:

51 = amount of I-135 decayed 0.3

• S1 = amount of Xa-135m buildup 0.7

• S1 = amount of Xe-135 buildup 5.7.1.4 The isotopic percentages are recalculated as:

1(w) = N(w)/ Sum (N)

• 100 where

['~

N(u) = adjusted / corrected postulated isotopic percentages Sum (N) = sum of the fifteen isotopic percentages I(w) = final isotopic parcentages based upon 100.

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Radiological controls Department 6610-PLN-4200.02 (s_,/

Title Rev.ision No.

TMI Emercency Dose Calculation Manual (EDCM) 1 5.8 Noble Gas to Todine Ratio Calculations 5.8.1 Whether performing dose projections based upon RMS readings, post accident samples or contingency calculations, it may be necessary to compute the NOBLE GAS TO IODINE RATIO. The uses of this ratio are discussed below.

An airborne release from a nuclear power plant will primarily consist of noble gases and radiolodines. Except in the sont severe and improbable accident scenarios, radioactive particulates are not expected to be important dose contributors.

The RAC program was designed to incorporate ten noble gases and five radioiodines.

I The 15 isotopes are considered to be the most radiologically l

significant gaseous isotopes availabla for release from an operating nuclear power plant. Pertinent radioactive decay parametera such as half life, average gamma energy per disintegration and average beta energy per disintegration for each isotope are stored in data statements within the program.

I Along with individual isotope source term information, this data is used to determine dose mate conversion factors and dose rates fr~'s that are specific to the isotopic nLxture being released. These N -)

calculated quantities can be adjusted to account for radioactive l

decay during the accident sequence.

s 5.8.2 The Tl!I-1 RAC model always projects both thyroid and whole body dose rates at specified downtind distances. Consequently un estimate of the isotopic release rate is necessary for both iodi.ies and noble gases. Under normal circumstances the program starts with a core inventory of all fifteen isotopes and traces the progress of each one through various systems or proceoses until it is released. Depending upon the type and severity of the accident and the engineered safety systems that have been activated, the isotopic ratios can vary widely. There are some circumstances where the release rates of specific isotopes may be zero or negligibly small.

But, in general, the program accounts for the fifteen isotopes listed above.

In certain circumstances it is not possible to obtain release rates for all fifteen isotopes individually. For example, some plant effluent monitors have only noble gas channels while others have particulate, lodine and gas channels. The MAP-5 sampling system yields only iodine information, where the CATPASS and the Marinelli gas sampling systems yield information on all fifteen isotopes. For release pathways where information on both noble gases and iodines is not available, the RAC program uses the noble gas to iodine ratio to fill in the missing information.

The following example illustrates the use of this ration

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Radiological Controls Department A610-PLN-4200.02

'd Title Revision No.

TH! Emergency Dose calculation Manual (EDcH) 3 A certain type of reactor accident has occurred. Based on an assessment of the degree of core damage and the accident type, the computer selects a default mixture of 15 nuclides and calculates the fraction of the mix that each isotope represents.

The noble grs to iodine ratio is also calculated. Assume that the ratio was equal to 5/1 in this case. Also aestume that an iodine sample was taken which indicated a total radioiodine release rate of 5000 pC1/sec. Usir.g the noble gas to iodine

{

ratio in the absence of specific noble gas measurements, the computer would calculate a gross noble gas release rate of 25,000 pCL/sec.

It would also calculate individual noble gas release rates by using the isotopic fractions from the default mix.

5.8.3 To summarire, the highest quality information available is a quantitative measurement of each nuclide. This type of information is available from RCS, gas Marinelli and profilter, and CATPASS samples. So there is no need to invoke the noble gae j

to iodine ratio in these cases.

The er,rond best measurement l

s would be one that yielded gross noble ran and gross iodine j

readings. This situation occurs in thst low range radiation 1

monitors which have individual noble gas and iodine channels.

Based upon the default mixture fractions, the release is (Q

apportioned among the fif teen nuclidas to arrive at isotopic release rates. Again, there is no need to use the noble gas to iodine ratio. It is used only in circumstances where either noble gas or i.odine measurements are not available, for example, I

when only noble gas or only iodine information is available.

5.8.4 There are some refinements and subtleties that the program user I

should be aware of.

The noble gas to iodine ratio changes with time because of radioactive decay. The RAC program has the ability to account for radioactive decay and to compute a decay corrected noble gas to iodine estio. As explained elsewhere in thiu manual, the program also corrects the Dose Rate Conversion Factor (DRCF) for decay of the isotopes in the mix. When performing dose projections several hours or more after the reactor has tripped, these two decay corrections can significantly alter the resultant projections. The computer operator is given the option cf whether or not to account for decay between reactor trip and dose projection.

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Radiological Controls Department 6610-PLN-4200.02

\\s_s Title Revision No.

TMI Emergency Dose calculation Manual (EDCM) 1 5.8.5 For dose projections based upon RMS readings, the decay correction to the noble gas to iodine ratio ie straightforward.

A def ault mixture of the fif teen noble gases and lodines is selected based upon an assessment of core damage. If the computer operator elects to decay the mix, he is prompted for the decay time between reactor trip and dose projection. All fifteen isotopes are decayed by the standard exponential decay law and the noble gases and iodines are totaled separately so that their ratio at dose projection time can be calculated.

(Ingrowth of Kanon isotopes from decay of iodine is accounted for.) The decay adjusted ratio can then be used to fill in the missing noble gas or iodine information, as explsined above.

5.8.6 When iodine samples are taken at the KAP-5 stations a two step decay process is used. As above, a default mixture is chosen, based upon the NRC core damagw classification. For a dose calculation based upon a radiciodine processor sample, if decay correction is desired, the user is prompted for two docay time intervalo 1.

The time between sampling and dose projection

'~'g 2.

The time between reactor trip and dose projection

\\~s/

Sample results from the radiochemistry lab are reported as of the sample collection time. When significant time has elapsed between sampling and dose projection, the results should be decayed from sampling time to dose projection time.

In order to compute the noble gas portion of the source term, the def ault mix in first decayed from reactor trip time to dose projection time.

The noble gas to iodine ratio is computed for the decayed default mix. This ratio, along with the gross radiciodine sample result, is used to compute a gross noble gas source term.

Isotopic source terms are calculated from the decayed mixture noble gas fractions. Note that the final source term is a combination of noble gases from a default mix and radioiodines from a sample.

Each has been decayed to the dose projection time.

A word of caution should be added at this point. The iodine released in certain types-of accidents may be reduced by various i

chemical and physical processes such as iodine plateout or formation of water soluble iodide salts. The noble gas to iodine ratio, as calculated above, may not account for this lodine reduction. As a consequence, the ratio, based upon the def ault mix, may be too low. This creates the potential for underestimating the noble gas portion of the source term.

RAC

]

personnel should be aware of this possibility. A comparison of field team data and the source term dose projections would reveal agreement for thyroid doses, but not for whole body doses.

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Radiological Controls Department 6610-PLN-4200.02 k _,

Title Revision No.

TMI Emeroency Dose Calculation Manual (EDCM) 1 5.9 Effluent Release Flow Rates - Flow rates for effluent releases to the environment are divided into four categories:

1.

Normal ventilation flow rates.

2.

Reactor Building leakage flow rates.

3.

Adjacent momentum plume rise (station vent and reactor purge i

concurrently releasing).

4.

Flow rates for OTso tube rupture release directly to atmosphere.

Buoyant plume rise source term calculation using RMG-26 or RMG-27 source term calculation using a contingsney calculation These flow rates for accident source terms to the environment are calculated as follows:

5.9.1 THI-1 Normal Ventilation Flow Rates O)

The Unit 1 RAC Program provides the option to use the actual 8

\\s_ /

ventilation flow rates as read frca the flow recorders or to use default flow rate (s).

Each normal plant flow path has predetermined flow rate ranges, and assigned flow recorders as follows:

1.

Reactor Building Purge

-FR909 0-20,000 CFM; Low Range

-TR148B 0-50,000 CFMJ High Range 2.

Reactor Building Purge and Make-up Exhaust

-FR148A 0-50,000 CFM 3.

Reactor Building Hydrogen Purge System

-FI282 5-50 CFM t

-FI283 20-200 CFM

-FI284 100-1000 CFM

-Total 5-1250 CFM 4.

Kidney Filter System

/'~'\\

-AHE-101, AH-F-12 P/I filters 20,200 SCFM-52.0 2042c

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Radiological controls Department s610-PLN-4200.02 Title Revision No.

TMI Emergency Dorg Calculation Manual'(EDCH) 1 5.

Auxiliary Building Exhaust-

-FR150 100,000 CFM 6.

Fuel-Handling Building Exhaust

-FR149 0-50,000 CFM 7.

Auxiliary and Fuel Handling Building Exhausts

-FR-151 0-150,000 CFM 8.-

Condenser off-Cas Exhaust

-RMR15 Recorder FT-lll3 Ch. A 0-200'CFM 9.

EST Fuel Handling Building Exhaust

-No Flow Recorder at this time 0-8000 CFM Def ault v.nlues are used in the RAC Program when a small value or an unknown value is required as input to a dose projection. The default values ares s,

5000 CFM Reactor Building Purgo

- Reactor Building Purge and Make-up Exhaust Auxiliary Building Exhaust Fuel Handling Exhaust Auxiliary and Fuel Handling Building Exhausts.

40 CFM Condenser Off-Gas 7000 CFM - ESF FHB Exhaust These default values allow the user to continue with dose projections even though a value is small or unknown. Therefore, once the dose projection is complete, the results may be ratioed up or down depending on the situation. For example,Jif the default value of 5000 CFM was used for a Reactor Building Purge and subsequently a Tech. Functions calculation was performed incicating 1000 CFM flow. The dose projection could be ratioed down by one fifth (1/5). Therefore, a dose projection of 10 mrom-would then be approxLmately 2 mram based on the reduced flow calculation, realizing that the I/O will also be affected by reducing flow.

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Radiological controls Department 6610-PLN-4200.02

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Title Revision No.

THI Emeroency Dose Calculation Manual (EDCM) 1 5.9.2 Reacter Building Leakage Flow Rate Another section of this program calculates a leakage flow rate out of the Reactor Building bassd on Reactor Building pressure.

The Reactor Building pressure indicator ic PT-291, 0-100 poig, located on control room panel CR.

The leakage out of the Reactor Building is based on the amount of pressure in the Reactor Building "ith all penetrations closed. The following equation ir used to calculate the Reactor Building Leak Rates L *SQRT(P,/P )

L,

=

a a

Reactor Building Leak Rate in CFM where:

L,

=

Maximum allowable integrated leakage rate at L

=

3 Pa La = 6.14 CFM Peak Reactor Building internal pressure at P

=

a design basis accident, P3= 50.6 peig 4

Actual Reactor Building internal pressure in P,

=

Psig

[

Therefore, the maxianum leakage allowed at a design basis accident (j

pressure of 50.6 peig is 6.14 CFM.

Leak rates at 0-60 peig can be calculated from the above formula. The default value in this subroutine is 50.6 psig. A graphic representation follows in Figure 5.9-1.

5.9.3 Adjacent Momentum Plume Rise (Station Vent and Reactor Purge)

For an isolated star.k, either the station vent or the Reactor Purge, the stack gas exit velocity can be calculated from the l

flow rate according to the following formula w=h (1) 2 where w = stack gas exit velocity V = flow rate or volume flux r = radius of stack 54.0 2042c

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Radiological rols Department 6610-PLN-4200.02

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Title Revision No.

TMT Emergency Does calculation Manyal !kDCM) 1 The station Vent and the Reactor Purge stack are situated close enough together that their plumes will mix as the plumes rise.

For two or more adjacent stacks that have different exit velocities, the effect of mixing on the exit velocities of non-buoyant plumes can be given by the following formulas E wv (2) w Ev Where w = exit velocity due to mixing If these adjacent stacks were modeled as a single stack, the radius of the stack would be given by U

(3) r=

nv At THI-1, the reactor building stack and station vent are g

adjacent stacks. For computing plume rise, the stack gas exit g ~s 1

velocity and stack radium were calculated according to Eq. (2)

\\s_ /

and (3) above. A comparison of the adjacent plume rise wit.h the plume rise from the individu. stacks is chown in Table 5.9-1.

5.9.4 Flow Rate Calculations for OTSG Tube Rupture Release Directly-to the Atmosphere (see Figure 5.9-2 and Figure 5.9-3).

TMT-1 has 22 main steam relief and atmospheric dump valves. Data on the valvos are presented in Table 2, which lists the valve identification number, function, 9anufacturer, pressure set point and flow rate. The set point pressures vary from 200 peig to 1092.5 peig, and the steam flow rate from 70,211 lbs/hr to 824,269 lbs/hr. Note that valves 4A&B are manually operated and do not have a set point pressure. These valves, MS-V-4A/B, can be operated from 0 to 100% open. The valve position openinge along with the secondary system pressure relate to a release flow and plume height. The percent open for these two (2) valves can be read at the center control panel under the turbine bypass dump controller for KS-V-4A/B from 0 - 100%.

Each of the 22 valves at THI-3 has a stack or vent where the steam is ejected into the atmosphere. The location of these stacks lo shown in Figure 5.9-2.

If a steam generator tube ruptures, each of the 22 valves and stacks acts as a throttle to limit the flow from the steam line to the atmosphere. When a valve opens, the flaw through it will be approximately equal to j

the rated flow, and the flow can be assumed to be approximately constant until the pressure in the steam line dtops to the point

[

whero the valve resest.9 For a stuck open valve, the pressure b.,0 2042c j

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Radiological Controls Department 6610-PLN-4200.02 1

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Title Revision No.

____ TMI Emergency Dose Calculation Manual (EDCM) 1 decreases rapidly with time, and the flow through the valve is only a small fraction of the rated flow. For either a normally

[

operating valve og stuck open valve, if the pressure and temperature in the steam line are known, the conditions just beyond tha etack exit can be satimated by assuming expansion of the steam to atmospheric pressure and temperature.

5.9.4.1 Buoyant Pluce Rise When the steam is released into the atmosphere, the rise of the steam plume is initially centrolled by its velocity, temperature and cross-sectional area. Depending on these variables and atmospheric cor.ditions, the plume rise can vary from hundreds to thousands of feet.

Plume rise le a very important f actor in determining maximum ground level doses. For a PWR, plume rise can increase the effective stack height by a factor of 5 to 50.

Since maximum ground level dose is roughly proportional to the inverse square of the effective stack neight, a plume rise of 200 feet, for eFample, gives a ground level Concentration 100 timos higher than that from a plume rise of 2000 feet.

I Modeling of plume rise begins with modeling the steam condition l

g'~'N at the valve inlet. Table d.9-3 outlines the calculational steps

(

)

required to compute buoyant plume rise, beginning with the valve I

inlet. The far left wide of the table identifies the area for s_-

l which the calculation applies: valve inlet, top of stack, jet l

origin, and plume rise. For each area, several quantities must i

be computed from various inputs, and these are also identified in the table. Buoyant plume rise was calculated according to Briggs (1984). The details of all the calculations are discussed in the Environmental Controle document Potentially Buoyant Releases at TMI-1.

.---------...------...------------=====- =.-----...----

CAUTION: In highly stable atmospheric conditions, th6 presence of layers of different temperature air can cause thermal boundaries resistant to plume vertical travel. In some l

l conditione, a buoyant plume may penetrate these layers and not come down to the surface as predicted.

In other cases i

the plume may be unable to penetrate the layer and the effective stack height will be reduced to the height of the layer. This may cause ground concentrations to be higher and closer to the plant than predicted. In these conditions (i.e., highly stable meteorology with a buoyant plume) off-site monitoring will provide an indicatior of the magnitude of the effect.

It may also be possible to estimate the effect through visual observation of the plume.

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,M1 Radiological Controls Departrent 6610-PLN-4200.02 Title Revision Wo.

_ THI Emergene) Dose calculation Manual (EDCM) 1 5.9.4.2 Source Term Calculation Using RMG-26 or RMB-27 'see Figurs 5.9-3)

RMG-26 and RMG-27 are effective in calcul? ting a primary to secondary release source term direct to the atmosphere when 1.

Atmospheric Dump Values ( ADV) MSV-4A or KSV-48 are open f r m 0-1004, as indicated on control Room Panel "CC", and releasing radioactive steam to the environment, end/or, 2.

Emergency Pood Pump (EFP) relief valves, MSV-22h or MSV-22B are open and releasing radioactive steam to the environment, and/or, 3.

EFP is in operation and releasing radioactive steam from the EFP exhaust to the environment.

4.

Steam bypass dump to the condenser through HSV-8A/B.

When the user chooses a release from an OTAG tubo rupture directly to the atmosphere and is using RM-026 or RM-G27 readings, the calculation Steam Flow Computation is used to deternine a release flow rate, depending on which of the valves rs are open (Table 5.9-2). The mars flow rate from each open valca f

is added up to give a total flow rate to the er.vironment. A gN,j/

source term is calculated using the flow rates in CFH and the RM-G26/27 readings converted to concentration using the monitor efficiencies to give pC1/second.

s._....

NOTE Calculation of a source term using RMS (RMG-26/27) is dependent on the Atmospheric Dump Va'sves (ADV) status.

If the ADV is open, the calculation is appropriate.

If the ADV is closed but plant conditions (0730 leakrate and core damage) have not changed significantly, or there is other l

flow paht the monitors as noted above, then the use of a I

RMG-26/27 peak reading will be appropriate. If the ADV is closed and plant conditions have changed significantly, and there is no other source of flow downstream of MSV-27/B, then the contingsney calculation applies.

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""* r TMI Radiological Controls Department 6610-PLN-4200.02 Title g

Revision No.

TMI Emerooney Dose calculation Manual (EDCM) 1 5.9.4.3 source Term caleulation unir a contingenev calculation When the user performs a Co.tingency calc lation due to the lack of sample results or RM-G26/A7 readings, t ' a flow rate corresponding to +,he set point > + essure i used, if the valve operates normally. ThJ s flow rat.e is gi >n in Table 5.9-2.

If, however, the valve sticks open, and cha steam generator pressure is less than the set point pressure, then the flow rate is based on the tables supplied by the valvo manufacturer. These tables have been incorporated.*nto ths RAC cceputer code. The flow from all the valves is totalAed and modeled as a release to the atmosphere.

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Title Revision No.

1 1

TMI Emergency Dose calculation Manual (EDcM) l TABLE 5.9-1 l

f Adjaccnt Plume Rise at TMI-1 Reactor Bldg Stack and Station Vent Stack Actual Flow Characteristica RAC Model MIDAS

~

Stability ' Reac Bid Stack Statiott Stack Flow Stack Pluate Flow Stack P!um.

i Stack D~ meter Vent Diarneter Rate Diameter Rise Rate Diameter Rise (cfm)

(m)

(cfm)

(m)

(cfm)

(m)

(ft)

(cfm)

(m)

(ft) l 10,000 1.1 10,000 1.7 20.000 1.847 30.3 10,000 1.1 25.4 l

10,000 1.7 16.4 j

A)'

10.000 1.1 120,000 1.7 130.000 1.827 199.2 65,000 1.1 165.4 65,000 1.7 107.0 StdMe" 10,000 1.1 10,000 1.7 20.000 1.647 24.8 10,000 1.1 22.1 dk$ '.

10,000 1,7 16.5 (Claes.F);

'W 10,000 1,1 120,000 1.7 130.000 1.827 85.8 65,000 1.1 75.9-65,000 1.7 57.0

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TH! Emerovriey Dose Calculgtion Manual (CDCH) 1 TABLE 5.9*2 TMl1 STEAM GENERATOR RELIEF VALVES Velve Stack VaNo Vahe Discharge set Point StecA Nurn*>w Functbn MenutscPl.'w kee Pressure l Flow Rete iAWter (MS V)

(64, in )

(pn6g)

(bs/hr)

(anches) 17AC Ro w Vtwes, Bank i Drnow /

16 1050 792.617 10 02 Conse6deted 16AC Ae6er Veves. Bank 2 Druse /

16 1060 000.065 10 02 Consondeted ISA D Re t \\ Nos. Bank 3 Oreiser /

16 1000 814.960 10 02 Conso6detod 20A&D Row VaNos. Bank i Dressw /

16 10$0 792.617 10.02 Conso6deted 208&C ReW VaNos. Bank 4 Druser /

10 1002.5 624.289 to 02 Conso6detod 21A&D Re6ef vnNet, Druser /

3.97 1940 194.820 10.02 Smet Gelety r.onsondsied W

6eW Re64 Lonergan 6.38 200 70.211 1313 4

Emergency Foert Pump MB Safety VaNo. Emerg.

Lonergan 6.Se 220 76,795 13.13 (F W.P.T. Bleam Irdet) 4A&B RdetVakee(manuar) Fwher variab6e 1010 402.792 13.13 to Atmosphwe

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g THI Emeroenev Dose Calculation Manual (EDCM) l 1

l l

TABLE 6.9-3 Calculational Steps for Computmg Plume Rise Source Stop Quentity Computed input Values Needed of input Velve inlet i

Steam Aow rate thru veno pressure of steam operator Top 2

Pressure of steem et top

a. steem low rete step t of of stock, bWow chamfor b, Mtemal radius of sistk constant

Stack, (Kchokoonow)

e. entheapy econstant.

Below Chamfor 3

Spade volume of steam at pressure at top of stacA stop 2 top of stack. below chamlw Below 4

Temperature of steem at pressure at top of stack slop 2 Chamfor top of ud. beow chamfer Alto 6

Voocny of steem at top of

a. srm voksme of steam step 3 Jet stack, be*ow chamler

b. 8.w rete of steam step i Origin c etemal rows of stack constant 6

DensRy of steam at

a. typerature of steere step 4

)st onen (amb+ent pressure)

b. pressure of antpent air sconstanta Jet 7

Jet radius at ongM

a. der,sRy of strom step 6 Origin (Neocod for MCAS only; b, vgkWty of steam stop 6 used h MC but not

c. Ihw rete of steam step i resty needed) 6 Plume noe e, )st redkJs at or6pn step 7 Plume b dansky of steam at onen step 6

c. velocay of steem si myn stop 6 Rise

d. densRy of amtwent air

< constant.

e. wind speed at onen operstor f.

150'43' delta T operster

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O Title Radiological Controls Department 6610-PLN-4200.02 Revision No.

r TMI Emergency Dose calculatio. Manual,(EDCM) 1 FICVRE 5.9-1 TMI-1 RB LEAK RATE VS RB PRESSURE L

PSIG 60 la*

A IEAK RATE M M.HCPW j

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Title Revision No.

l TMI Emergency Dose Calcalation Manual (EDcH) 1 n.10 Two-Phase Steam Flow Determination - A t wo-phase (liquid and gas) release calculation was included for an OT80 tube rupture accident in response to INPo soEn 83-2 (Recommendation $12). INPO SOER 83-2 " Steam Generator Tube Ruptures" was developed based upon the steam generator tube rupture events at R. E. Cinna, Oconeo and Rancho seco. Recommendation #12 states

Emergency Plan Implementation Procedures should... ensure that estimates of doses can be made for two-phase or liquid releases through the steam generator oafety relief valves." GPUN Corporation is required to respond to all soER recommendations. The calculational method used to Laplement this recomoendation is based upon the assumptions that the valve inlet fluid condition is either pure liquid or steam (as indicated by the 078G wide range level instrumentation) and following discharge, the steam fraction is described by assuming that there is no change in total energy content. If the OTSO wide range level instrument is indicating that the valve inlet fluid condition is pure liquid, greater than 600 inches, and the fluid is near saturation for the pressure and temperature, then the fraction of gas vapor present in the release is a function of the OTSG pressure as indicated on the PCL panel, PI950A and 951A, or the console center, SPGh PT1 and 2 or SPGB PT 1 and 2.

5.10.1 The program determines a two-phase correction factor (Tfcf) which is a function of OTs0 pressure in psia. This factor is only calculated if the OTSG water level is indicating a liquid release s s (greater than 600 inenes on the wide range level instrument k,

reeding). The correction factors are used to account for the i

s I

radiciodine that would remain in the liquid portion of the resultant two-phase release to the environment.

Upon input of the OTSG pressure in Psia the code selects a correction factor which is subsequently used in the radiciodine source term equation to correct the radioiodine source term.

RIST = D1

• D2

• D3

• RI/100

• 63.09

• 1/Tfcf NOTE:

Increasing 0750 A/B water level will possibly help cut down the release of radiolodine due to the partitioning effect of the iodine in water. Increasing OTs0 level should be discussed with the Emergency Director as a means of reducing offsite doses, j

1 65.0 2042C 1

MUClear

,,3 Radiolooleal controls Department 6610-PLN-4200.02 Title Revision No.

TM! Emeroeney Dose Calculation Manual (EDCM) 1 5.11 Source Term Filtration - The THI-1 RAC e'o? ram provides the option to include or disregard reduction of source term through filtration. The source term reduction is applied to radioiodine species only.

5.11.1 Radioiodine normally exists in chemical forms which are highly reactive. They readily adsorb onto surfaces and can be scrubbed chemically from the atmosphere. The radiciodine renoval methods available in the plant by design are the charcoal filter banks in the ventilation eystem. 79 applied in the RAC tape, anytime a sample is obtained upstream of a charcoal filter a filter reduction can be applied. Anytime a def ault source term is used, a filter reduction and/or building spray reduction can be applied.

5.11.2 In any case where the RAC program will apply the source term reductions a prompt is proviood by the program.

In a general form, the program will promph "ARE T:3 CHARCOAL FILTERS OPERATIONAL" or "!S IHE REACTOR BUILDING SPRAY ACTIVATED".

Answering

'Y' to these questions will apply the source term reduction factors associated with each system. The charcoal filters are generally better than 95% efficient for ree: val of radiolodine species. However, the RAC program tak&s a conservative approach and assumes a 90% reduction from fully operational charcoal filters. Note that the prompt for either is N

a "yes" or "no" answer.

If the charecals are known to be degraded (for exaaple, due to meisture) but still partially functioning, this is accounted for in the RAC model. The filters are either functioning at the full capacity - 90% or from 0 - 90%

due to dsgrad& tion.

5.11.3 Application of the filter reduction is available in Reactor Building and Aux /FHB releases. Reactor Building releases monitored off of RMA-2, the CATPAS8 system and contingency def ault source terms generated from a spent fuel accident or LOCA in the Reactor Building, including RM-L1 readings, RCS sample results, cladding damage, or fuel melting scenarios all provide the opportunity to apply the filtration to the iodine source term. Normally, application of the filter fraction will simply reduce the radiciodine source term by a factor of ten. However, if the source term is generated from a gas channel reading, the iodina will be reduced but the noble gas will be increased due to a constant meter reading and a change in the isotopic ratios.

Application of the filtor reduction for iodines in the Aux /FHB is available for source terme generated from RMA-4, RMA-6, a 4 the contingency def ault source terms from a sample result, damaged fuel rods, a fuel cask, or a waste gas release.

It is Lmportant to remember that filter application can significantly change the noble gas to iodine ratios.

66.0 2042C

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NI gg Radiological controls Department 6610-PLN-4200.02 Title Revision No.

)

b TN! Emergency Do6e Calculation Manual (3DCN)

J

- i Ar.,4tmo a filter correction f actor is applied to the iodines downstream of the filters, the noble gas source term must be increased, while iodines obtained upstream of a filter will reduce the iodine source term without changing the noble gases.

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Radiologi 41 Controls Department 6610*PLN-4200.02 I'

Title Revision No.

1

, TM1 Emergency Dose Calculation Manual (EDCM) t 5.12 Heteorology Inquiry - Upon initiation of the metectological data input

{

section of the RAC program, the computer places a telephone call to the IRH PC located at the base of the TM1 met tower and requests f rom it the most l

recent 15 minute average of the met data stored in the met tower PC.

The RAC program is able to continue if the telephone call is not completed, the met tower phone is busy, or the met tower fails to respond after the call has begun.

If any of the above conditions occur, the data is marked as i

missing by the RAC program. Af ter the call is terminated, a new screen is presented to the user listing the collected data in a tabular format and l

asking the operator f or the wind speed. If the "A" sensor value is

[

non-missing, the operator is allowed to default to it.

If the "A" sensor value is missing and the "B" sensor value is non-missing, the operator it allowed to default to the "B" sensor value.

If both "A" and "B" sensor values are missing, no default value is presented or allowed. To obtain i

the def ault value the operator presses the Return key while the cursor is on the first character of the input field. The operator is free to enter his own value.

j Af ter the Return key is pressed, the RAC program subjects the inputted value to limit checking. The limits ares Lower Limit Upper Limit

\\

Wind Speed 0.5 99 s_/

Wind Direction 0

360 Delta T

-30

+30 If the entered value is lower than the lower ILmit or higher than the upper limit, the operator is reqvested to input a new value. An exception is the i

wind speed value. If the entered value is less than 0.5 mph, the RAC program uses 0.5 mph and informs the user of this fact.

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Radiologiesi controls Department 6610-PLN-4200.02 i

Title Revision No.

TM! Emergency Dome calculation Manual (EDCH) 1 5.13 Diepersion Model - The THI-1 RAC model computes both Whole Body Dose using a finite come model and thyroid dose using a semi-infinite model. Many l

subroutines are called by both models. The results of the two models are presented on a single output page.

5.13.1 Finite (Whole Body) Dose Model The THI-1 RAC model calculates external whole body gassna dose rate using a finite model for both ground and elevated releases.

The finite gamma dose algorithm is licensed f rom Dr. John Hamawi i

of Intech Engineering through Pickard, Lowe & Garrick, Inc.

(Dr. Hamawi was the author of the dose integral routine listed in Appendix F of Reg. Guide 1.109). The dose is computed by I

multiplying the dose rate by the expected duration of release.

The finite gamma dose algorithm in the THI-1 RAC model has the same structure as Pickard, Lowe & Garrick's MIDAS finite gamma dose algorithm. The basis for the algorithm to a four dimensional array of finite gamma factors. These finite gamma factors are pre-computed three dimensional numerical integrations which appear in the theory of the finite cloud model and represent the spatial distribution of the radioactive material in the finite plume. These f actors depend upon the plume dimensions f

at the downwind distance of interest, the crosswind distance, the plume elevation and the average gemma energy of the nuclide mix in the cloud. They are tomettres referred to as " gamma X/Q" i; the literature although they are not derived from typical X/Q ealculations. The finite gamma factors in the array correspond to 28 downwind distances, 6 crosswind distances, 6 heights above ground, and 6 energy groups. Specifically, the downwind distances are 400, 500, 600, 700, 800, 900, 1100, 1250, 1500, 1750, 2000, 2250, 2500, 3000, 3500, 4000, 4500, 5000, 5500, 6000, 6500, 7000, 7500, 8000, 9000, 10000, 15000, and 20000 meters.

The 6 crosswind distances ares 0, 50, 100, 150, 250 and 500 meters. The 6 heights above ground ares 0, 30, 60, 100, 150, l

and 300 meters. The 6 energy groupe are

.032,.081,.15,

.25,

.53, and 1.0 MeV.

The abundances of the noble gases for the six energy groups were taken from MIDAS.

For effective release heights other than the 6 fixed heights, the finite gamma f actors are extrapolated to that height by the subroutine (Interpolate). For downwind distances other than the

~

28 fixed downwind distances, the finite gamme factor of the nearest fixed distance is assigned to that distance, i.e.,

no i'

horizontal interpolation is done, as is consistent with MIDAS.

The TMI-1 RAC model explicitly includes the contribution of I-131, I-132, I-133, I-134, and I-135 to the external whole body gamma dose. This method of handling the contribution from the radioiodines is note accurate than the method used in MIDAS. The

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MUClS8r

""*'* r Radiological controls Departn.ent 6610-PLN-4200.02 Revision No$

i Title TMt Emerceney Dose calculation Manual (EDcH) l 1

abundances of the radioiodines were taken from the Radioactive

. 11.

All r'adionuclides are Decay Da6a Tables, D.C. Kocher, decayed during plume travel.

5.13.2 semi-Infinite Dose Model The THI-1 RAC model calculates the thyroid dose rate due to inhalation of 2-131, 2-132, 1-133 1-134, and 1-135.

The thyroid dose rate is proportional to X/r The constant of proportionality is the p' 1 child breathing rate and the child inhalation do-

.no program uses the child breathing rate of 0.4:

. (trom 43hle E*5, Reg. Guide 1.109) and the child inhalation dose f actors are from Table E-9, Reg.

Guide 1.109 to compute the dose rate conversion factors. The dose is computed by multiplying the dose rate by the expected duration of release.

The radioiodines are decayed during plume travel time. The decay constants for 2-131 thrcugh 1-135 are from the Radiological Health Hananoak.

5.1.3. 2.1 x/o calculationt

[

The basis for the I/O calculation is the caussian diffusion

\\s_-

equation and a 10 x 7 array of sigma y's and sigma s's.

The array of valuse correspond to sigma y's and sigra s's for 7 stabili y classes and at 10 fixed downwind distances. For distances other than the fixed downwind distances, the F.!gma y's and sigma t's are linearly interpolated before I/Q is computed for that distance. The ten fixed distances ares 200, 500, 1000, 2000, 3000, 6000, 10000, 30000, 50000, and 80000 meters.

5.13.2.2 compute Building Effect Returns one of seven pre-computed virtual source diets.nces, depending on stability class. The virtual source distances fer each of the seven stability categories are 209,209,209,308,465,770 and 1254 meters, respectively. These values were computed based on the cross-sectional area of the nearest large building. Building wake effects are simulated by adding the virtual source distance, for a particular stability class, to the actual downwind distance for the purpose of computing X/Q. For example, suppose we wanted to know X/Q without building wake effects at 800 meters downwind with stability class D.

X/Q would then be computed at 800 meters downwind. With building wake effects, X/Q would be computed at 1108 meters downwind (800 + 308) Thus building wake effect is simulated by computing X/Q at a distance greater than the actual downwind distance and is called only for ground level pertion of release.

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MUClG8r

""***r yx, Title Radiological controls Department 6610-PLN-4200.02 Revision No.

TMI Emergency Dose calculation Monual (EDCM) 1 5.13.3 Subroutines Used by Both Finite and Semi-infinite Model -

5.13.3.1 Compute TMI-1 Emergency Action Level. Declares the emergency action level f rom highest dose whether whole body ce thyroid.

Emer. Action tevel Maximum Dese Rate (mrom/hr) within 10 miles WHOLE BODY TNYROID Hone 0 $ dose rate <

10 and 0 $ does rate <

50 Alert 10 $ dose rate <

50 or 50 $ dose rate <

250 Site Area Emergency 50 $ dose rate < 1000 or 250 $ dose rate < 5000 General Emergency dose rate 1 1000 dose rate 1 5000 The subroutine computes the EAL for whole body and thyroid dose and then reports the more severe of the two.

5.13.3.2 compute site _ Boundary The whole body and thyroid doses are computed at the site boundary. The distance to the eits boandary varies with the compass sector that the wind is blowing to.

This routine returns this distance in meters.

5.13.3.3 compute Terrain Factor Computes terrain height in meters for a given downwind distance.

At downwind distances other than those in the subroutine, terrain height is computed by linear interpolation, except at distances closer.i.sn 610 meters. Between the plant and 610 meters downwirs. the terrain height is set equal to the terrain height at 610 meters.

Terrain further from the plant is never lower than terrain closer to the plant due to mathematical approximations.

5.13.3.4 compute stability class As measured by the THI Meteorological Tower from the 150 ft minus 33 f t temperature dif ference. Table 5.13-1 relates the temperature difference to the stability class. The equivalent temperature dif ference per 100 f t is shown in the last column of the table. With regard to the temperature dif ference, MIDAS can be confusing because MIDAS expects the input in degrees per 117 feet, but prints it out in degrees per 100 ft.

stability class is datermined by the measured temperature lapse rate per Reg.

Guide 1.21.

_s 71.0 2042C

i (UCIS8r

""*6" Radiological controls Department 6610-PLN-4200.02 i

Revision No.

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Title TMI Emergency Dose Calculation Manual (RDCM) i t

5.13.3.5 Ad$ust Wind speed Adjusts wind speed from the anemometer height to the release j

height. The wind speed is adjusted according to the following i

equation s u = u,(h/h,)P where the subscript "0" denotes the anemometer height and "u" and "h" are the wind speed and height above ground, respectively.

The exponent p is a function of stability 0.25, 0.03 and 0.50.

I for unstable, neutral and stable cases, respectively. If the adjusted witad speed is less than 0.5 mph, the adjusted wind speed l

is set equal to 0.5 mph.

5.13.3.6 Compute Enit velocity Computes exit velocity of the release. satorial in feet per second by dividing cubic feet per minute by the stack l

cross-sectional area.

5.13.3.7 Compute Plume Rise computes the plume rise in meters for the elevated portion of a split wake release. Two formulas are used to calculate the plume rise; for unstable and neutral conditions jet plume rise, I

momentum dominated, is calculated from Briggs' Plume Rise, Eq.

4.33; for stable stability, it is calculated using Eq. 4.28 ftom Briggs' Plume Rise.

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72.0 2042C i

4 MUClear Radiological Controls Department 6610-PLN-4200.02 Title Revision No.

TM! Emergency Dose Calculation Manual (IDCM) 1 5.13.3.8 Compute Entrainment Factor Computes entrainment factor for wake split flows. A mixed mode release is assumed when: (1) the release point is at the level of or above adjacent solid structures but lower than elevated release points. (2) the ratio of plume exit velocity to horisontal wind speed is between one and five, specifically, the entrainment factor, E, is computed according to the following g

iormulas:

1.0 for w,/u le 1 E =

g

2. 58 - 1. 58 (w,/u ) f or i gt w,/u le 1. 5 E =

g Es = 0. 30 - 0. 06 ( w,/u ) f or 1. 5 yt w,/u le 5. 0 0 for w,/u gt 5.0 E =

g whors w is the stack gas exit velocity and u is the wind speed e

at stack height in miles per hour.

Note that the entrainmer.t factor does not address the case of two adjacent plumes mixing with each other, as would be the case in THI-1, where it is possible for a clean plume and a contaminated plume to be emitted from adjacent but separate stacks. These plumes are examples of co-located adjacent jets; little is known about the modeling of co-located adjacent jets such as the ones at TMI-1.

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gg Radiological controls Department 6610-PLN-4200.02 l

Title Revision No.

A m! Emergency Dose calculation Manual (EDcM) 1 TABLE 5.13-1

)

classification of Atmospheric stability stability Pasquill Delta T Delta T classification categories 1150' - 33')

(*F/100')

{

t 6

(*F)

('F)

Extremely Unstable A

< -1.22

<--1.04 t

Moderately Unstable a

3 -1.22 to < -1.09 1 -1.04 to < -0.93 slightly Unstable C

3 -1.09 to < -0.96 3 -0.93 to < -0.82 Neutral D

> -0.96 to < -0.32

> -0.82 to -0.27 j

slightly stable E

> -0.32 to < +0.96

> -0.27 to < +0.82 I

Moderately stable r

1 +0.96 to < +2.56 3 +0.82 to < +2.19 i

Extremely Stable Q

> +2.56

+2.19

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Title Radiological Controls Department 6610-PLN-4200.02 f

Revision No.

f TMI Emergency Dose Calculation Manual (EDCM) 1 i

5.14 offsite Air sample Analysis

{

5.14.1 Introduction The "Of f aite Air Sample Analysis" portion of the RAC code is used in conjunction w!th results provided from field teams to assess thyroid doce commitment.

The method involves collection of an air sample using a low flow i

(about 50 LPM) sampler with bo,n a particulate filter and an iodine adsorber cartridge. The flow rate of the sampler, the duration of sample collection, the background of the frisker used to count the sample, the gross counts on the particulate filter, and the gross counts on the lodine cartridge are called into the RAC or EACC from the field teams. The RAC or EACC staff then uses the RAC code to estimate the of f site dose commitnent based on the sample.

5.14.2 Assumption!

A calibrated face loaded iodine cartridge was obtained and was used to determine the actual efficiency of a Eberline E140N with a HP-210/260 type probe to be used for counting in the field.

The results of several tests on combinations of different probes and ratemeters showed a consistent 0.0039 (0.39%) counting efficiency.

(Reference 7.7, 7.10, 7.11).

Since 1-131 has a fairly strong beta (0.6 MeV max.), the usual particulate filter counting efficioney of 0.1 (10%) is used. The collection efficiency for both filters for these calculations are assumed to be 1.0.

5.14.3 calculation The method first calculates the not counts per minute for tha particulate and iodine cartridge. Then, using the given efficiencies separately, it cateulates the air concentration of gaseous and particulate iodines. These are then combined for a total air concentration. A child breathing rate and dose conversion f actor is then applied along with the estimated duration of exposure to obtair. the offsite dose commitment.

Since the plant RAC code normally accounts for five different iodine isotopes, the dose cot. version f actor (DCF) used is a weighted average of the child DCFs based on the relative i

abundances of the five isotopes at damage classes of on6 and five with 100 minutes decay. This accounts for counts on the samples which will be caused by isotopes other than 1-131.

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Radiological Controls Department 66_10-PLN-4200.02

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Title Revision No.

THI En.drgency Dose Calculation Manual (EDCM) 1 5.14.4 Examples Given an of fsite air sample was takan with the following results:

Background = 100 cpm gross cartridge countrate = 200 cpm gross particulate countrate = 200 cpm flow rate through sampler = 50 LPH sample duration a 10 min.

exposure duration = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> DRCF m-4.0E8 mram/hr pCL/cc The RAC program follows the logic below to calculate an of f-site thyroid dose commitment for this sample.

a.

not particulate countrate = 200 - 100 = 100 cpm b.

not particulate activity = 100/.1 = 1000 dpa c.

not cartridge count rate = 200 - 100 = 100 cpm d.

not cartridge activity = 100/0.0039 = 25600 dpm g'~'

k_/

e.

total activity in sample = 1000 + 25600 = 26600 dpm s

f.

total microcuries = 26600/2.2286 = 0.012 pCi g.

sample volume = 50

• 1000

• 10 = 5E5 cc h.

air concentration = 0.012/5E6 = 2.4E-8 pCi/cc 1.

dose commitment = 2.4E-8pci/cc*1 hour *4E8 mrom/hr = 9.6 mrom pCL/cc 5.14.5 The items listed below appear while performing this section of the RAC program. Once all input is entered, the resultant tnyroid dose commitment is displayed in mrom, or mrom/hr if a duration of one (1) hour is entered.

OFFSITE AIR SAMPLE ANALYSIS EKAMPLE SAMPLE TIME (military clock)

SAMPLE LOCATION FIELD TEAM DESIONATION BACKGROVND COUNTRATE 100 Cpm GROSS PARTICULATE COUNTRATE 200 Cpm s

76.0 2042C i

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i L UClear Radiological controls Department 6610-PLN-4200.02 Revision No.

Title TNT Emergency Dose Calculation Manual (EDCH) 1 1

CROSS CHARCOAL COUNTRATE 200 cpm TLOWRATE THROUGH SAMPLER 50 LPH DURATION OP SAMPLE COLLECTION 10 HIN t

EXPECTED DURATION OF RELEASE 1 HOURS For Dose Rate Enter Duration of One Hour THYROID DOSE CONHITMENT 9.663+00 MREM (using weighted child DRCF) i d

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__ Title Revision No.

TMI Emerger'cy Dose calculation Manual (EDCH) 1 i

s 5.16 Liquid Release calculation - In this section of the program calculations are performed for liquid source term determination, (see Figures $.15-1, 6.1$-2, and Tables 6.3 bl, 5.15-2) MPC's in the river, travel time to 7

downstream users, and ingestion dose consnitment calculatione. The methods l

used to perform the calculations are as follows:

1.

The concentrations of the liquid effluents are determined by one of i

the following methods. Each method uses only the four usual lodine isotopes and Co-134 and co-137. All selections are sienu driven.

j a.

Normal Miscellaneous Liquids. An ' average' isotopic content for miscellaneous plant liquid wastes is called if this option is selected. The isotopic concentrations used a.'

typical values for default use only. The values cannot be changed.

If other i

isotopics are known to be present, other options should be used.

l The discharge rate is used with the concentrations to calculate l

the source term.

f b.

Known Isotopic Concentrations.

If the concentrations of the five i

lodines and two cesiums are known from gamma analysis of ther i

liquid, then the option for use of actual isotopic concentrations l

can be used. The program will prompt the user for each isotope.

The discharge rate is prompted for in order to calculate the O

source term from the concentration.

c.

primary to Secondary Leakage with Secondary Liquid Release.

If primary to secondary leakage occurs and the secondary liquids are released to the river, this option is appropriate if the actual

{

concentrations are not known. The program will prompt for the l'

RCS temperature and pressure in order to calculate the NRC damage class and associated isotopic percentages. The isotope percentagee for the five iodines are normalized to a sum of 1 and the cesium activities are calculated using ratios for the co-134 and Cs-137 to the 1-131 based on the NRC damage class. The t

ration are 0.0075 for Co-137 and 0.036 for co-134. These are assumed to represent a damage class 7 or greater accident. To adjust for other damage classas, the percent of core matrix activity given in 5.3,3 is used to adjust the abundance of the cesiums. Thus at high damage classes the full ratio is used, i

while at lower damage classes, the ratio may be adjusted down by

.5,

.1 or 0.

The leakrate and duration of primary to secondary is used to

[

estimate the maximum activity in the secondary system. The discharge rate of secondary to the river is used to estimate the source term.

d.

RCS Direct to River. The damage class and isotopic percentages

[

are used in a manner similar to that for primary to secondary leakage. The isotopic concentrations and the discharge rate are

[

used to calculate the source term.

^

i 78.0 2042C

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t 7,3 Radiological controls Department 6610-PLN-4200.02 Title Revision No.

TM! Emergency Dose Calculation Manual (EDCM) 1 l

The activity in the turbine building sump following a primary to e.

secondary leak is calculated using the isotopic percentages based l

on damage class and the count rate of RML 10.

The response of RML 10 in cpm /pci/mi-is used to estimate the total concentration and the isotopic percentages are used to calculate individual concentrations. The discharge rate from the turbine building sump is then used to calculate the source term.

2.

The dilution in the river is calculated by first obtaining-the river flow rate and inputting the value in the program.- Instructions are i

provided for obtaintag the flow rate. The river flow. rate in then used along with the discharge flow rate to calculate the concentration in the river. The concentration in the river is then divided by the water MPC to determine the MPCs in the river to downstream users.

3.

The river concentration is used along with the total discharge time, to calculate the dose commitment to an individual from drinking the river water f rom one of the downstream intakes. The river concentration is multiplied by the duration of the reloaea, the usage f actor, the ingestion dose commitment f actor for inf ants, and the infant usage factor (3301/yr) to obtain an estimated dose commitment f or the downstream drinker. The infant dose is used because the product of the usage factor and DCF shows that the infant is the maximum age group for~ all seven isotopes.

I 4.

A fiume arrival time is estimated for each known downstream user and printed out. The river volume flow is used in an algorithm based on a model derived from river dye dilution and flow studies conducted from the TM! discharge.

i 5.

If the concentration of any nuclide exceeds le-6 in the river, downstream users must be informed and recommended to curtail usage.

j 6.

If any MPC fraction exceeds 500 MPCs, the NRC must be notified within l

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per 10 CFR 20.403.

If the NPCs exceed 5000, immediato notification of the NRC is required.

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REVISION g Unit 2 Liquid Pathways Control Control + I i

Building Senice Sump Area Sump j Turbine

. Gen. 'A' Diesel 3

i

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=

l Sump j

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81.0 2042c

IUClear rol Padiological _ Controls Department 6610-PLN-4200.02

yggi, Revision No.

TMf Emergency Dose Calculation Manual (EDCM) i TABLE 5.15-1 i

a THI-2 Sump Capacity j

Sump Total Capacity Gallons / inch j

Turbine Bldg. Sump 1346 gald 22.43 l

Circulating Water Pump House Sump 572 gals 10.59 Control Bldg. Area sump 718 gals 9.96 Tendon Access calley Sump 538 gals 9.96 Control to Service Bldg. Sump 1346 gals 22.43 Emergency Diesel Generator Sump A/B wet 837 gals 9.96 A/B dry 1200 gale 14.29 Chlorinator House sump Water Treatment Sump 1615 gals 22.43 Air Intake Tunnel l

Normal Sump 700 gals.

Emergency sump 100,000 gals 766.00 Condensate Polisher Sump 2617 gals 62.31 Sludge Collection Sump 1106 gals 26.33 Heater Drain Sump Solid Waste Staging Facility Sump 1476 gals 24..

Aux. Bldg. Sump 10,102 gals

-202 Decay Heat Vault Sump 478.5 gals or 957 gal? (total)

- Building Spray Vault Sump 478.5 gale or 957 gals (total)

- 10 l

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2042C v

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d W UV yg Radiolegical controls Department 6610-PLN-4200.02 Title Revision No.

TMI_ Emergency Dose Calculation Manual (20CH) 1 TABLE 5.15-2 THI-1 Sump /Tenk Capacities s, ump capacity (Gallone) um Turbine Building sump (TBS) 10,000 Auxil'.sey Building sump (abs) 10,000 R 6 * *. Building Sump (RBs) 10,000 Intermediate Building sump West 1,000 Tendon Access Gallery sump 1,000 Intermediate Building snap East 1,000 Auxiliary Boiler sump 2,000 Powdex Sump 40,000 Industrial Waste Treatment System Sump (!WTS) 300,000 Industrial Waste Filtration System Sump (IWFS) 80,000 Tanks THI-2 Condensate B Tank 250,000 TMI-1 OTSG A or B (secondary) 25,000 THI-1 WECST A or B 8,000 Neutraliser Tank 100,000 BWST 350,000 condensatt Tank A/B 265,000 1

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luCISSF exx Radiological controls Deparg gnt 6610-PLN-4200.02 Title R**I'i " N

• TMI._ Emergency Dose Calculation Manual (EDCM) 1 5.16 Protective Action Recommendation Logie - The Lop W Diagram is designed to enable the user to develop protective actions based upon plant conditions, release duration and dose assessmente. The logic is diagramed in Figure 5.16-1 f or TM1-1 and Figure 5.16-2 for 7NI-2.

For the most up to I

date PAR logie diagrams see procedure 9471-1MP-1300 02.

t 1

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84.0 2042c

_ _ _.., _ _ _ - _., _ -.. - - ~.

ENuclear Radiological Controls Department 6610-PLN-4200.02 Revision No.

l THI Emergency Dose calculation Manual (EDCH) 1 TM IE.E M 1AET 1 FIGURE 5.16-1 mm (Example)

N IF PN ACTgp arremem**TEkt (PW WERAL DERIENCY 18 MCLAKD

$!TE AfEA EKnM)CT g arn.ars b

sysv M umC mAm m a0 Taum Paa unic meAn TD Mar FMPALTB FM A PRMAL EMBL PAA VITHIN 45 MNJTES. F A ME!SEM (MM M E!!JmfB MPWETM CAN DCT BE M43E VIT)Si 15 >IW"U TMN ACTDi EC90ENDAfDi gg urm&*WT VITH MIC RADANCE, A8 A MINDEM Kttpece SPELTENHO SJT TD 4 MILES F W ImER M II WAMAffft3 RAstus ANS fL 90LES IIMMMB.

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DELTUDIG FUt t 18,1 AMN 7

D AMS S MILES ElMMME AND Ct3(TDAK A'er***WT 1

U 13 TM3E A 1.AME F13333t 13 EP4 PAG uNER LDET PEIRET DfVENTtRY IN T4 DC mwarn 3 PELECTD TD CistTADeeft AT)c3PM1 E ""W'1 1 M38 N EilBT calsTICENT TD EXIED EPA 5 MM 08 3 TWTMES l.

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85.0 2042c

1 AUCl98r

,,I RadioloGieel controls Department 6610.PLN-4290.02 Title Revision No.

TMI_F.merceney Dose calculation Manual (EDCM) 1 FIGURE is 16 1 (Cont'd)

(EEAMPLE)

PROTECTIVE ACTION RECOMMENDATIONS (PAR) LOGIC DIAGRAM NOTE la AS INDICATED BY ONE OF THE FOLLOWING 1.

CAT-PAS SAMPLE RESUI.TS OFs TOTAL NOBLE GAS CONCENTRATIOA > 2300 MICRO CI/CC TOTAL RADIO!ODINE CONCENTRATION > 420 MICRO CI/CC 2.

HIGH RANGE CONTAINMENT AREA MONITOR READINGS OFI RMG-22 OR RMG-23 > 400 R/HR (HIGH ALARM)

....................................................................~......

NOTE 2:

AS INDICATED SY ONE OF THE FOLLOWING 1.

RCS POST ACCIDENT SAMPLE ANALYSIS IhDICATES GREATER THAN NRC D WAGE CODE 2.

2.

RM SOFT'fhKE CODE CALCULATES GREATER THAN HRC DAMAGE C01)E 2.

(THIS IS BAEED ON RCS PRES $UPE AND THE AVERAGE OF TO FIVE HIGHEST INCORE THERMOCOUPLES. POINT C4006 ON COMPUTER) 3.

LETDOWN MONITOR READINGS RML-1 LOW AND RML.1 HIGH ARE OFFSCALE HIGH (THE ISOLATION INTERLOCKS MUST BE BYPASSED TO GET THIS READING).

[

NOTE 31 AS INDICATED BY ONE OF THE FOLLOWING 1. REACTOR BUILDING PRESSURE > 30 PSIG 2 - REACTOR BUILDING HYDROGEN CONCENTRATION > 36 BY VOLUME 3. SIGNIFICANT OTSG LEAKAGE INVCLVINC MULTIPLE TURE FAILURES.

i 4. A DIRECT REACTOR BUILDING TO ATHOSPHERE RELEASE PATH.

l WAY SUCH AS RB PURGE VALVES FAILURE TO CLOSE.

86.0 2042C l

1 ENuclear Tm Radiological controls Department 6610.PLN.4200.02_

Title Revision No.

TM! Emergency Dose Calculatica Manual (EDCM) 1 FIGURE 5 16 1 (Cont'd)

(EEAMPLE)

PROTECTIVE ACTION RECOMMENDATIONS (PAR) LOGIC DIACRAM NOTE 4:

THI EVACUATION TIME ESTIMATES

(

LOWER fHOURS) UPPER (HOURS)

BEST ESTIMATE (NIGHT) 6.75 9.50 TYPICAL WEEKDAY (NORMAL) 6.26 10.25 i

ADVERSE WEATHER 10.00 12.25 LOWER - GOOD STATE OF EMERGENCY READINESS (SLOW SCENARIO)

UPPER LACE OF ADEQUATE PREPARATION TIME (FAST SCENARIO) 1 NOTE $1 CONSIDERATION SHOULD BE GIVEN TO ThE PROJECTED EEPOSURE TO BE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATES, IN SO DOING, YOU HUST FACTOR RELEASE DURATION, RELEASE MAGNITUDE AND ASSUME A PROTECTION FACTOR OF 2 FOR UP TO THE FIRST 2 HOURS OF RELEASE DURATION AND A PF OF 1 FOR >

2 HOURS RELEASE DURATION. THE PATHWAY OF LEAST EEPOSURE

)

SHOULD SE CHOSEN. 3F THE DOGE RATE IS 400 MREM /HR; SHELTERING FOR 3 HOURS WOULD RESULT IN AN EEPOSURE OF 800 MREM.

NOTE 6:

PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICA-TION OF THE KEYHOLE CONCEPT.

(CONSIDER USING 2 MILE RADIUS AND 10 MILE DOWNWIND).

i I

\\

87.0 2042C D

MUCl88r r,a Radiological Controls Department 6610-PLN-4200.02 4

Title Revision No.

?

• %I Emergency Dose calculation Manual (EDCM) 1 FIGURE 5.16-2 (EIAMPLE) went m a asuge e rf a 3

MM MVELpIENT F PWTECTIVE 6.M MIMSEpenTWe M

.j 1 ent assa ' Bouse, a suum I (men. ossoc? e unuss I s

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88.0 2042C s

MUCIScr

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[

RadioloSical Controls Department 6610-PLN-4200.02 5

Title Revision No.

THI Energency Dose calculation Manual (EDCM) 1 FIGURE 5.16-2 (Cont'd)

(EEAMPLE)

THREE MILE ISLAND UNIT 2 PAR LOGIC DI AGPJl' NOTES NOTE 18 AS INDICATED BY ONE OF THE FOLLOWING 1 - REACTOR BUILDING PRESSURE > 4 PSIG 2 - REACTOR BUILDING HYDROGEN CONCENTRATION > 3% BY VOLUME 3 - A DIRECT REACTOR BUILDING TC ATHOSPHERE RELEASE PATH

• WAY SUCH AS RB PURGE VALVES FAILURE TO CLOSE.

4 - A DIRECT FUEL HANDLING BUILDING TO ATHOSP?iER". RELEASE PATHWAY SUCH AS A FILTER TRAIN FAILURE.

NOTE 2:

TMI EVACUATION TIME ESTIMATES LOWER (UOURS) UPPER (HOURS)

BEST KSTIMATE (NIGHT) 5.75 9.50 TYPICAL WEEKDAY (NORMAL) 6.25 10.25 ADVERSE WEATHER 10.00 12.25

)

V LOWER - GOOD STATE OF EMERGENCY READINESS (SLOW SCENARIO)

UPPER - LACK OF ADEQUAIE PREPARATION TIME (FAST SCENARIO)

NOTE 3:

CONSIDERATION SHOULD BE GIVEN TO THE PROJECTED EXPOSURE TO BE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATES, IN SO DOING, YOU MUST FACTOR RELEASE DURATION, RELEASE MAGNITUDE AND ASSUME A PROTECTION FACTOR OF 2 FOR UP TO THE FIRST 2 HOURS OF RELEASE DURATION AND A PF OF 1 FOR > 2 HOURS RELEASE DURATION. THE PATHWAY OF LEAST EXPOSURE SHOULD BE CHOSEN. IF THE DOSE RATE IS 400 MREM /HR; SHELTERING FOR 3 HOURS WOULD RESULT IN AN EXPOSURE OF 800 MREM.

NOTE 4:

PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICATION OF THE KEYHOLE CONCEPT (CONSIDER USING 2 MILE RADIUS AND 10 MILES DOWNWIND)

/

89.0 2042C

MUCl88r yx, Radiological Controls Department 6610-PLN-42OO.02 Title Revision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 i

5.17 Dose Projection Model Overview TMI-1 5.17.1 The dose projection model may be regarded as three distinct sections. Certain variables are passed between those sections n.

source term generation b.

met data input c.

dose calculation model Indeed, the operator may separately update any one of these sections using specially defined keys. To save time tho l'

operator, upon learning that the met conditions have changed and the source term has not, may update the met conditions and then update the dose calculation without having to re-enter the source term parameters. The program retains the most recent set of source term and met parameters.

5.17.2 Source term parameters that are used in the dose projection portion of the model aren 1.

An array of the 15 isotopes of interest in pCi/sec.

}

2.

The release flow rate in efm.

s_s l

3.

The release point etack height of 48.6 meters.

7 4.

The relsase point stack diameter in meters - 1.7 meters for l

the Station Vent and 1.1 meters for the Reactor Building.

5.

A single letter code for the type of release.

a.

G - Ground level b.

S - Split wake c.

E - Elevated 6.

A character string describing the source, i.e., RMA-6 300 cpm.

7.

Tne date and time of the source term parameter entry.

The met conditions section requires as input the type of

release, i.e.,

G, S, or E.

If this variable is not present the operator is required to specify it before the program continues.

i I

~.

90.0 2042C I

1

MUClear

""*b" yx3

[

Radiological controls Department 6610-PLN-4200.02

\\s,-

Tit le Revision No.

J 1

TMI Emergency Dose Calculation Manual (EDCM) 5.18 TMI-2 Ecurce Term Calculation

-.~.-----~~.-------.---------------~~~---------------------------------------

NOTE:

Follow Figure 5.18-1 THI-2 RAC Program Flowchart following this section.

~~~---------.---.--~~-------------------------------------------~~~~~--------

5.18.1 Source Term Calculations - The source term portion of the THI-2 dose assessment program is used to generate the quantity and radionuclide make up of the radioketive material released (or available for release) to the environment. Once the source term is nessured or estimated, meteorological and dosimetry models are applied to the assesstent. Some specific accident scenarios are used to calculate radionuclice release factors and assess the accident consequences. These assessments are documented in Technical Evaluation Reports (TER's) or Safety Evaluation Reports (SER's). Source Term Calculations are performed by three methods, once the release pathway is chosen. These methods are 1) using Radiation Honitoring System readings, 2) Actual sample results, or 3) Contingency calculations.

5.18.2 The following facilities are considered as being radioactive

s material release pathways for the TMI-2 RAC program, i

^~s DEFAULT VENTILATION CURIES FACILITY RMS FLOW RATE (CFM) AVAILABLE 1 - WHPF EBERLINE PING 7100 100 2 - RLM EBERLINE PING 900 100 l

3 - CACE AMS-3 (2) 2000 100 4 - EPICOR II EBERLINE PING 9000 100 5-6 - STATION VENT HPR219 VICTOREEN PING 120K - 130K 100 1

7 - STATION VENT HPR219A EBERLINE PING 120K - 130K 100 8 - ISWSF/ PAINT SEED NONE NONE 100 9 - RAD. INST. SHOP NONE 4000 100 The above list shows the associated. default data concerning each f-s release point in the THI-2 RAC program. The station vent release e

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exz Radiological Controls Department 6610-PLN-4200.02

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Revision No.

s Title THI Emergency Dose Calculation Manual (EDCM)

I pathway also includes an option for a dropped fuel canister accident. Figure 5.18-2 shows the main THI-2 ventilation.

5.18.3 Radiation Monitoring System (RMS) Source Term Calculation only the RMS channel available for a selected release pathway is offered to the user. The following parameters are used to calculate a'THI-2 source terms 1.

RMS reading (CPM, pC1) 2.

RMS Channel Efficiency (CPM /pci/cc, CPM / min /pCL/cc) 3.

The release flow rate (CTM) 4.

Cs-137/Sr-90 Rati 5.10.4 Post Accident Sample Sou.

erm Calculation This option for a particular release pethway is to use actual sample results to develope a source term. This option would be the most preferable method for calculating a THI-2 source term

/'~'N since this method eliminates built in conservatisms from RMS or

(

)

contingency calculations.

In this option the sample results or defaults, if required, will be used in conjunction with the release flow rate (CrM) to calculato a source term.

5.18.5 Contingency calculation Source Term Calculation This option utilises referenced technical documents, such as Safety Evaluation Reports (SER's) and Technical Evaluation Reports (TER's), to define a maximum source term for each facility. The quantity of radioactive material for.a given facility along with the associated release flow rate are used to calculate a conservative source term.

5.18.5.1 Source Term Filtration (Contingency Calc only)

During the final calculation of a source term a filtration efficioney will be prompted to determine the fraction of the source term that will not reach the environment due to filtration.

5.18.6 Dose Calculation once the source term is established for a release from a particular pathway. This section of the.RAC program will proceed very similar to the TMI-l dose projection process. A meteorology option will gather meteorology data and combine it with the source term information to complete a dose projection, b

sa, / -

92.0 2042C

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Radiological Controle Department 6610-PLN-42OO.02

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Title Revision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 5.18.7 Dose Pro $ection Model Overview THI-2 5.18.7.1 The dose projection model may be regarded as three distinct sections. Certain variables are passed between those sections:

I a.

source term generation b.

met data input c.

dose calculation model i

Indeed, the operator may separately update any one of these sections using specially defined keys. To save time the operator, upon learning that the met cor.4. cions have changed and the source term has not, may update the met conditions and then update the dose calculation without having to re-enter the source term parameters. The program retains the most recent set of source term and met parameters.

]

5.18.7.2 Source term parameters that are used in the dose projection portion of the model ares

['~'}

1.

The radionuclidsa Sr-90 and Cs-137 in their chosen ratio.

'V 2.

The release flow rate in efm.

i 3.

The choice of fuel related source term or not.

4.

A singlo lutter code for the type of release.

a.

G - Ground level t

b.

s - split wake c.

E - Elevated S.

The date and time of the source term parameter entry.

l The met conditions section requires as input the type of release.

4 1.e.,

G, S, or E.

If this variable is not present the operator is required to specify it before the program continues.

e V

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AUCIO8r Radiological controls Department 6610-PLN-4200.02 (s_,)

Title Revision No.

THI Emergency Dese calculation Manual (EDcM) 1 5.18.7.3 Met parin.sters that are passed to the dose projection portion of the model aret 1.

Delta T in oF per 117 feet.

2.

Wind speed in mph.

l 3.

Wind direction in degrees, from.

4.

The date and time that met conditions were entered.

Upon being invoked, the dose calculation oogment checks to set if the 15 isotopic values are defined. If they are not, the source term segment is invoked. After this the dose calculation segment checks to see if the wind speed value is defined. If it is not, the met condition segment is called. Only after these two conditions are satisfied is the dose model started.

b.18.'.4 To perform an entire dose calculation with new source term and met, with user key-1, the source term and met variables are initialized to missing and the dose projection segment is called.

The dose projection segment does not retain any variables for the gj program's use elsewhere in the program but sends all its output

• d to the screen.

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94.0 2042C

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f-Radiological Controls Department _

6610-PLN-4200.02 Title Revision No.

TMI Emergency Dose calculation Manual (EDCM) 1 FIGURE 5.10-1 THREE WILE ISLAND UNIT 2 RAC PROGRAM FLOWCHART s am 3 CACE START 4. artaa 11 5

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TMI Emeroency Dose calculation Manual (EDCM) 1 FIGURE 5.18~1 (Cont'd)

THREE MILE ISLAND UNIT 2 RAC PROGRAM FLOWCHART A /

O CALCULATIONS I

PERFORMED Y

FIN AL PRINTOUT:

(1) SOURCE TERM (2) METEOROLOGY (3) a CROUND RELE ASE (4) EMERGENCY ACTION LEVEL (5) DOSE PROJECTION (mrem /hr)

(6) INTEGR ATED DOSE (mrem)

(7) RELE ASE PolNT (8) RMS/S AMPLE RESULT D AT A (9) CS/SR R ATIO tion NOTE: ALPHA EFFECT ON SR SOURCE TERM.

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.m Radiological controls Department 6610-PLN-42OO.02 Title Revision No.

g TMI Emergency Dose calculation Manual (EDCM) 1 FIGURE 5.18-1 (Cont'd)

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""*b" rx, Radiological Controls Department 6610-PLN-4200.02 Title Revision No.

TMI Emeroency Dose Calculation Manual (EDCM) 1 6.0 RESPpNSIBILITIES 6.1 The RAC is responsible to ensure that done $ssessments using the methodology in the EDCM are performed upon it.plementation of the Emergency Plan.

6.2 The RASE has the responsibility to support the RAC in performance of radiological controls and dose assess:ent using the methodology in the EDQ4.

6.3 The Chemistry Cocedinator has the responsibility to support the RAC in the procurement and analysis of in-plant samples required to quantify the accident.

6.4 Radiological and Envircnmental Controle has the responsibility of proper review, and evaluation of the EDCM and the RAC program software.

Radiological Controls is responsible for ensuring that the EDCH and the RAC.

program software are current and compatible.

6.5 The Technical Support Center (TSC) has the responsibility to provide the RAC with appropriate fuel damage data and other information pertinent to performing dose calculations.

O o

V 101.0 2042C

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Radiological Controls Department 6610-PLN-4200.02

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Title Revision No.

l TMI Emergency Dose Calculation Manual (EDCM) 1 7.0 RETERENCES 7.1 American National Standard (ANS), ANSI /ANS-18.1 - 1984, Radioactive Source Term for Normal Operations of Light Water Reactott 7.2 APS Source Term Report - Report to the American Pnycical Society of the f

Study Group on Radionuclide Releases From Severe Acc. dents at Nuclear Power Plants, February 1985 7.3 Dose Assessment Manual for Emergency Preparednece Coordinators, February 1986, INPO 86-008 7.4 EDCM Flowchart Block Diagram

+

7.5 Efficiency Check using an Air 2-131 Source Cartridge and a Ba-133 Source j

Cartridge, Memorandum 9502-88-0139, September 28, 1988 7.6 Emergency Dose Calculation Manual (EDCM) Source Code Listing

+

7.7 EPA 520/1-75-001 - Manual of Protective Action Guides and Protective Actions for Nuclear Incidento

{

j'~'N 7.8 EPIP 9471-IMP-1300.07 - off-site /On-site Dose Projections I

t\\' /

7.9 Evaluation of a F;ont Loaded Iodine Cartridge using Various Survey Equipment, Memorandum 9100-88-0194, May 12, 1988 7.10 Field Measurements of Airborne Releases of Radioactive Material, Memorandum I

9502-88-0098, May 25, 1988 7.11 FSAR, TMI-l Chapter 11, Radioactive Waste and Radiation Protection 7.12 FSAR, TMI-1 Chapter 14 - lafety Analysis 7.13 FSAR, THI-2, Volume 10, Section 15, Accident Analysis 7.14 GPUNC Emergency Plan, 1000-PLN-1300.01 7.15 ICRP Report of the Task Group on Reference Man, 1981 7.16 INPO SOER 83-2

" Steam Generator Tube Ruptures" 7.17 Introduction to Health Physics, Herman Comber, 2nd Edition,1985

?

7.18 NRC-BNL Source Term Report 7.19 KURZG-0017 Rev. 1 - PWR - GALE Code; Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWR, April 1976 7.20 NUREG-0133 - Prnparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, October 1978.

/A)

U 102.0 2042C

QUCl68r yx2 O

Title Radiological Controle Department 6610-PLN-4200.02 Revision No.

?

TMI Err.orgency Dose Calculation Manusi (EDCM) 1 7.21 NUREG-0591 - Environmental Assessment for use of EPICOR II at Three Mile Island Unit 2, October 3, 1979 7.22 NUREG-0654 - Revision I - Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plante 7.23 NUREG-0737 - Clarification of THI Action Plant Requirements, U.S. Nuclear Regulatory Consnission, November 1980, O'eneric Letter 82-33, Supplement 1 to NUREG-0737 - Requirements for En.orgency Reeponse Capability, U.S. Nuclear Regulatory Consiscion, Washington, D.C., December 1982 7.24 NUREG-1228 - Source Term Estimation during Incident Response to Severe i

Nuclear Power Plant Accidents, October 1988 7.25 NUREG/CR-3011 - Dose Prtjaction Considerations for Emergency Conditions at Nuclear Power Plantu 7.26 N1830 - Post Accident Reactor Coolant Sampling 7.27 N1831 - Post Accident Atmospheric Sampling I

p 7.28 N1832 - Pos: Accident Sample Analysis 7.29 N1P33 - Post Accident Core Damage Calculations 7.30 OP1202 Abnormal Transients Rules, Guides and Graphs 7.31 OP1202 RCS Super Heated 7.32 OP1210 Excessive Radiation Levels

?

7.33 operational Quality Assurance Plan, 1000-PLN-7200.01 7.34 Proprietary Midas User Documentation, Pickard, Lowe, and O u rick l

7.35 Radioactive Decay Data Tables, David C. Kocher, ORNL, DOE / TIC-11026,1981.

7.36 Radio.ogical Health Handbook, Revised Edition Jan. 1970, US Dept. IEW.

7.37 Rao. Guide 1.21 Measuring, Evaluating, And Reporting Radioactivity in Solid dantes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants, Rev. 1, June 1974.

7.38 Reg. Guide 1.109 - Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix 1, October 1977, Rev. 1 7.39 SER-15737-2-GO7-108, Rev. 4, March 5, 1985. Safety Evaluation Report for a l

THI-2 Fuel Canister Accident l

[

7.40 SER-419628-003, Rev. 7, Instrument Calibration Facility, Feb. 12, 1988 NJ 103.0 2042C i

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MUCl68r

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<) Title Radiological controls Department 6610-PLN-4200.02 Revision No.

TMI Emergency Dose Calculation Manual (EDCM) 1 7.41 TDR-390 - THI-1; Primary-to-Secondary OTSO Leakage and its On-site /Off-site Radiological Impact, April 1983 7.42 TDR-405 - TMI-1; Evaluation of Plant Radiation Release and its 10CTR50, Appendix I conformance for Different opetating conditions 7.43 TDR-431 - Method for Estimating Extent of Core Damage Under severe Accident conditions 7.44 TER-13587-02-003-015, Rev. 6, January 21, 1985. Interim solid Waste staging Facility Technical Evaluation Report 7.45 TER-15737-2-003-104, Rev. 5, May 8, 1985. Technical Evaluation Report for the CACE 7.46 TER-15737-2-003-107, Rev. 6, Feb. 2, 1987. Technical Evaluation Report for

)

the WHPF.

7.47 WASH-1400 - 197's Nuclear Safety Study WASH-1400 (also known as Rasmussen Report) j 8.0 ELHIBITS 8.1 Exhibit 1, RAC Data Collection (Example)

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G 104.0 2042C

MUCIS8r "u** r G

Radioloolcal controls Department 6610-PLN-4200.02

[U I

Title Revision No.

THI Emergency Dose Calculation Manual (EDCM) 0 EXHIBIT 1 5

RAC Data Collection (Example)

Pese i RAC DATA C0tttCTION prime,v to secoadery Reisese l SYSTEM l PARAMETER l INDICATION ]

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Radiological controls Department 6610-PLN-42OO.02

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Title Revision No.

TMI Emeroency Dose Calculation Manual (EDCM)

O EXHIBIT 1 (Cont'd) osee a RAC DRfA COLLECT!DN Reactee Buildine Release lSYSTIM l PARAMtita l IN0! CATION l TIME or READ!>G l

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Revision No.

TNI Emergency Dose Calculation Manual (EDCM) 0 EXHIBIT 1 (Cont'd)

1. age 3 RAC DATA COLLECTION Auxiliary and Fuel Mandlinn Buildino Release l SYSTEM l PARAMETER l !NDICAf!DN l TIME Or atADING l

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