ML20059H903
| ML20059H903 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/07/1990 |
| From: | Danaky R, Shawn Williams GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| 6610-PLN-4200.0, NUDOCS 9009190083 | |
| Download: ML20059H903 (108) | |
Text
{{#Wiki_filter:Date 9-/O-96 TMI-1 EMERGENCY DOSE CALCULATION MANUAL (EDCM) INSTRUCTION MEMO l Office Nuc. Reactor Regulation Document Control Desk U.S. Nuclear Regul. Comm. i Washington, 0.c. ; RETURN TO: Betty Nash 20s55 j Procedure Distributton Control INFO (2 copies) ! Unit 2 Admin, Bldg. '
'TMI Please update your. Copy of theiTHI-1. Emergency Dose C41ctJation Manual (EDCM) as '
instructed below. .Also, please sign'the acknowledgement at the bottom of this memo and return to Betty Nash as shown above. REMOVE INSERT i 9/00 -?LN-stado. 0.? b lelo - ? L N .4/3 0 0 03 Wars C1 _& & h-JJ T lco ~b"* b? l 10 L Wb . 7)d A ~lh" 0
/O ADDITIONAL INSTRUCTIONS / COMMENTS 7
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! hereby acknowledge recolpt of the item above and have complied with the instructions.
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%.s TMI Emergency Dose Calculation Manual (EDCM) 0 Applicability / Scope- The EDCM is applicable to all qualified Responsible Office Radiological Assessment Coordinators. This manual provides the
~ methods uced to perform dose projections during emergencies 6610 This document is within QA plan scope K Yes .No Effective Dats 1 Safety Reviews Required K Yes No I List-of Effective Pages J Page Revision g Revision P_aage Revision P_a age Revision 1.0 0 29.0 0 57.0 0 85.0 0' '
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Radiological' controls Department Revision No. . 1
'TMI Emergency Dose Calculation Manual (EDCM)' 0-
- k Table of Contents Section _Page 1.0 PURPOSE .4.0 2.0 APPLICABILITY / SCOPE 5.0 l.
3.0 -DEFINITIONS 6.0-4.0 PP.EREQUISITES 15.0 - 5.0 PROCEDURE 16~.0 f 5.1 Source Term Calculations .18.0 5.2 Selection of Release Pathways and Characteristics 19.0-5.3 Calculation of NRC Damage Class and Isotopic Percentages 23.0 5.4 Radiation Monitoring Siystem (RMS) Source Term Calculation 35.0
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j )) 5.5 Post Accident Samplet dource Term Calculation 39.0-p; 5.6 Contingency Calculations Source Term Generation 41.0 5.7 Decay Scheme calculation 46.0 5.8 Noble Gas to Iodine Ratio calculations' 48.0 5.9 Effluent Release Flow Rates 51.0 5.10 Two-Phase Stoam Flow Determination 64.0 5.11 Source Term Filtration 65.0 5.12 Meteorology Inquiry 67.0 5.13 Dispersion Model 68.0 5.14 offsite Air Sample Analysis 74.0 q 5.15 Liquid Release calculation 77.0 , l 5.16 Protective Action Recommendation Logic 83.0
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5.17 Dose Projection Model Overview THI-1 89.0 s 5.18 THI-2 Source-Term Calculation 90.0
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k #~ Ind g gf Number Radiological controls Department 6610-PLN-4200.02' Title Revision No. TNT Emergency Dose' Calculation Manual (EDCM) 0 ,, Table of contents (Cont'd). i Section Page. I i 6.0 RESPONSIBILITIES 100.0 ! l
7.0 REFERENCES
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TMI Emergency Dose Calculation Manual (EDCM) 0 -i 1.0 PURPOSE- .
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The' purpose of this manual is to provide a document that describes the. assumptions and methodology used11n the current TMI-l and THI-2 Radiological , Assessment Coordinator,(RAC) programs. This includes calculating projected. on-site and off-site doses from releases of radioactive' material to the envircnment in accident conditions upon implementation of the Emergency Plan. As'
.such, this document describes methods 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 results, or contingency calculations, if RMS and. sample'results are not available.- Theses dose projections are performed by computer using the current version of the THI-l' or TMI-2 RAC programs.' The Radiological Assessment Coordinator is responsible for implementing the dose projection process for THI-l and TMI-2.
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Number i p-- g g , TMI. Radiological Controls Department 6630-PLN-4200.02-l{. Tit le -- Revision No. TMI Emergency Dose Calculation Mancal'(EDCM)-- 0 ll 1 I i 2.0 APPLICABILITY / SCOPE The EDCM is applicable to all quell!ied TMI, Radiological. Assessment Coordinator
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i- (RAC) personnel involved in the projection of on-site and off-site doses during-an emergency.- This manual provides the methods used in performance of doce ~ I projections during emergency situations where radioactive material has been or is predicted to be released to the environment. 4 i
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TMI Emergency Dose Calculation Manual (EDCM) 0 3.0 DEFINITIONS 3.1 BARCOCK AND WILCOX TRAILER (B&W TRAILER).- that was used for THI-2 sample analyses after the TMI-2 accident.- 'This facility has its own ventilation system including installed HEPA filters. This facility has an AMS-3> instelled as a-ventilation radiation monitor.J The ventilation system is run at 1300 CFM. 3.2 BUILDING WAKE EFFECTS - When an atmospheric release occurs et, near, or below the top of a building (or any structure) the dispersion. of the=
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-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 down to the ground on the downwind side of the structure. This has two effects: -it increases on-site concentrations dramatically, while slightly reducing concentrations downwind for a short distance. Far. downwind j concentrations are affected very little by building wake. Building wake ;
effects are most noticeable for ground level or low flow stack releases. t such as the condenser off-gas exhaust. Normal plant ventilation usually has a high enough flow that building wake does not affect the plume-significantly. Building wake is accounted for as part of the split wake release modeling. , [ - . 3.3 " CHI over Q" (X/Q) is the dispersion of a gaseous release in the-t environment calculated by-the split wake dispersion model. Parmal units of h X/Q are sec/ cubic meter. X/Q is used to determine environmental atmospheric concentrations by multiplying the source term represented.by Q. I Thus dispersion, X/Q (sec/ cubic meter) times source term, Q -(pci/sec)- "
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yields environmental concentration X=(pci/ cubic meter). X/Q is a function of many parameters including wind speed, delta T (change in temperature with height), release point height, buildir.f size, and-release velocity, among others. The release model takes all these into account when. calculating atmospheric dispersion. . 1 3.4 CONTAINMENT AIR CONTROL ENVELOPE (CACE) - This facility provides'a containment outside of the THI-2 equipment hatch. . This containment uses two AMS-3 radiation monitors to monitor releases, at 2000 CFM, from this facility when the facility is-in use. 3.5 CONTAINMENT ATHOSPHERIC POST-ACCIDENT SAMPLING SYSTEM (CATPASS) - Post accident sampling system capable of providing. sample (s) following an accident condition, coincident with a blackout, with ilmited personnel j exposure. The sampling system, located in a post-accident accessible area, provides the capability for obtaining samples of the Reactor Building ' atmosphere, within one hour after the decision has been made to acquire the sample (s). The samples (s) are then used- for radiological and -hydrogen analysis. These results will provide an indication of the extent of core
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damage and provide good data for the Reactor Building source term if a Reactor Building release is possible. 1 1 6.0 2042c
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3.6 CONTINGENCY CALCULATION - A source term calculation performed in the absr ace of suf ficierc. ef fluent radiation monitoring system readings or post : ; accident _ sample data. It'is a mathematical calculation based upon.the most representative physical model of. actual accident plant conditions.- j i 3.7 -CORE DAMAGE - (Notes This definition to be'used in lieu of'. defective / failed fuel.)- A set of core classifications used to address the i requirements of the NRC NUREG 0737. Criterion 2(a) upon implementation of 1 the Emergency Plan. ~ Based upon RCS pressure and incore thermocouple readings, an assessment is made of the degree of cladding failure, fuel' .j overheat, and fuel molt. 3.8 DEFECTIVE FUEL / FAILED FUEL - See definition of core' damage. i 3.9 DOSE RATE CONVERSION FACTOR (DRCF) - A parameter calculated by the methods- ~; and models of internal dosimetry, which indicates the committed dose 'l equivalent (to the whole body or sn. organ)-per unit activity inhaled or; ; ingested. This parameter is specific to the radionuclide and the dose l pathway. Dose conversion factors are ccamonly' tabulated in units of ! 3 mrom/hr per curie /m inhaled or ingested. I i 3.10 ELEVATED RELEASE - An airborne effluent plume which is well above any :. building wake effects so asito be essentially unentrained is termed an ! elevated release. The source of the~ plume.may be elevated either.by virtue of the physical height of the source above the ground elevation and Lulldings.or by a combination of the physical' height and the jet plume 4 rise. Semi infinite modeling of elevated releases generally will not produce any significant ground level concentrations uithinithe_first-few-hundred yards of.the source. . Semi intinite modeling of' elevated = releases generally have less dose consequence,to the public due:to the greater. y ' downwind distance to the ground concentration maximum compared to ground releases. Elevated releases as used in the EDCM actually means "not at j ground" in the split wake plume model. LOther definitions of " elevated" - l with respect to plumes abound in literature. 3.11 EMERGENCY ACTION LEVEL (EAL) - Predetermined conditions or values, including radiation dose rates; specific levels of airborne; waterborne; or surface-deposited contamination; events such as natural disasters or fires; or specific' instrument indicators which, when reached or exceeded, require implementation of the Emergency Plan. 1 3.12 EMERGENCY DIRECTOR (ED).- Designated on-site individual having the responsibility and authority to implement the Emergency Plan, and who will coordinate efforts to limit consequences of, and bring under control,' the emergency. I 3.13 EMERGENCY DOSE CALCULATION MANUAL (EDCM) - This controlled dose calculationt l manual is the documentation describing the content and calculational methods of the Radiological Assessment Coordinator (RAC) program. r-l' 7.0 2042c
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Title Revision No. J *YMI Emergency Dose Calculation Manual (EDCM) 0- ( 3.14 EMERGENCY OPERATIONS FACILITY (EOF) - The' Emergency.Operatione Facilities- - p serve.as the primary locations for management of the Corporation's overall l emergency response.- These facilities are equipped for and staffed by the
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'-j Emergency Support Organisation to coordinate emergency response with eff-site support agencies and to assess.tho' environmental' impact of the,
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~ emergency. The.ROF participates in accident assessment and transmits L ,
. appropriate data and recommended protective actions to Federal,1 State and-Local agencies.
3.15 EMERGENCY PLANNING EONE (EPB) - There-sre two Emergency' Planning: Bones.L . , .; 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 Plume Exposure Pathway EPZ. The second is an area approximately 50 miles.in' radius around the siter for which emergency planning consideration of the ingestion exposure pathway has been given..
This is called the Ingestion Exposure Pathway EPE.. l 3.16 ENTRAINMENT - When a release in treated as a wake split release an (; entrainment factor is applied to specify how much of the release =1s to be~ 7 considered elevated and how much is-to be considered a ground release.
} Entrainment. factor is related to the building wake effect.. The entrainment-g G factor.is computed on a case by case basis and is dependent on both the . ,
stack exit velocity and the wind speed.. At low wind speeds and high exit i velocities, building effects are lowest and the entrainment-factor-selects for elevated release.- At high wind speeds and/or low exit-velocities the building effect is highest 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 form for the application of the entrainment factor (Ef) is: 1 J X/Q(splitwake)=X/Q(ground)*Ef + X/Q(elevated)*(1-Ef). to can be seen from'the formula, when the entrainment factor is one, the
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release is entirely ground and when the entrainment~factorJim. sero, the release is entirely elevated. When 0 ( Ef < 1 then the-release is split. 3.17 EPICOR II - Radioactive Liquid Waste Processing-Facility' located on the , east side of THI-2. This facility is'used'to process-TMI-2 radioactive-l waste. An Eberline PING radiation monitor is located on the ventilation-system of thin facility. The ventilation system average flow rate is 9000'CFM. 3.18 EXCLUSION AREA (EA) - As defined'in 10CFR100.3; "that area surrounding the reactor, in which the reactor licensee has the1 authority to. determine all' U activities including exclusion or removal of. personnel and property from the area". At TMI this is an area'with a 2000 ft. radius from the-point . equidistant between the centers of the TMI-l and THI-2 reactor buildings. O 8.0 2042c-
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t Radiological Controls Department 6610-PLN-4200.02 J Title Revision No. l l TMI Emergency Dose Calculation Manual (EDCM) 0 -] 4 3.19 EXIT VELOCITY AND PLUME RISE - Atmospheric dispersion and' ground = concentrations are in part dependent on release height. . Higher release helv5te will cause: lower maximum _ concentrations at ground 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 j 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 effective stack height.
-3.20 FINITE PLUME HODEL - Atmo'pheric s dispersion and dose assessment'model.which l is based on the assumption that the horisontal_ and vertical dimensions of - ,
an effluent' 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 factor, t.sd the air. attenuation of the gamma rays coming from the cloud. -This model can estimate the dose to a1 receptor who is not submerged in the radioactive cloud. It is particularly useful in evaluating. doses from an elevated plume.or when the receptor is'near tho' effluent source. .1 p 3.21. FUEL HANDLING BUILDING ENGINEERED SAFETY FEATURE VENTILATION SYSTEM - The i Fuel Handling Building ESF Ventilation System, is being added to TMI-1 :in
\g, accordance with a commitment to the NRC.- This commitment has been included i in the NRC TMI-1. restart report. The Fuel Handling Building ESF Ventilation System is, installed to containi, confine,. control, mitigate,.
monitor and recor! 1diation release resulting from a THI-1 postulated spent fuel accident in the Fuel Handling Building as described in-FSAR, Section 14.2.2.1, Update 1, 7/82'.1 Normal' operation of'the Fuel Handling. Building ESF Ventilation System will be during- THI spent fuel ~ movements . in the Fuel. Handling Building. The system design shall include adequate air filtration and exhaust capacity to ensure that no uncontrolled radioactive release to atmosphere occurs. 'The System.shall include' effluent radiation monitoring capability. i 3.22 GAUSSIAN PLUME EQUATION - An equation which takes input parameters of plume ' E height, sigma-Y, sigma-2, and wind speed, which explicitly. calculates the straight line Gaussian Plume Dispersion. The Gaussian Pluae equation , actually averages short term; variations to produce a mean effective plume, j so short term measurements of the plume may not be duplicated by the-Gaussian Plume Model. 3.23 GROUND RELEASE - An airborne. effluent plume which contacts the ground . , essentially at the point of release either from a source actually located-L at the ground elevation or from a source well above the ground elevation- l 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' l 1evel 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 i g releases also have generally lower X/Qs all the way downwind.
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. i 3.24l HYDROGEN PURGE SYSTEM - Post-accident containment purge system is designed i to maintain the hydrogen concentration of the post-accident containment 0 atmosphere below the. lower flammability. limit. :The system does this by introducing'outside air into the Reactor Building, which allows the !
displaced containment atmosphere to be discharged in a controlled manner into the normal. Reactor Building exhaust duct.- In the flow path three- , release rates exist which can be additive to give flow from 1.to 1250 CFM. 3.25 INTERIM SOLID WASTE STAGING FACILITY.(ISWSF/ PAINT SHED) and~the VMI-2 PAINT: ' SHED OR RADWASTE NATERIAL' STORAGE 7ACILITY (RMSF) - These faci?.ities have ~ no ventilation system or radiation monitor, but hayw the potential to- - release radioactive material to the environment.. L 3.26 LOW POPULATION EONE (LPZ)=- As defined.in-10CFR100.3 "the area immediately 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 measures'could be taken.in their behalf in the eventi of a serious accident. , t 3.27 HETEOROLOGICAL'INFORMATION AND DOSE ASSESSMENT SYSTEM (MIDAS)' ,This'is the acronym for the computer program that can be used by the. Environmental. f'~'g Assessment Command Center (EACC) to. project'off-site doses for routinec 7
-effluents and releases during emergencies.;J The MIDAS' program runs on a -
main frame computer. Some features.of MIDAS that are'not in the RAC < program are ingestion pathway doses, liquid and gas population doses,' dose ~ ' projections at any desired point of interest, and-sector dose-integration. 3.28 NRC DAMAGE CLASS - A method of estimating the extent: of core damage per j NUREG-0737 Criterion 2 (a) under accident conditions requiring. implementation of the Emergency Plan.. The' initial estimate of the. degree. ' of reactor core damage is cerived from the calcule.ted radionuclide , concentratione'that are measured 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". , 3.29- OFF-CENTERLINE DOSE CALCULATIONS - Dose calculations that are calculated at i various distances from the calculated' plume centerline (0, 50,'100, 150,
'250, and 500 meters). These calculations are performed at.2-8 distances frem the plant.
3.30 OFFSITE AIR SAMPLE ANALYSIS SYSTEM - An air; sampling and analysis system i specifically designed for iodine air sam'pling and1 thyroid dose assessment.'. The system cm.aists of an air pump ' unit which draws air'through a canister containing a highly efficient iodine adsorbing material and a-Geiger l Mueller detector for canister evaluation. 3.31 PARTITION FACTOR - (Condenser), see NUREG-0017
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3.32 POST-ACCIDENT SAMPLING SYSTEM (PASS) - System used for acquiring a pressurised liquid sample of the RCf. during omrergency conditions. The i post-accident reactor coolant sampling system provides a-means of obtaining ) a representative. sample of reactor coolant, including dissolved gaaet,. .i reactor coolant bleed tank contento and reactor containment sump centents,. within one hour after the decisior.to acquire the sample, without excessive
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operator exposure or compromise of interfacing safety-related systems. 3.33 5ROTECTIVE ACTION OUIDE (PAG)l- ?rojected radiological dose or dose
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commitment values to individualte of the general' population end to emergency _ wtrkers that warrant protective' action before or after a release of ] radioactive material. Protective actions would be warranted provided the-reduction in individual dose taxpected to be achieved by carrying out the i prot ective action -is not of f set by excessive risks to individual- safety in I l takitg the protective action. _ The protective action guide does not include -
' the c.ose that has unavoidably occurred prior to the assessment.
3.34- PROTECTIVE ACTION RECOMMENDATION (PAR)'- Those actions taken during or after en emergency situation that are intended to minimize or eliminate the hazard to the health and safety of the general;public and/or'on-site personnel. < l' 3.35 RADIATIOli INSTRUMENT SHOP - Facility used to repair / calibrate / maintain ( instrumentation used by' Radiological and Environmental' Controls Department s. This facility includes calibration sources that'may possibly i i be released in a worst case accident-(e.g., fire). This facility has a-ventilation system, (rated at 4000 CFM), but no installed radiation . y monitor, or filtration.
- f. 3.36 RADIATION MCNITORING SYSTEM-(RMS) - The.RMS detects,; indicates,'_
annunciates, and records the radiation level at selected' locations inside and outside the plant to verify compliance with applicable Code of Federal Regulations (OFR) limits. The'RMS consists-of tte followingLsubsystems: area monitoriag, atmospheric monitoring, and liquid monitoring. 3.37 RADIOIODINE PIJ TEOUT - Iodines are very chemical .y < reactive, being members J of the halogen family. As such, iodines have a eery high probability of , reacting with a.most any other material they coem in contact with. / Radiciodine platsout-is a generic term for~the nochanisms by which radioactive iodines are removed from a waste stream by. contact with materials not specifically designed or engineered for radioiodine removal. 4 Examples of potential radioiodine' plateout reaf:tions are the removal of iodine from gaseots wastes by adsorption onto interior surfaces of-ductwork- i and piping and on any exposed surfaces of the room or building originating l l' the release. i
.l l 3.38 RADIOIODINE PROCESSOR STATIONS (HAP-S) - System used'for acquiring '
particulate and iodine camples from the Reactor Building Exhaust, Auxiliary and Fuel Handling Building Exhaust or the Cor. denser Off-gas Exhaust during-emergency conditions. The stations are controlled by solenoid valves which x
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Title Revision No. 4 TMI Emergency Dose Calculation Manual (EDCM) 0 , stream. Flow'is actuated through<(3) parallel filter cartridges pe.- station. The' sampling times are adjustable on each Ic. cal control panel. The filter cartridges must be removed manually for a.mlysis. 3.39 RADIOLOGICAL AssEsshENT COORDINATOR (RAC) - The RAC la responsible for all f
-on-site radiological assessment activities.- Initially, the PAC is.
. responsible for directing the on site and off-site sarvey teams. The'RAC-~ ,
is relieved of off-site radiological monitoring respnsibilities by'the t Environmental Assessment Coordinat.or. The RAC performa dose projections, based upon source term estimates and provides information to the RAC. Initially the Group Radiological Control Supervisor assumes the. role of the RAC until relieved by nho 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,. j and overall' assessment and control'of radiological ~ hazards.- Normally one j RAC and one RASE are on duty at all times.. l 3.41 ~ REACTOR COOLANT SYSTEM (l'.CS) - This system contains the necessary piping .l ' and components to provide suf ficient watur flow to cool the reactor. - This' . A) system provides for the transfer of thermal energy from the reactor core tof (d the once through steam oenerators-(OTS0) to make steam, acts as a moderator-for thermal fission, and provides a boundary,to separate fission products from the atre.osphere.- 4 i l 3.42 RELEASE DURATION - Release duration refers to.the time interval during which radionuclides are roleased 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 estimates of the-plant technical staff. In the absence of other information, use two hours as the expected release d2 ration. For purposes of-determining whether to .{ take a protective action on the basis of projected dose from an' airborne-- ; plume, the projected dose should not include the dose that.has already been received prior to the tina the dose projection is done. ; 3.43 RELEASE RATE - This term refers to the rate at which radionuclides'are j ' released to=the environment. Normally, it will be expressed in curies per second (C1/sec) or micro:uries per second'(pci/sec). 3.44 RESPIRATOR AND LAUNDRY MAINTENANCE FACILITY (RLM) - This facility is used to process clean and maintain laundry and respirators for TMI-l and TMI-2.- .j This facility's 900 CrH ventilation system is monitored using a Eberline ~ j PING radiation monitor. 1 4 3.45 RMS RESPONSE FACTOR - Parameter which is used to convert RMS monitor count .i rates to total microcurie per cubic centimeter of the assumed or measured j radionuclide spectrum passing by the monitor. This is different from a ' meter calibration-factor which does the same thing for.a single calibration 4 nuclide. These factors are adjusted for changes in mixture decay. '
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1 TMI' Radiological Controls Department' 6610-PLN-4200.02
-Title Revision No.
TMI Emergency Dose calculation ~ Manual (EDCM) 0-3.46. SEMI-INFINITE PLUME HODEL - Dose assessment model-which is based on the. 1 assumption that the dimensions of an effluent plume are large compared to l 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. i just as well be infinite since radiation emitted from beyond a certain ,- 1 l distance will not reach the receptor. The ground is considered to be'an infinitely'large flat plate and the receptor is assumed to be standing at 3 the center of a hemispherical cloud of-infinite radius. The. radioactive ' cloud is limited to the space above the ground plar.o. . This-'is the origin of the name SEMI-I:! FINITE PLUME.~ The noble gas MPC's were developed on the' ,
-basis of the semi-infinite plume model. }
3.47 SIGMA-Y AND SIGMA-Z - Parameters of the Gaussian dif fusion 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; activity available for release.' The. common units for the' source term are , curies, curies /Sec, or multiples thereof:(e.g., microcuries). .The term t
" Source Term" derives from the equations involved.in doing dose. '
calculations, since the equations'contain many terms:(a term being mathematical nomenclature for a portion of an-equation), the " Source Term" t is that portion of the' equation which addresses the activity released. ; Although the term " Source Term" is used loosely to mean almost any. activity l ~ for airborne, liquids, and even dose rate calculations in plant, strictly .; j speaking " Source Term" applies only to radioactive material actually ' released. 3.49 SPLIT WAKE RELEASE - Airborne releases, for purposes of assessing off-aite i dispersion, must address the elevation of the release since wind speed
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changes with height, buildings affect dispersion for low releases and even l wind direction can be different. Mtny release points are actually at a height where, given different conditions.of' release flow rate and-l l meteorology, could either be most accurately described as, ground or
. elevated releases, or some mixture between the two.' The purpose.of l treating a release as a split wake release'is to address this problem.- i When a release' point 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 intermediate conditions, the model j 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. 3.50 STABILITY CLASS - Dispersion of an effluent plume in the atmosphere-is a function of the amount of mixing occurring between the plume and the atmosphere around the plume. The~ amount of mixing is related to what is referred to as the stability of the atmosphere. Conditions which create' good mixing are unstable and conditions which create poorer mixing are. . stable. Pasquill stability class is a breakdown of the relative atmospheric stabi'ity into seven groups, denoted as A through G, from most I 13.0 2042c
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k c ENuclear ,,, ~ Radiological' Controls Department a= ~ r 6610-PLN-4200.02 N'- -Title Revision No. .; TMI Emergency Dose Calculation Manual'(EDCM)- 'O unstable to most stable. In' the pasquill stability class system, stability . I l is related to the relative change in temperature 'with. height, d/ elta T. The
.mcre negative the change in temperature with increasing heighti the more' ,
unstable the atmosphere.~ The RAC program uses sensors on the t Noteorological tower at 33 feet and 150 feet to determine the delta T.~ r once the delta T is determined, a stability class'is selected based on the delta T and the atmospheric dispersion (X/Q).is calculated based on the. . selected' stability class. j 3.51 STATION [ VENT HPR-219 - This radiation monitor and release pathway is:the- ~ main release point for TMI-2. All ventilation from the Reactor Building, i Auxiliary Building, and Fuel Har.dling Building era routed to the station vent release point, at an average flow rate of 120,000 to 130,000 CFM.. 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-l radiation monitor.RM-A2 with a moving particulate filter. 3.52 STATION VENT HPa-219A - Eberline PING unit installed in THI-2 to monitor-
.the THI-2 Station Vent Stack. The read out unit is located in the TMI Turbine Building.'.This unit is used in conjunction with HPR-219 to monitor the main THI-2 release pathway.
3.53 -TERRAIN FACTOR'- The terrain factor is the terrain-height in meters above: 1 plant grade.- Terrain factor varies with sector and distance from the: release point. 3.54 TWO PHASE RELEASE --Liquid and steam release from the main steam safety
' relief valves. .Following discharge to the environment the steam fraction- .
is_ calculated assuming there is no change in system entropy and that the OTSG 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-952A). 3.55 WASTE HANDLING AND PACKAGING FACILITY (WHPF) This facility.is used to handle and package radioactive waste mainly from TMI-2.= This facility'n ventilation.is monitored by a PING /AMS-3 radiation monitor, and runs at 7100 CFM. 3.56 WIND SPEED ADJUSTMENTS - Since wind speed varies 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 depandent'on tho' stability class.
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f (? Ib & &( .yg Radiological Controls Department 6610-PLN-4200.02 Revision No.. Title TMI Emergency Dose Calculation'Hanual=(EDCM)' 0' 4.0 PREREQUISITES ! 4.1 The following are the prerequisites for performance of TMI projected doses l using the methods in the EDCM, and the current TMI-1 or TMI-2 RAC Program. l 4.1.1 The Emergency Plan is being implemented.
.i 4.1.2 The RAC station is manned and functional, q
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WINuclear ,,, am=~ Radiological Controls Department-- -6610-PLN-4200.02 Title Revision No. I TMI Emergency Dose Calculation Manual (EDCM) O
- l 5.0 PROCEDURE
.This section of the EDCM is' divided into the programs that are contained in the RAC computer programs for THI-l and TMI-2. Listed below is a table of contents for the procedure section of the EDC#
5.1 THI-1 Source Term Calculations 5.2 Selection of. Release Pathways and Characteristics. ,l 5.3 -Calculation of NRC Damage Class and Isotopic Percentages ; 1 5.4 Radiation Monitoring System (RMS) Source' Term Calculation. 5.5 Post Accident Samples Source Term Calculation .
-i 5.6 Contingency calculations Source Term Generation 5.7 Decay Scheme Calculation 5.8 Noble Gas to Iodine Ratio Calculations i
5.9 Effluent Release Flow Rates 5.10 Two-Phase Steam Flow Determination 5.11 Source Term Filtration 5.12 Meteorology Inquiry ; i 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, THI-1 5.18 TMI-2 Source Term Calculation Each part of this section explains what each program does and how it does it.. To use the TMI-1 or TMI-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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( Radioloolcal Controls Department 6610-PLN-4200.02 _\ Title Revision No. TMI Emergency Dose calculation Manual (EDCM) O
- 3. Turn on printer-(ON/OFF switch). !
! 4. Insert THI-l or THI-2 Disk into the A disk drive (3.5" disks).
- a. Turn on computer (right side ON/OFF switch)-
1. ! b. RAC program will load within 1 minute.- To reload or interrupt program with the computer ON - hiti NOTE: the Ctrl, Alt, Del keys at tho'same time, with disk in l the PC.
- 5. For computers with RAC program on a hard drives .,
q
- a. Turn on computer.
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- b. Menu will appear, choose appropriate.RAC Program, or
- c. If no menu, use directory RACl for THI-1 or RAC2 for THI-2.
O h d. Type RAC. i
\
- e. The RAC program will load within 1 minute.
- 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'l-2 minutes and will' print when completed. ;
- 9. When dones Remove all disks, turn off CRT, computer, and printer.' Store-disks appropriately. ,
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6610-PLN-4200.02 Title. Revision-No. TMI Emergency Dose Calculation' Manual (EDCM) 0 5.1. Source-Term Calculations -.The source term pordion of:the THI-1 dose.
-assessment program is used to generate the quantity and radionuclide make up of the radioactive material- released (or available for release) to the environment. -once the source term is measured.or estimated, meteorological and dosimetry modvis are applied to the assessment.. Some specific accidenti scenarios are'used to calculate-radionuclide release factors and assess the-accident consequences.
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MUClear. "" ** r - L ,, l 'F Radiological Controls Department 6610-PLN-4200.02 "A ( s Title Revision No. . TMI Emergency Dose Calculation Manual (EDCM)- 0 5.2 Selection of Release Pathways and Characteristics - The THI I computer. ! -program will prompt for the release pathway'and the release ' i l characteristics.. 5.2.1 The following are the Release Pathwayst' , t J
- 1. OTSO Tube Rap ale Release -
-- Includes:' via the condenser off-gas or'directly to atmosphere.- >
- 2. Reactor. Building' Release
- 3. Station Vent Release'
- ' Includes: ' Auxiliary Building and Fuel Handling Building.
.5.2.2 The following are the Release Characteristics {
l.. -0TSG Tube' Rupture.via condenser off-gas I
?[m 2. OTSO Tube Rupture directly to atmosphere via the Main Steam
( Reliefs or. Atmospheric Dump Valves .; t
- 3. LOCA in the Reactor Building
- 4. Fuel Handling Accident =in the Reactor Building-
- 5. Fuel Handling Accident'in'the Fuel Handling Building, including ESF Fuel Handling-Building Releases t
6.. LOCA in the Auxiliary Building
- 7. Waste Gas Release 5.2.3 The following choices are now offered for the method to be used in the source term generation:
- 1. Use RMS
- 2. Use Post Accident Sample' Result
- 3. Use contingency calculation
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f Nuclear: ,,I Radiological Controls Department "y6610-PLN-4200.02'
'i I
A Title, , Revision No.
'TMI~ Emergency Dose calculation Manual (EDCH) 0 1
4
'5.2.4 The TMI-1 RAC computer program accommodates airborne releases, from the following pathways-(See Figure.5.2-1):'
.[
~
A.- The OT80 Tube Rupture'such as
- 1. - RM-A5 condenser off-gas >
-2. RM-A5 High-condenser off-gas
- 3. RM-G25 condenser Off-gas. '-
t
- 4. RM-G26 Main Steam Reliefs and Atmospheric Dump Valves: b
- 5. RM-G27 Main Steam Reliefs-and Atmospheric. Dump Valver. ,
- 6. RM-A5 MAP-5 Samples
- 7. Main steam Release directly to the atmosphereL
- 8. Contingency Calculations without RMS or: Samples B.
ps -The Reactor' Building such as:
.5 . .
\, 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. MAP-5 Samples
- 7. Contingency calculations without RMS or Samples C. The Station Vent such as:
1.. RM-A4 Fuel Handling Building Exhaust D .,
- 2. RM-A6 Auxiliary Building Exhaust
- 3. RM-A8 Station Vent (Auxiliary and Fuel Handling Buildings)-
- 4. RM-A8 High-Auxiliary Building Exhaust ,
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' Number AN . , TMI i Radiological Controls Department ' 6610-PLN-4200.02 (
_f^*): sd Title - Revision No. TMI Emergency Dose calculation Manual '(EDCM) 0 c
- 5. RM-A14 - ESF Fuel Handling Building' Exhaust t
- 6. MAP-5 Samplee 7 -Contingency Calculations without RMS or Samples
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,,oer 6610 PLN-4200.02' Title - Revision No.
1 TMI Emeroency Dose calculation Manual (EDCM) 0 - t FIGURE 5.2-1
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.r Nuclear a= 6 < j p Radiological-Controls Department TMI 6610-PLN-4200.02- 'F i'
Title Revision-No. ; TMI Emergency Dose Calculation Man'ual'(EDCM) 0 4 5.3 Calculation of NRC Damage Class and: Isotopic Percentages This calculation . will determine the mix or percentages of.the following fifteen
- radionuclides.-
5.3.1 Ten' Noble Gases- Five Radiolodines
- 1. Kr-8om 1. 1-131
- 2. Kr-8C , 2. I-132
- 3. Kr-87 :3. -I-133
- 4. Kr-88 ~4. 1-134
- 5. Xe-131m 5. 1-135- {
- 6. .Xe-133m ;
- 7. Xe-133
-T
- 8. Xe-135m-p3
[ 9. Xe-135
'~~
- 10. Xe-138 5.3.2 The NRC Damage Class Determination- '.
The deiermination of the NRC. Damage Class is performed using . various core temperature regions from operations Procedure 1210-8, see Figure 5.3-1. The cor= comperatures used in this section of the program come from operatings computer pt C4006, which is the average of the-five highest incore_ . thermocouples.. The curves relating to saturation, and claddir.g , failures are approximated by straight line equations. NRC damage' I 5 classes 11 - 10 are based on the different pressure.and
- temperature regions of Figure 5.3-1~. .
' NOTE: A 5% allowance is made for the accuracy of the average of 5 incore temperatures from the C4006 reading from Operations. i 5.3.2.1 Core Temperature Regions - Figure 5.3-1 The region to the left'of cus. , C represents normal RCS_ activity, NRC Class-1. The region between curves C and D represents-RCS plus a percentage of gap activity, NRC Class 2 - 4. The region .
between Curve D and curve E represents RCS plus all gap' activity' ' s plus a~ percentage of noble and volatile fission ~ product _ release from fuel grain boundaries, (CS,'I, Rb), NRC Class 7;. The ,
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_ __ M M o,Qe i entrols Department _6610-PLN-4200.02 Title Revision No. TWI Emergency Dose Calculation,Manag]. (EDCH) 0 region between Ctrve E and 25500" incore temperature reptosente RC8 activity plus 100% of the gap activity and 100% of the in veessi melt release assuming NURE0-1220 release fractions, NRC Class 4 - 10. , 5.).2.1.1 The matrix below shows the theory of fuel damage based on k' TDR-431. NRC DAMAGE CLASS hEGREE OF MINOR INTERMEDIATE MAJOR DEGRADATION <104 10 - 50% >50% , No Fuel DagsOe (RCR) -No Damege : Class 1 ~ Cladding _ Failure (CAP) 2 3 4 Fuel Overheat (Fuel Matrix) 5 6 7 Fuel Melt (Fuel Matrix) 8 9 10' 5.3.3 Calculation of Radionuclide Mix Percentages Based on NRf .amage 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 Fuel Matrix Activity. The program models the various combinations of activities for each NRC Demage Class as follows hCS GAP FUEL NPC DAMAGE ACTIVITY ACTIVITY MATRIX PLASS FRACTION FRACTIE 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 1 8 1 1 1 9 1 1 1 10 1 1 1 Therefore, as an examplet NRC Damage Class-6 would consist of 100% RCs Activity, plus 100% of the GAP Activity, plus 50% of the Fuel Matrix Activity. 24.0 2042c
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N I TN! g Radiological Controle Department 6610-PLN-4200.02
\ 71tle Revision No. I TN! Emereeney Dose calculation Manual (EDCN) 0 !
6.3.3.1 TNI-1 Normal RCS Activity - TNI-1 normal RCS Activity (WRC Damage Claes-1) and the rapid power transient RCS Activity (NRC Damage Clace-1A) are listed in the next table. The normal $efault RCS Activity is based on operational RCs activity frosi cyci. a. on August 3, 1990 at 0450. This normal RC8 activity (NP2 Damage Clace-1) can be modified in the RAC program, if a current RCS . sample result is available. l i 5.3.3.1.1 Rapid Power Transient RCS Act.'vity represente the " spiking" of , the radiciodines and noble gases during power changes. Rapid Power Transiente are a power change of greater than or I equal to 10% over one minute. The Technical Support Center (TSC) ; should be requested to confirm a rapid power transient. )! The default " spiking" factore for radiolodines and noble gasee ) ' are: Radioiodines timee 50. (50X) Noble gases times 2. (2X) The Tsc should be requested to provide the actual " spiking" ,) factors for the situat! 3. The default factors are to be used if I l . information le not available from the Tsc. I l- This " spiking" of radiciodine and noble gas activities represent ] l an increase in RCS radioactivity due to a plant evolution and do not represent an indication of fuel damage. i F 4 4 1 25.0 2042c
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l ( Radiological Controle Department 6610-PLN-4200.02
\ Title Revision No. ]
TMI Emergency Dope Calculation Manual (EDCM) 0 1 5 3 3.2 Tables of NRC Damage Classes 1 and th-
- m -1 NORMAL RCS mAPID POwtR TRai,'-en' acs l 1
' DAMAGE I NRC DAMAGE CLASS 1 CLASS 1A uC1/cc** Percent curies ** DC1/cc Percent Curies **
1-131 1.298-03 0.66 2.768-01 6.465-02 1.33 1.385+0) j 1-132 2.088-02 10.52 4.46E+00 1.04E+00 21.40 2.23E+02 I-133 1.45E-02 7.34 ~3 11E+00 7.258-01 14.92 1.55E+02 ; I-134 3.228-02 16.29 6.90E+00 1.61E+00 33.14 3.45E+02 1 1-1.15 2.425-02 12.24 5.18t+00 1.21E+00 24.90 2.593+02 SUBTOTAL 9.30E-02 47.05 19.92 4.65E+00 95.69 9.96E+02 KR-85M 3.508-03 1.77 7.50E-01 7.00E-03 0.14 1.50B+00- , KR-85 0.00E+00 .0.00 0.00E+00 0.00E+00 0.00 0.00E+00 i KR-87 6.81E-03 3.45 1.46E+00 1.36E-02 0.28 2.92E+00 i KR 8.45E-03 '4.28 1.81E+00 1.69E-02 0.35' 3.62E+00 J XE-131M 0.00E+00 0.00 0.00E+00 0.00E+00 0.00 0.00B+00 -! XE-133M 0.00B+00 0.00. 0.00E+00 0.00E+00 0.00 0.00E+00 1 XE-133 3.09E-02 15.63 6.62E+00 6.18E-02 1.27' 1.32E+01 ) XE-135M 7.91E-03 4.00 1.69E+00 1.58E-02 0.33 3.39E+00 1 XE-135 2.71E-02 13.71 5.81E+10 5.42E-02 1.12 1.16t+01 ' XE-138 2.00E-02 10.12 4.28E+00 4.00E-02 0.82 8.575+00 i SUSTOTAL 1.05E-01 52.95 22.42 2.09E-01 4.31 4.48E+01 TOTAL 1.98E-01 100.00 4.23E+01 4.86E+00 100.00- 1.04E+03 NOBLE GAS TO 1.13' 1.13 1.13 0.05 0.05 0.05 ; T IODINE RATIO
* " NORMAL" RCS CONCENTRATION AND PERCENTAGES ARE FROM NORMAL CYCLE 8 RCS OPERATIONAL DATA
**THESE CURIE VALUES ARE BAWED ON NORMAL RCS VOLUME OF $6,595 GALLONS l e
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Radiological Controls Jepartment 6610-PLN-4200.02 l Title Revision No. TM! Emergency Dose Calculation Manual._(EDW) 0 5.3.3.3 3ap Activity and Fv. Matrix Activity - Gap activity and Fuel. Matrix activity uoss in the program are determined from TDR-431. l these curie activities are based on irradiation of the entire .j core at full power, 2535 MW,, for 930 days. l TMI Unit 1 l GAP Fuel Matrix Fuel ) Act. GAP Act. Matrix ; Isotope Curies % Curies % j Kr-85m 4.84E+04 0.45 2.13E+07 2.36 Kr-85 7.48E+04 0.70 8.59E+04 0.01 ; Kr-87 2.63E+04 0.25 3.90E+07 4.33 ! Kr-88 6.67E+04 0.63 5.91E+07 6.56 ] Xe-131m 7.96E+04 0.75 5.40E+05 0.06 i Xe-133m 9.30E+04 0.87 3.09E+06 0.34 l Xe-133 8.34E+06 78.31 1.28E+08 14.20 -; Xe-135m 2.72E+04 0.26 3.37E+07 3.74 i ' Xe-135 3.45E+04 0.32 1.595+07 1.76 Xe-138 0.00E+00 0.00 0.00E+00' O.00 2-131 1.29E+06 12.11 6.37E+07 7.07 3 g 1-132 1.85E+05 1.74- 9.70E+07 10.76-I-133 2.79E+05 2.62 1.43E+08 15.86 1-134 1.74E+04 0.16 1.67E+08 18.53 2-135 8.83E+04 0.83 1.30E+0B 14.42 I Sum 1.13E+07 100.0 Sum 9.02E+08 100.0 l l l J s I i i 27.0 2042c 1 i, 4 -. , , _ , . , , . . . , , . , _ , _ , _ . _
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s ENuoloar ,, Radiological Controls Department
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6610-FLN-4200.02 Y,1tle Revision No. TMI Emergency Dose Calculation Manual (EDCM) 0 5.3.3.4 The following tables represent the NRC Damage class 2 - 10 mimes, percentages, curies, and concentrations used in the RAC code. DAMAGE DAMAGE CLASS 2 CLASS 3 uct/cc** Percent curies DC1/cc** Percent Curies 2-131 6.03E+02 12.0) 1.29E+05 3.01E+03 12.11 6.45E+05 I-132 8 668+01 ~1.74 '1.86E+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-134 8.43E+00 0.17 1.80E+03 4.10E+01 0.16 8.763+03 2-J.3 5 4.15E+01 0.83 A.88E+03- 2.07E+02 0.83 4.42B+04 SU8 TOTAL 8.70E+02 17.45 1.86E+05 4.35E+03 17.46 9.30E+05 KR-85M 2.288+01 0.46 4.87E+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 1 K' *8 3.15E+01 0.63 6.73E+03 1.56E+02 0.63 3.34E+04 XE .J1M 3.72E+01 0.75 7.96E+03 1.86E+02 0.75 3.98E+04 XE-133M 4.36E+01 0.87 9.32E+03 2.17E+02 0.87 4.65E+04 XE-133 3.90E+03 78.29 8.35E+05 1.95E+04 78.30 4.172+06 fN XE-135M XE-135 1.28E+01 1.72E+01 0.26 0.34 2.74E+03 3.68E+03 6.37E+01 8.17E+01 0.26 0.33 1.36E+04 1.75E+04 i ( EE-138 1 70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 ' 1 l 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+06 ) l l NOBLE GAS 4.73 4.73 4.73 4.73 4.73 4.73 TO. IODINE RATIO
**THESE pC1/cc VALUES ARE BASED ON NORMAL RCS VOLUKE OF 56,595 GALLONS j 1
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I p & 5 yg Radlological Controls Department 6610-PLN-4200.02 i Title Revision No. l
'tWI Emergency Dose calculation Man'ual (EDCM) 0 I,
DAMAGE DAMAGE CLASS 4 CLASS 5 uci/cc** Percent Curies NCi/cc** Percent Curies j 1-131 6.03E+03 12.11 1.29E+06 3.5EE+04 7.60 7.66t+06 I-132 8.65E+02 1.74 1.85E+05 4.62E+04 9.81 9.89E+06 l I-133 1.30E+03 2.62 2.798+05 6.81E+04 14.46 1.468+07 l I-134 8.16t+01 0.16 1.75E+04 7.81E+04 16.59 1.67E+07 1-135 4.13E+02 0.83 8.83E+04 6.12E+04 12.99 1.31E+07 sUSToTAL 8.69E+03 17.46 1.868+06 2.89E+05 61.44 6.19E+07 I KR-85M 2.26E+02 0.45 4.84E+04 1.02E+04 2.16 2.18E+06 i KR-85 3.50E+02 0.70 7.48E+04 3.90E+02 0.08 8.34E+h4 ) KR-87 1.23E+02 0.25 2.63E+04 1.83E+04 3.90 3.93E+D6 KR-88 3.12E+02 0.63 6.68t+04 2.79E+04 5.93 5.98t+06 XE-131H 3.72E+02 0.75 7.96E+04 6.24E+02 0.13 1.34E+05 XE-133M 4.35E+02 0.87 9.30E+04 1.88E+03 0.40 4.02E+05 l XE-133 3.90E+04 78.31 8.34E+06 9.88E+04 20.97 2.11E+07 ) XE-135M 1.27t+02 0.26 2.72E+04 1.59E+04 3.37 3.40E+06 1.62E+06 j XE-135 1.62E+02 0.33 3.47E+04 7.59E+03 1.61 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 A ' l SU8 TOTAL 4.11E+04 82.54 8.795+06 1.82E+05 38.56 3.89E+07 TOTAL 4.98E+04 100.00 1.07E+07 4.71E+05 100.00 1.01E+08 NOBLE GAS 4.73 4.73 4.73 0.63 0.63 0.63 . IODINE RATIO
**THESE pCi/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS O
29.0 2042c
bl ImL @ E gg Radiological Controls Department 6610+PLN-4200.02 Title Revision No. TMI Emereeney Dose Calculation Manual (EDCM) O DAMAGE DAMAGE CLASS 6 CLASS 7 pC1/cce* Percent Curies WC1/cc** Percent Curies 1-131 1.55E+05 7.18 3.31E+07 3.04E+05 7.13 6.50E+07 1-132 2.28E+05 10.55 4.87E+07 4.54E+05 10.66 9.72E+07 I-133 3.35E+05 15.56 7.18t+07 6.70E+05 15.71 1.43E408 1-134 3.90E+05 18.10 8.35E+07 7. 80E+05 - 18.31 1.673+08 1-135 3.04E+05 14.11 6.51E+07 6.08E+0$ 14.26 1.30E+08 SUBTOTAL 1.41E+06 65.50 .3.02E+08 2.82E+06 66.07 6.03E+08 KR-85M 5.00E+04 2.32 1.07t+07 9.90E+04 2.34 2.13E+07 KR-85 5.50E+02 0.03 1.18t+G5 7.51E+02 0.02 1.61E+05 KR 9.12E+04 4.23 '1.95E+07 1.82E+05 4.28 3.90E+07 KR-88 1.38t+05 6.42 2.96t+07 2.76E+05 6.49 5.92E+07 XE-131H 1.63E+03 0.08 3.50E+05 2.90E+03 0.07 6.20E+05 XE-133M 7.65E+03 0.36 1.64E+06 1.49E+04 0.35 3.18E+06 XE-133 3.38E+05 15.68 7.23E+07 6.37E+05 14.95 1.36E+08 XE-135M 7.89E+04 3.66 1.69E+07 1.58E+05 3.70 3.378+07 KE-135 3.73E+04 1.73 7.98E+06 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 7.44E+05 34.50 1.59E+08 1.45E+06 33.93 3.10E+08 9.12E+08 TOTAL 2.16E+06 100.b0 4.61E+08 4.26E+06 100.00 NOBLE GAS 0.53 0.53 0.53 0.51 0.51 0.51 IODINE RATIO "THESE pC1/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS 30.0 2042c
,, ENuclear ,,
Radiological controls Department 6610-FLN-4200.02 ( Title i Revision No. TMI Emergency Dose Calculation Manual (EDCM) 0 i DAMAGE DAMAGE CLh!s 8 CLhss 9 ucl/cc** Percent Curles pC1/cc** Percent curies l I-131 3.04E+05 7.13 6.50E+07 3.04E+05 7.13 6.50B+07 I-132 4.54E+05 10.66 9.72E+07 4.54E+05 10.66 9.72B+07 2-133 6.70E+05 15.71 1.43E+08 6.70E+05 15.71 1.43E+08 2-134 7.80E+05 18.31 1.675+08 7.80E+05 18.31 1.675+08 ! 2-135 6.085+05 14.26 1.30E+08 6.087.+05 14.26 1.30B+08 l SUBTOTAL 2.82E+06 66.07 6.03E+08 2.82E+06 66.07 6.03E+08 KR-85M 9.98t+04 2.34 2.13E+07 e.98E+04 2.34 2.13E+07 ; KR-85 7.51E+02 0.02 1.61E+05 l 7.51E+02 'O.02 1.61E+05 l KR-87 1.82E+05 4.28 3.90E+07 1.82E+05 4.28 3.90E+07 KR-88 2.76E+05 6.49 5.92E+07 2.76E+05- 6.49 5.92E+07 XE-131M 2.90E+03 0.07 6.20E+05 2.90E+03 0.07' 6.20E+00 XE-133M 1.49E+04 0.35 3.18E+06 1.49E+04 0.35 3.18E+06 ) XE-133 6.37E+05 14.95 1.36t+08 6.37E+05 14.95 1.36E+08 j XE-135M 1.58t+05 3.70 3.37E+07 1.58E+05 3.70 3.37E+07 ] XE-135 7.45E+04 1.75 1.59E+07 7.45E+04 1.75 1.595+07 1 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01-l
.1 SUBTOTAL 1.45E+06 33.93 3.10E+08 1.45E+06 33.93 3.10E+08 .
k TOTAL 4.26E+06 100.00 9.12E+08 4.26E+06 100.00 9.12E+08 NOBLE CAS 0.51 0.51 0.51 0.51 0.51 0.51 IODINE RATIO
**THES3 pC1/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF $6,595 GALLONS 4
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Nuclear T,1 Radiological Controls Department 6610-PLN-4200.02 s_- Title Revision No.
'TMI Emergency Dose calculation Manual (EDCM) 0 j DAMAGE CLASS 10 l uci/cce* Percent curies 1-131 3.04E+05 7.13 6.50E+07 1-132 4.54E+05 10.66 9.72E+07 1-133 6.70E+05' 15.71 1.43E+08 4 1-134 7.80E+05 18.31 1.675+08 1-135 6.08t+05 14.26 1.30E+08 l
SUBTOTAL 2.82E+06 66.07 6.03E+08 l KR-85M 9.98E+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-131M 2.90E+03 0.07 6.20E+05 XE-133M 1.49E+04 0.35 3.18E+06 1 XE-133 6.37E+05 14.95 1.36E+08 i XE-135M 1.58t+05 3.70 3.37E+07 ; XE-135 7.45E+04 1.75 1.59E+07 l , XE-138 1.70E-01 0.00 3.64E+01 l SUBTOTAL 1.45E+06 33.93 3.10E+08
\s.
TOTAL 4.26E+06 100.00 9.12E+00 NOBLE CAS 0.51 0.51 0.51 IODINE RATIO
**THESE pCi/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF $6,595 GALLONS 5.3.3.5 Calculation of Radionuclide Mix Percentages - Once the computer has determined the total amount of combined activities or curies for a certain NRC Damage Class, these curies are then normalised to 1004, i.e., the percentage of-each radionuclide in the total 1 mix is calculated. These percentages are then displayed on the screen along with the NRC Damage Class, 1 - 10.
Exceptions - The above percentages are replaced.in cases where an i assumed mix is more appropriate. These cases are:
- 1. Contingency calculations' fore
- a. Spent Fuel Accident in the Fuel Handling Building -
FSAR mix is assumed.
- b. Fuel Cask Accident in the Fuel Handling Building - FSAR mix is assumed. ,
[N _ c. Spent Fuel Accident in the Reactor Building - FSAR mix is assumed. 32.0 2042c-
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=~r Radiological controls Department 6610-P1,N-4200.02 Title- Revision No.
TMI Emergency Dose calculation Manual (EDCM) 0
- d. Waste Gas Decay Tank - FSAR six is assumed.
M
- e. LOCA in Reactor Building using NRC Damage Class Default concentration.
- f. OTSG tube rupture directly to atmosphere using NRC Damage class Default concentration.
- g. OTSG tube rupture via condenser off gas using NRC Damage class Default concentration.
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l 33.0 2042c
. .RCS INDICATED PRESSURE (psig)
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Radiolooleal controls Department 6610-PLM-4200.02 ; Title Revision No. TMI Emergency Dose calculation Manhal (EDCM) 0 ) l 5.4 Radiation Monitoring System (RMS) Source Term Calculation - This section of the program allows the user to determine an effluent source term from . readings on the TMI-1 Radiation Monitoring System. . 5.4.1 Only those RMS channels available for_a selected release pathway '{ are of f ered to the user. These are listed in Section 6.2. To J calculate a source term from a RMS reading the following ] parameters are used:' q
- 1. RMS READING: CPM, mR/HR, OR CPM / MIN l
- 2. RMS CHANNEL EFFICIENCY RELATING TO THE CALIBRATION NUCLIDE ' i CPM /pCI/CC ;
1
- 3. THE METER RESPONSE FACTOR i
- 4. THE NRC DAMAGE CLASS MIXTURE I
- 5. THE RELEASE FLOW RATE 5.4.2 In order to gather the above information the program will proceed in the following manner.
- 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 l decays the eventual mixture based on the hours input by the . l
- user. l l
~
- 2. The appropriate radiation monitors for the pathway chosen J are then displayed. The user then chooses the radiation monitor that is 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 prompt is for the flow rate'for the release .
I pathway. The applicable flow rates are discussed in the ] effluent flow rate section. Appropriate default values are l also listed. l
- 5. Once the above data has been entered, the program will use the RMS reading, the particular radiation monitor's efficiency, the monitor response factor, the nuclide i' fraction from the isotopic percentage section of the program i
relating to NRC damage class determination, radionuclide mix.
- . 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.
\ l 35.0 2042c l
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I lp - Nuclear ,, Radiological Controls Department a~r 6610-PLN-4200.02 l i Title Revision No. TMI Emeroency Dose calculatien Manual (EDcM) 0 l 1 The calculations are performed in the following fashion
~
5.4.3
- 1. First the total monitor response factor is calculated by l roultiplying the . individual nuclide percentages from the NRC . j damage class determination by the individual nuclide. monitor response factors.
a
'15 M= E P.* In 1 i i
Nheres M = total monitor response factor P = individual nuclide percentages from NRC damage class I, = individual nuclide monitor response factors l 1 r The I,'s for the various RMS detectors are listed as i l
- f. follows:
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Nuclear ,,,
=~r 1 RadioloQical Controls Departnent 6610-PLN-4200.02 I
\ Title Revision No.
i' TMI EmerQency Dose Calculation Manual (EDCM) 0 INDIVIDUAL MIXTURE RESPONSE FACTORS (1,) Beta Scint. Detectors RM-A2Lo, RMA4Lo, 1 Scintillation Ion ON Tubes RM-A6Lo, RM-ASLo, Detectors Chamber RM-ASHi, RM-ASHi(3), RM-ASLo, RM-A9Lo(4) i NUCLIDE_ RM-G26 & RM-027(1) RM-024(2) RM-A9Hi, RM-025 RM-A15Lo ! Kr-85m 70.7 212.2 2.35 1.92 j 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 j i Xe-131m 9.01 4.88 0.054 0.0 j Xe-133m 18.6 35.15 0.378 0.0 J Xe-133 0.0 90.24 1 1.0 (<80 kev) Xe-135m 193 195.12 2.16 0.0 Xe-135 111 221.95 2.54 2.59 ) Xe-138 1560 939.02 10.41 4.62 2-131 172 240 2.66
- 2-132 1030 747.8 8.286
- J I-133 274 219.51 2.432
- 3-134 1100 542.44 6.011.
- I-135m 706 341.46 3.784
- i (1) E MeV* Dis Nuclide E HeV* Die cal Nuclide (calibration isotope is Kr-85, threshold set to exclude Xe-133 at 80 kev.)
(2) In Probability Nuclide It Probability Cal. Nuclide (calibration nuclide is Kr-85) (3) Et Probability Nuclide Et 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
/',,s E Beta decay probability
- Beta end pt. energy cal. nuclido (cal. isotope.
is Xe-133) L '\ *Radiciodines filtered cut prior to noble gas channel 37.0 2042c E o , e - . - w. - - e w
s Nuclear ,,,
,,Radiologitw dentrole Department 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose calculation M,anual (EDCM) 0
- 2. The noble gas source tors in pC1/sec is now calculated using the following seguation and input data:
Ngst; = ( 1
- ACT
- _\_
- Ng/100) * (Flow)
- 472 M h4 Where Ngst = Moble Gas source term in pCi/sec.
M = Total monitor response factor. ; het = eps, mR/hr, cpm /sta reading from the monitor. ' a He =1 Monitor sensitivity in epm /pC1/cc, mR/Hr/pci/cc, ; or epm / min /pi/cc. Flow = Flow rate in CFM t 472 = cc/sec/ CPM.
"g = Sum of Nobic gas percentages from the selected ,
NRC classification.
- 3. The radioiodine source term in pi/sec is then calculated using the . ,
noble gas source term and the noble gas to iodine ratio, as discussed in Section 5.8. , 4 Rist = Ngst
- Ri
- 1 Ng Tfcf (if apolicable) l Where Rist = Radioiodine source term in pCi/sec.
Ri = The radiolodine to noble gas ratio. Ng ; L Tfcf = Two phase steam factor - If this is a steam release with water; the wrter will tend to keep the radioiodines in solution. ;
- 4. The noble gas and the radiciodine source terms ate _then multiplied by i the individual isotopic percentages of the NRC danage class mixture to determine the pC1/sec of each of the 15 nuclides, 10 noble gas and 5 radiolodines. i i
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) Title Radiological contrals Department 6610-PLN-4200.02 Revision No.
i TM1 Emergency Dose Calculation Manual _(EDCM) 0 ; l 5.5 Post Accident samples source. Te rm osiculation - One option of the RAC program is to use actual plant of flaent sample results to develop the l release source terms. This is in f act the preferable method for estimating i release quantities if the samplo r esults are available since the method , oltninates some of the built in censervatisms of using monitor readings, or .] cor.tingency calculations. Using-sample results also eliminates errors.in i I the source term when the actual telease mixture is dif ferent from the ass amed mix. The routines which allow use of t.he post accident samplee 1 contained in the RAC programs ptovide the menu selectors to call the dif ferent subroutines for each t)pe of poet accident sample. L
- 5.5.1 One menu selection to tue sample station / method to be used. For the Reactor Buildi ag.thtee options are presented
- 1) CATPASS' ;
(Containment Atmos phere Post Accident sample System), .
- 2) Marinelli/profilter imarinelli with a particulate and iodine filter upstream), )r 3) MAP-5, Radioiodine Processor Station.
For the condenser of f-gas or the Aux /FHB release pathways, only the Marinelli/profilter and MAP-5 samples are available. A menu will appear on the computer screen which wtl1 list the two or three available sample methods and prompt with ' enter choice'. l I When a choice - 1, 2, rar 3 as appropriate ;is entered the ' j
"' program will continue.
]
5.5.2 The MAP-5 program will' prompt for each of the identified I radiciodine species in the silver soolite sample from the MAP-5 -I Processor station. It will place the value for each of'the five J radioiodine species into one of the elements of.the five element .i
- array in microcuries por ec. It will then sum the results and I print out the sum. The user is provided the options to decay the mixture from time of shutdown and.from. time of the sample. {
NORMALLY THE DECAY CMAECTION WOULD NOT BE APPLIED FROM TIME OF : SHUTDOWN SINCE THE AllALYSIS ITSELF ACCOUNTS FOR TRAT. If the' , release is from the Heactor Building, an option selection is i provided to determine if the Reactor Building is isolated or not, and if the release is proposed or in progrees. It then will adjust the noble gases. since the MAP-5 only provides
- information on the radioiodines, the expected. ratio between the !
, iodines and noble gosos is used to approximate the noble gas activity. Since thu MAP-5 is downstream of the charcoal filtears, if the charcoal fil':ere are ef fective, the noble gas activity' will be increased bf a factor of ten to account for the filtor
. re, duction of the iodines. Based on the release rate in CFM the 1 isotopic concentrations in pCi/cc are converted to a releases rate in pC1/sec for the final source term. f 1 39.0 2042c l
- i
j N ggg 1 / Radiological Controla Department 6610-PLN-4200.02 Title Devision No. t I _ TM! Emereeney Do_se Calculation Manual (EDCM) 0 l 5.5.3 The CATPASS program prompts for the 10 noble gas and the five radiciodine nuclides identified from air sampling. The noble gas f' and iodine activities in microcuries per cc are put into en array. Options are then provided for decaying the mix from the sample to dose projection time. Since the CATPAss only= applies to the containment, the options are again provided to select l 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 now. isotopic percentages and the activities are changed frosa pci/cc to pCi/sec based on the release rate > defined to arrive at the final source tera. 5.5.4 The marinelli program is called if the marinelli/profilter option , is selected. This option is available for all three release ' pathways. Since the production of the source term is based on , the measured isotopics, as did the CATFA58 program, the marinelli i program proceeds in an identical mannov. e
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p Nuclear ,,1 Radiological Controls Department 6610-PLN-4200.02 __ Title Revision Ro.
]
*:MI Emergency Dose Calculation Manual (EDCM) 0 j 5.6 Contingency calculations Source Term Generation - The contingency I calculations attempt to determine a source ters based upon a prioritised 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 i the FSAR default parameters, are replaced with real+ time accident i 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 utilised to select: l
- 1. Secondary side Release
- 2. Reactor Building Release l
- 3. station Ventiilation Release ;
The " secondary Side Release" includes accidents that result in ' release via The condenser off-gas, the atmospheric dump valves, the main steam reliefs, and a main steam line rupture. j i The " Reactor 3uilding Release" includes accidents that result in i a release from the Reactor Building vias the purge duct, when ; the purge valves are open, or design basis leakage, when the ; purge valves are closed. , i The " station ventilation Release" includes accidents that result h in a release from the Auxiliary or Fuel Handling Buildings. 5.6.2 A " secondary side Release" is calculated by identifying four parameters: j I 6 l 1. RCs Activity [D1] pCL/cc i
- 2. Primary to secondary Leakage [D2) gym
- 3. Transport Fraction [D3)
I
- 4. Two Phase Release [Tfcf) 5.6.2.1 The "RCS Activity" is determined utilising ,
?
- 1. RM-L1 High [A1] cpm, D1 = A1/22.2 pC1/cc i
- 2. RW-L1 Lo (Ali epm, D1 = A1/1270 pCi/cc !
- 3. Host recent RCS sample, in pC1/ce, or 5
l 41.0 2042c
b$h N I gg
} Madiological controlo Department 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM)' 0
- 4. Def ault to a RCs concentration dependent on the NRC Damage Clase RC5 Default hRC Damage Clase WCL/cc 1 1.985-01 1A 4.86t+00 (Using default spiking factors)-
2 4.99E+03 3 2.498+04 4 .4.98E+04
$ 4.71E+05 6 2.16t+06 7 4.26E+06 8 4.26t+06 9 4.26E+06 10 4.26E+06 5.6.2.2 The " Primary to secondary Leakage" is determined utilising
, 1. RCS identified leakage [D2] gpm
- 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 release via the condenser off-gas. For a release through the condoneer off-gas the noble gas transport is 1.00, the radiciodine transport fraction.ie 0.0075. The radioiodine transport fraction is a product of: The fraction of radioiodine entering the 0T80 from the RCS that is a volatile iodine species (.05) and the partition factors for volatile iodine species in the main condenser (.15). Non-volatile iodine species have a partition factor of sero in the condenser off-gas. For a release direct to atmosphere the noble gas and radiciodine trsnoport f ractions use the fraction of steam released to total steam flow from the 0T80. An additional partition factor (Ticf) is applicable for a two phase direct release. The resultant source terms in pCi/sec are calculated by: Ngot = D1
- D2
- Ng/100
- 63.09 i Riot = D1
- D2
- D3
- RI/100
- 63.09
- 1/Tfcf.
5.6.3 A " Reactor Building Release" le 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 42.0 2042c
ENuclear m -- Title- Radiological Controls Departnent ; 6610-PLN-4200.02 I Revision No. 7N! Beeroency Do_a_e Calculation Manuel-(EDCW) 0 3. Building spray statusTransport Fraction, E4 = 0.1 or 1 0 depending on th j , 4.
. Rolease ylow Rate CPMs E3 = flow
- 412 to convert CPM to ec/eec 5.6.3.1 The RCs Activity is determined utilising:
1. RM-L1 High Channel (A1) cpap [A2)-= [A1)/22.2 pC1/cc 2. MM-L1 Low Channel (A1) cpm; (A2) ='[A1)/1270 pCi/cc 3. Representative RCS sample results (A2) in pC1/cc 4. Default Class: to a RCs concentration dependent on the NRC Damage _ NRC Damage Class RCs Default pCi/cc-1 1A 1.98E-01 2 '4.86E+00 3 4.99E+03 4 2.49E+04 5 4.98t+04 6 4.71E+05 7 2.16t+06 6 4.26E+06 9 4.26E+06-10 4.26t+06 4.26t+06 4. Default identified. six according to core condition as previously Reactor Bui1 Jing is calculated as follows:The noble gas and radicio El = A2
- A3
- Ng/100
- 3785/5.6E10 pCi E2 = A2
- A3
- Ri/100
- 3785/5.6810 pC1 5.
The RCS leakage to the Reactor Building is determin e db roguesting the " total gallons of RCs leakage into they Rs" The transport fraction is determined on the basis eof th status of Reactor Building spray. fraction is assumed to be 1.00. The Noble gas transport transpori fraction for instantaneous radiciodineis 0.5. . The reduc An additional adjustment of the radiciodineplateout is Reactor Building spray is activated. concentration in the Reacto iodine an assumed concentration 90t. in the Reactor Building ere atmosphThe by spray reduce 43.0 2042c
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[\ Title Radioloalcal controls Department 6610-PLN-4200.02 Revision No. t "D(I Emeroency Dose calculation Manual (EDCN) 0 i P iodinc concentration in the Reactor Building atmosphere by an assumed 90%. i o
- 6. 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 R3 leaktate adjusted for actual Rs internal t 3= pressure as indicated on PT-291. , 5.6.3.2 If the accident type is a Fuel Handling Accident in the Reactor j Building, then the number of. damaged fuel rode is identified by ; the " user" or an FsAR default condition is used. ;
't El = 1.7
- Num rod /208 E2 = 0.05
- Num rod /208 5.6.4 A " station Ventilation Release" is calculated fors
- 1. LOCA in Auxiliary Building' l
- 2. Waste Gas Accident ,
l
- 3. Fuel Handling Accident in the Fuel Handling Buildirg,.
including.E8F Fuel Han611ng Building releases. 5.6.4.1 If a grab sample is available from the affected area, it is used to determine the source term. If not, the accident is modeled by: 1
- 1. Determining the number of damaged fuel rods for a fuel f handling accident in the Fuel Handling Building.
Ngst = 4.2E6
- Num rods /56 pCi/sec Riot = 750
- Num rods /56 pCi/sec i
-t 1
l l 44.0 20A2c 'I i J 1
-__________ _ _ ____ ___ _ ._- _ - . - _ _ _ . _ _ . _ . . _ . . _ _ - ~ , . _ .
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& UV ggg Radiological controls Department 6610-PLN-4200.02 j Title Revision No. ]
l TN! Emergency Doce Calculstion Manual (EDCM) 0 l 1 I
- 2. Using FSAR assumed conditions for a cask-drop accident.
Ngst = 1200 pCi/ sue 1 Riot = 450 pC1/sec
- 3. Determining the curies released for a Waste Gas Accident.
Typical source term based on a' typical inventory of j 1 Ngst.= 1.0E9/Dr FCi/sec- ] Riot = 1.055/Dr pci/sec or
]
FSAR worst cases Ngst = 1.0E10/Dr pC1/sec ; 1 Rist = SE6/Dr pCi/see
]
Whores Dr a duration of release. ) l l l l L t l !
'l L
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I i 1 i ' I TMI [.N Radiological controls Department 6610-PLN-4200.02 l l ( Title Revision No.
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TMt Emergency Dose Calculation Manual (EDCM) 0 1 5.7 Decay Scheme calculation - The user has the option to (1) decay the postulated mixsure from the time of reactor shutdown to time of the dose i projection or- (2) decay sample data from the tinae the sample is obtained to .! the time of does projection. Subroutine (decay) only decays forward in 'I time. This oubroutine (decay) adjusts the individual nuclite percentages according to the conventional exponential decay equations j A = A, exp (* *) 5.7.1 -Fifteen isotopes are decayed according to the equatton N(w) = I(w)
- EXP (-decay _ time
- f(w)) ]
4 where:- 2(w) = posculated isotopic percentage l decay time a user input time f(w) = isotopic decay constants read from data files j
~
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 = 1(11) - N(11) N(5) = 0.88
- S1 + N(5) ;
where S1 = amount of I-131 decayed 0.88
- S1 = amount of Xe-131M buildup j 5.7.1.2 For Xe-133M the equations are 81 = I(13) - N(13) J l
N(6) = 0.02
- S1 + N(6) where: I l
S1 = amount of I-133 decayed 0.02 * $1 = amount of Xe-133m buildup i N(7) = 0.98
- S1 + N(7) calculates the amount of Xe-133 I buildup from I-133 I i
i l 46.0 '2042c-
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l 1 ENuolear ,,, Radiological' controls Department
=~r 6610-PLN-4200.02 l
j Title Revision No. j
. \
TW1 Emergency Dose calculation Manual (EDCM) 0 I I 5.7.1.3 For Xe-135m and Ae-135 the equations are 31 = 1(16) - N(ll). ) i N(8) = 0.3
- S1 + N(8) ;
N(9) = 0.7
- 31 + N(9) ,
where: f 4 ! l 51 = amount of 1-135 decayed 0.3
- S1 = amount of Xe-135m buildup 0.7
- S1 = amount of Xe-135 buildup l 5.7.1.4 The isotopic percentages are recalculated aos l 1(w) = N(w)/ Sum (N)
- 100:
wheres g N(w) = adjusted / corrected postulated isotopic percentages som (N) = sum of the fif teen ' isotopic percentages 1(w) = final isotopic percentages based upon 100. . t 6 i ) i
,f 4
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N TMI ,, (' Title Radiological controls Department 6610-PLN-4200.02 Revision No. TMI Emeroency Dose Calculation Manual (EDCM) 0 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 NOSLE GAS TO IODINE RATIO. The uses of this ratio are discussed below. An airborne reisase from a nuclear power plant will primarily consist of noble gases and radiniodines. Except in the most i severe and improbable accident ecenarios, radioactive : particulates are not expected to bo important dose contributors.- 1 The RAC program was designed to~ incorporate ten noble gases and- ! five radiciodines. 2 The 15 isotopes are considered to be the most radiologically significant gaseous isotopes available for release from an ; operating nuclear power plant. Pertinent radioactive decay parameters such as half life, average gamma energy per disintegration and average beta energy.per disintegration for i each-isotope are stored in data statements within the program.
. Along with individual isotope source term:information, this data.
is used to determine dose rate conversion factors and dose-rates that are specific to the isotopic mixture being released. These i calculated quantities can be adjusted to account for radioactive ) decay during the accident sequence. 5.8.2 The THI-1 RAC model always projects both thyroid and whole body 't I dose rates at specified downwind distances. Consequently an estimate of the isotopic release rate is necessary for both , iodines 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 processes 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 l circumstances where the release rates of specific isotopes may be sero or negligibly small. But, in general, the program accounts for the fifteen isotopes listed above. I In certain circuastances it is not possible to obtain relvaeo rates for all fifteen isotopes individually. For example, some plant offluent monitors have only noble gas channels while othere i have particulate, iodine and gas channels. The MAP-5 sampling l system yields only iodine information, where the CATPASS and the l Marinelli gas sampling systeme yield information on all fifteen 1 isotopes. For release pathways where information on both noble I 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 1 l i l 48.0 2042c ) l
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, , . __.__._,m. _, - _ _ . . . , . - , __, ____._m_______I
I I i i Nuclear ,,,
"-- i Radiological Copt,rcle Department _
6610-PLN-4200.02 j s_s Title Revision No. TMI Emergency Dose Calculation Manual (ED';M) ,___ 0 , i A certain type of reactor acciden: has occurred. Based on an j assessment of the degree of core damage and the accident type, j the computer selects a default mixture of 15 nuclides and j l calculates the fraction of the mix that each lootope represents. J The noble gas to iodine ratio is also calculated. Assume t'nat the retto was equal to 5/1 in this case. Also assume that an iodine sample was taken which indicated a total radioicdine ; release rate of 5000 pC1/sec.- Using 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 j pC1/sec. It would also calculate individual noble gas release i rates by using the isotopic fractions from the default mix. ) l 5.8.3 To summarize, the highest quality information available is a' l quantitative measurement of each nuclide. This type of , information is available from RCs, gas Marinelli and profilter, j and CATPASS samples. So there is no need to invoke the noble gas-. to iodine ratio in these cases. The second best measurement -)
. would be one that yielded gross noble gas and gross iodine ;
readings. This situation occurs in the low range radiation , monitore which have individual noble gas and iodine channels. p Based upon the default mixture fractions, the release is : I apportioned among the fif teen nuclides 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 iodine measurements are not available, for example, when only noble gas or only iodine information is available. i.8.4 There are some refinements and subtleties that the program user 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 ratio. As. explained elsewhere in i this manual, the program also corrects the Dose Rate Conversion ' Factor.(DRCF) for decay of the. isotopes in the mix. When l performing dose projections several hourspor more after the reactor has tripped, these two decay corrections can ! significantly alter the resultant projections. The computer operator is given the option of whether or not to account for decay between reactor trip and done projection. r I i l 49.0 2042c
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ENuolear ,M1 Radioloolesl controls Department
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6610-PLN-4200.02 Title Revision No. TMI smorgency Dose calculation Manual (EDcM) 0 l l ( 1 1 5.8.5 For dose projections based upon RMs readings, the decay I correction to the noble gas to iodine ratio is straightforward. 1 A default mixture of the fifteen noble gases and iodines le ! selected based upon an assessment of core damage. If the j computer operator elects to decay the mix, he is prompted for the ! ! decay time between reactor trip and dose projection. All fifteen j l isotopes are decayed by the standard exponential decay law and ) i the noble gases and lodines are totaled separately so that their ! ratio at dose projectitn time can be calculated.- (Ingrowth of 1 menon isotopes from decay of iodine is accounted for.)- The decay l adjusted ratio can ther, be used to fill in the missing noble gas l or iodine information,.as explained above. , 5.8.6 When iodine samples are taken at'the MAP-$ stations & two step l decay process is used. As above, a default mixture is chosen, j based upon the NRC core damage classification. For a dose . - calculation based upon a radioiodine processor sample, if decay' l correction is desired, the user is prompted for two decay time j i intervals: ! 1. The time between sampling and dose projection ,
- 2. The time between reacter trip and dose projection ;
( 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 default mix is 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 radioiodine 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 radiciodines from a sample. Each has been decayed to the dose projection time. A word of caution should be added at this point.- The lodine released in certain types of accidents may be reduced by various chemical and physical processes such as iodine plateout or ; formation of water soluble iodide salto. The noble gas to iodine i ratio, as calculated above, may not account for this iodine reduction. As a consequence, the ratio, based upon the default mix, may be too low. -This creates the potential for _ i 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 , agreemer't for thyroid doses, but net f or whole body doses. 50.0 2042c I
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Title Radiologleal controls Department 6610-PLN-4200.02 Revision No. TMI Emergency Dose calculation Manual (gDcM) 0 5.9 Ef fluent Release Flow R6tes - Flow rates for ef fluent releases to the environment are divided into four categories: , i
- 1. Normal ventilat.on flow rates. ,
- 2. Reactor suilding leakage flow rates. ,
- 3. Adjacent momentum plume rise (station vent and reactor purge ,
concurrently releasing).
- 4. Flow rates for 0780 tube rupture release directly to atmosphere.
. . .i
- Buoyant plume rise l 1
- source term calculation using RMG-26 or RMG-27
- source term calculation using a contingency calculation !
These flow rates for accident source terms to the environment are calculated as follows:
/'
( 5.9.1 THI-1 Normal Ventilation Flow Rates ' The Unit 1 RAC Program provides the -option to use the actual ventilation flow rates as read from the flow recorders or to use =i default flow rate (s). Each normal plant flow path has ! predetermined flow rate ranges, and assigned flow recorders as follows: 4
- 1. Reactor Building Purge
-FR909 0-20,000 CFM; Low Range-
-TR148B 0-50,000 CFM; 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
-FI283 20-200 CFM .
-FI284 100-1000 CFM
-Tatal 5-1250 CFM y- s 4. Kidney Filter System k_,/
s
-AHE-101, AH-F-12 P/I filters 20,200 SCFM _
51.0 2042c , h
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y 1 ENuclear ,, Radiological controls Department 6610-PLN7 4200.02
. Title 3 Revision No..
TMI Emergency Dose Calculation Manual (EDCM) 0
- 5. Auxiliary. Building Exhaust?
- m. ,
-FR150L 0-100,000 CFM !
- 6. Fuel. Handling Building Exhaust
-FM149i '0-50,000 CFM-7.' . Auxiliary.and Fuel Handling Building Exhausts.
-FR-151 0-150,000 CFM~ ..
l'
- 8. Condenser Off-Gas Exhaust- ]
-RMR15 Recorder FT-lll3 Ch. A 0-200 CFM- ;
- 9. - ESF' Fuel Handling Building Exhaust
-No Flow Recorder at this time 0-8000 CFM Default. values are.used in the RAC Program.when a small value or j
~
an unknown value is required as input to a dose projection. The default-values'are 5000 CFM - Reactor Building Purge ' Reactor Building Purge and Make-up. Exhaust I Auxiliary. Building Exhaust
+2el Handling Exhaust
- .ESF Fuel' Handling Building Exhaust !
- ; Auxiliary and Fuel Handling. Building. Exhausts.
40 CFM - Condenser Off-Gas
-These default values allow the user to continue'with dose h projections'aven though a value'is small or unknown. ~ Therefore, ,
once the dose projection is complete,' the results may be ratioed- j up or down-depe. ding on the situation. For example, if the default value of 5000 CFM was used for a Reactor Building Purge , snd subsequently a Tech. Functions calculation was performed { indicating 1000 CFM flow.; The dose projection could be ratioed o
- i. down by one fifth (1/5).- Therefore, a dose projection of -'10 mrom
. q would then be approximately 2 mrom based on the reduced flow.
calculation, realizing that the X/Q will'also be affected by reducing flow.
.i 52.0 2042c 1 1
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V yg - Radiological' controls Department' 6610-PLN-4200.02
' Title Revision.No.
TMI Bmergency Dose calcalation Man'ual (EDCM)' O ,, 1
^?
5.'9.2 Reactor Building Leakage Flow Rate. .
~
Another.section of this program calculates a leakage flow rate out of the Reactor Building based on Reactor Building pressure. The Reactor. Building pressure indicator is PT-291, 0-100 peig, J
.lc.:sted on control room panel CR.' The leakage out of the Reactor Building is based on the amount of pressure in the Reactor..
Building with all penetrations' closed. The following. equation is used to calculate the Reactot Building Leak Rates . L
- t. L, = L3*SQRT(P,/Eg ) ,
i- where: - L, = Reactor Building Leak Rate in CFM - 1 L3 '= Maximum allowable integrated leakage rate at. Pg' - Lg = 6.14 CFM . i Pa = Peak' Reactor Building internal pressure at. design bata, accident, Pg = 50.6 poig - l= P, =1 1 Actual Reactor Building internal pressure 11n'
\ Psig-
, Therefore,-the maximum:}itkage allowed at:a design basis accident-l pressure of 50.6 peig is'6.14 CFM._' Leak rates-atz 0-60.psig can be calculated from the above formula. -The default value-in this * , subroutine-is 50.6 peig. cA graphic' representation follows in. l Figure 5.?-1. 5.9.3 Adjacent Homentum Plume Riseg(Station Vent and Reactor Purge)' F I. I For an isolated stack, either the station vent or the Reactor < Purge,-the stack gas exit velocity can be calculated from the-l flow rate according to tho following formula i 4 V (1) w. nr2 { where w = stack gas exit velocity e
- V = flow rate or.valume flux r = radius of staet 4
)
d' k 53.0 2042c 4 4
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r 9 Title-TMI Radiological Controls Department' 6610-PLN-4200.02 Revision No. TMI Emeroency Dose Calculation Manual (EDCM) , 0-The station Vent and the Reactor Purge stack are situated close j enough' together that their plumes will mix as the plumes rise. f
.( iFor'two or more adjacent stacks that have difforent exit' , ;
velocities, the offact of mixing on tho' exit velocities of; non-buoyant plumes can be given by the following formulas (2)[ E wV I w-Ev-l i b where w = exit velocity due to mixing k ~ If these adjacent stacks were anodeled as a single stack,, the - p l radius of the stack would be given by: U (3) r= _- > rr w i At THI-1, the reactor building stack and stationivent are 9* adjacent stacks. For computing plume rise, the stack gas exit
= velocity and stack radius were calculated according to Eq. (2)--
and (3) above. -A comparison ofs the adjacent plume rise with the
=
4 plume rise from the individual stacks is shown in Table 5.9-1.- l 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). THI-l has 22 main steam rel'ief and atmospherii dump. valves. Data on the valves are presented in Table 2, which lists the valve identification number, function, manufacturer,. pressure set point.
.and flow rate. The set point pressures vary from 200 psig-to.
1092.5 psig, 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.. Tne valve position. openings along with the secondary system pressure relate to a release flow- {' and plume height. The percent open for'thece'two'(2)~ valves can be read at the center control panel under the turbine bypass dump controller for MS-V-4A/B from 0 - 100%. Each of the 22 valves at THI-l has a stack or vent where the steam is ejected-into the atmosphere. The location of these < stacks is 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 flow through it will.be approximately equal to the rated flow, and the flow can be assiuned to be approximately O constant until the pressure in the steam line drops to the point where the valve reseats. For a :: tuck open valve, the pressure 54.0 2042c
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_/y i!' x 'b -; . c 4 :- GNuclear TMI ,
== - -
Radiological controls Department 6610-PLN-4200.02' Title Revision No. TM! Emergency Dose Calculation Manual'(EDCH) n decreases rapidly with time, and the flow through the valve is on?,y a small. f raction of the rated flow. ~ For either a normally operating valve or stuck opem valve, if the pressure and' temperature in the steam _iine are'known, the: conditions just 'i ' beyond the stack exit can be estimated by assuming expansion of the steam to atmospheric pressure and temperature.- 5.9.4.1 Buoyant Plume Rise
.i When the steam is released into the atmosphere, the rise of the. ;j steam plume is initially controlled by~1ts: velocity, temperature ;
and cross-sectional area. Depending on these variables and
- s. atmospheric conditions, the plume rise can1 vary from hundrede to '
thousands of feet. . Plume rise is a very'important factor in determining maximum ground' level doses. For 's 79m,: plume. rise can increase the effective stack height by a factor _of 5 to.50.- =! Since maximum ground level dose is reeghly proportional to the inverse square of the ef fective stack helget, a plume rise of 200 feet, for example, gives a ground level concentration 100 times higher than that from a plume-rise if 2000 feet. < Modeling of plume rise begins with modeling the steam condition 9 at the valve inlet.. Table'5.9-3 outlines the calcul=*ional steps required to compute buoyant plume rise,.beginning with the valve inlet. The far left side of the table identifies the area'for 'I which the calculation applies:- valve; inlet, top'of stack, jet 4. origin, and plume rise. For each area, several; quantities must 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 controls document Potentially Buoyant-Releases at TMI-1. CAUTION: In highly stable atmospheric' conditions, the presence of-layers of diff3 rent temperature air lcan cause thermal boundaries resistant to plume vertical' travel. In some conditions, a buoyant. plume may penetrate thene layers and not come down to the surface as predicted. In other cases -j the plume may be unable.to penetrate the. layer and the - j effective stack height will be reduced to the height of tho' 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)1off. .
site monitoring'will provide an indication of.the magnitude ' of the effect. It may also be possible to estimate-the effect through visual observation of the plume.
~
till> 55.0 2042c
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.- . TMI
- 1 Radiological ~ Controls Department 6610-PLN-4200 02 Title - Revision No.
n: . . TMI Ec.orgency Dose Calewi= tion Manual IRDCM) O !,i i p 5.9.4.2 Source Term Calculation Using RMG-26 or RMS-27. (see Figure. 5.9-3) z i. l- RMG-26 and RMG-27 are effective in calculating a primary to: secondary release source term direct to the atacephere whens l1. . Atmospheric Dump Values (ADV) MsV-4A or MSV-43 are.open from-0 7 1004, as indicated on control Room Panel "CC", and z releasing radioactive steam to the environnent, and/or,-
- 2. Emergency Feed Pump'(EFP)-relief-valves,.MsV-22A or MSV-22B' are open and releasing radioactive steam to the environment, and/or,~-
V ~ [ 3. EFP is in operation and releasing radios.ctive' steam from tho :l EFP' exhaust to the environment ]
- 4. Steam bypass dump to the,conden.wr through MSV-8A/8.
When the user chooses a release from an OTs0 tube rupture. f directly to the atmosphere and is using RM-G26 or RM-G27 l l- readings, the calculation Steam Flow Computation is used to l determine a release flow rate, depending on which of the valves ! are open (Table.5.9-2).- The mass flow rate from each open valve i is added up.to give a total flow rate to the environment. A-l' source term.is calculated using.the flow rates in CrH and the , RM-026/27 readings converted to concentration using the monitor j efficiencies to give pCi/second. ') NOTE: Calculation of a source term using RMS (RHG-26/27) is j dependent'on the Atmospheric Dump Valves (ADV) status. If 1 the ADV is open,-the calculation.is appropriate. If the '! ADV is closed but plant conditions-(01SG 1eakrate and core f damage) have not changed significantly, or there is other flow past the monitors as noted above, then.the use of a- - 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 7 2A/B, . then the contingency calculation applies. h q n l 1 i P ! 5 l t 56.0 2042c I e
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,, . g { - Number p TMI Radiologleal controls Department- 6610-PLN-42OO.02-
\
. Title - Revision No.
- l. TMI Emergency Dose Calculation Manual (RDcM)~ O 1
i 5.9.4.3 Source Term Calculation Using a. contingency calculation I When the user performs a contingency calculation'due to.the lack of sample results or RM-G26/27 readings, the flow rate. ' corrraponding to the set point pressure is used, if the valve > operates normally, This flow rate is given in Table 5.9-2. If, I L.
- however, the valve sticks open, and the steam generator pressure l--
is less than tho' set point pressure,.then the. flow rate is based en the tables supplied by the valve manufacturer. .These tables have been incorporated into the RAC computer code. The flow from , all the valves is totalled and modelod-as a release to the- - atmosphere.- F E
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57.0 2042c -l l l
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m 1 (ff hIj g 7_ Number Radiological ~ Controls Department 6610-PLN-4200.02 Title 4
. Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 0 '-- TABLE 5.9-1
. Adjacent Plume Rise at TMI-1 i Reactor Bldg Stack and Station Vent Stack
.- Actual Flow Characteristics - RAC Model . MIDAS -
% Reac Bad Stack Station Stack Flow Stack Plume - Flow Stack- . Plume Stack D6ameter Vent Diameter Rate Diameter R!as Rate Diameter Rise -
(cfm) ' (m), (ofm) (m) - (ofm) .(m). -(ft) (cfm) (m). . (ft) 10.000 1.1 10.000 1.7 20.000 , t.647 30.3 10.000 1.1 25.4 .; H , e 10.000 1.7 16.4
]'
10.000 1.1 120.000 1.7 ~ 130.000 1.627 - 199.2 65.000 65.000 1.1 1,7 165.4 107.0 I
$$ 10.000 1.1 10.000 1.7 20.000 1.647 24.8 10,000 1.1 22.1 ;
fif,. 10.000 1.7 16.5 N 10,000 ipp{t 1.1 120.000 1.7 130.000 1.627 85.8 65,000 1.1 75.9-65,000 1.7 57.0.
, 5?.0 2042c-
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-Radiological Controls Department 6610-PLN-4200.02 l Title Revision No.:
TMI Ernergency Dose Calculation Manuel (EDCM)' D TABLE 5.9-2 ) TMI1 STEAM GENERATOR RELIEF VALVES vehe stack - vowe , veke - D.cheroe see Point sieck ; Manutecturer Aree Diameter i Number Function Wessure l Flow Rate
, (us V) (sq. in.) (psig) (ibs/hr) (eches) 17A D Relief Vekes, Bank 1 - - Dresser /- 16 1050 - 792,617 ' 10.02 .
. Consolidated 16A D Relief VeNes, Bank 2 Dresser / 16 1000 - 600,065 l b..
. Consolidated 1060, 614.960 l ISA D Relief Vekes, Bank 3 Dresser / 16 10.02 Consolidated 20A&D Relief VeNee, Bank 1 Dresser / 16 -1050 792,617 10.02 Consolidated Relief Vahes, Bank 4 Dresser / 16 1092.5 824,260 10.02 20B&C .
Consolidated Railef Vekes, Dresser / 3.97 1040 194.620 - 10.02 l 21A&B Smel Safety Cor:Wdeted ! 22A Safety Relief,' Lonergen 6.36 200 70,211 13.13 Emergency Feed Pump ' 22B Safety Vane, Emerg. Lonergen - 6.36 220' 76,795 '13.13-(F.W,P.T. Steam inlet) - 4AAB Variable - 1010 402,792 13.13' Relief Vekes (manual) Fisher to Atmosphere i l 1 59.0 2042c 4
'3
'n N I M*f Radiological controls-Department 6610-PLN-4200.02~ Title Revision No. TMI-Emergency Dose'Calculatlon Manual (EDCM)' O i
. TABLE 5.9-3 Calculational Steps for _Cbmputing' Plume Rise-Source -
Quantity Computed input Values Nee ( 3d . ~of input Step a Valve . preneure of steem operator - q Inlet 1 steem tow rete ttwu vente : Preeeure'of steem at top a. steem b rate ' - step 1 l Top 2 I b.' intemel radius of stack - constant of of sesek below sha.eder .
' Stack, cJ enthalpy < constant.
[ choked b ) -i
' Below specac volume of steem at pressure at top of stack step 2 l Chamfer 3 :; ,
top of sis &, boiow chamler pressure at top of stack step 2 Solow 4 Temperature of steam at Chamtsr top of stack, be6ow enenwer veioeny of steam at top of a. specac volume of steem fstep3 i Also S
- b. b rate of steam " step 1 f Jet stack, below chamfer
' c. Intemel radius of stack constant Origin _
Density of steam at a. ten 1mreture of steam step 4 6 I jet origin (smbient pressure); b.' pressure of ambient air < constants Jet
' Jet radius at ortgin . a. ' density of steam . Step 6 7
- b. velocay of : team step 5 Origin (Needed for MIDAS ony,
- c. b rete of steem step 1 J used in RAC but not reaty needed)
- a. jet redus et origin step 7 8 Plume rise
- b. denemy of steam at origin step e -
Plume
- c. velocity of steem at origin step 5
- d. density of amtwent air <constante Rise
- e. wind speed at ongin operator ,
- f. 150'.33' dona T . operator s
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- CL Radiological Controls Department 6610-PLN-42OO.O?_
Title Revision No.= 4 TMI Emergency Dose Calculation Manual'(EDCM) o_ f FIGURE 5.P-2 TMI-i RB LEAK RATE VS RB PRESSURE 1
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1 FMNuclear == 6 < 9 4 1 Title Radiological controls Department 6610-PLN-4200.02 Revision No. TMI Emergency Dose calculation Manual (EDcM)' 0 l 5.10 'two-Phase steam Flow Determination -A two-phase (liquid and gas) release calculation was included'for.an OTsG tube rupture accident in response to INPO SOER 83-2 (Recommendation #12).. IMPO SOER 63-2 " steam Generator Tube. -! Ruptures" was developed based upon the steam generator tube rupture events-
.at_R. E. Cinna, Oconee and Rancho Seco. Recommendation #12 states
'" Emergency Plan. Implementation Procedures should , .
. ensure that, estimates of doses can: be made for' two-phase orfliquid releasee .through the steam generator safety relief valves."' GPUN-Corporation'is required to'.
respond to all sOER' recommendations. The calculational method used to ; Eimplement this recommendation is based 'upon the assumptions that the valve - inlet fluid condition 11e either-pure liquid or steam (as indicated by the.
-0T80 wide range level instrumentation):and followingl discharge, the steam.
-fraction is described by-assuming that there is'no change in~ total energy.
content.- If the OTSG 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, SPCA PTl and 2 or SPCB PT 1 and 2.
i 5.10.1 The program determines a two-phase correction factor [Tfcf) which. 9 is'a function of OTSG pressure in psia. . This factor is only-calculated if the OTso water level is indicating a liquid release (greater than 600 inches on-the wide range level instrument, reading). The correction factors are used to account'for the. radioiodine that-would remain'in thecliquid 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 cubsequently~used in the radiciodine - source term equation to correct'the radiolodine source term. RIST = D1
- D2
- D3
- RI/100
- 63.09'* 1/Tfcf:
__________________________________________________=-- l NgTE t '
. Increasing OTSG A/B water level willfpossibly help cut.down.
the release of radiolodine due to the partitioning'effect-of the iodine in water. Increasing.oTSG: level should bec discussed with'the Emergency Director as a means.of-reducing
~
.offsite doses. ,
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l 4 N -THI Radiological' controls Department ' 6610-PLPf-4200.02 b Title; . Revision No. , TMI Emergency Dose Calculation Manual (EDCM) 0 _{ 5.11 -Source Term Filtration - The TMI-1 RAC. Program provides the optien tof , include orl disregard reduction of source term through' filtration. The- :j source ters toduction is applied to radiciodine species only. 5.11.1 Radiciodine normally exists in chemical' forms which rare highly reactive. They readily adsorb onto surfaces and can be scrubbedi chemically free the atmosphere.: The radiciodine removal methods available in the. plant by design are the unarcoal filter banks-in the ventilation system.1 As applied in the RAC tape, anytime a sample is obtained upstream of a charcoal. filter a filter. x! reduction can be applied.. Anytime's default source' term is used,. a filter reduction and/or building spray reduction can bei applied.- I i 5.11.2 In any case where the RAC program will' apply the source term, reductions a prompt is provided by the program. . In a, general' l form, the program will prompt "ARE THE CHARCOAL FILTERS l OPERATIONAL" or "IS THE. 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; removal of. radiciodine species. 'However, the RAC program takes a conservative approach and assumes a 90% reduction from fully operational charcoal. filters. Note that;the prompt-for.either is a "yes" or "no" answer. If the charcoals are known to be degraded-(for example, due to moisture).but=still. partially, ,~ functioning, this is accounted for in the RAC model. . The filters-are either functioning at the full _ capacity - 90% or from 90% due to degradation. i 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 CATPASS system and contingancy default source terms generated from a spent fuel accident or LOCA in the Reactor Building, including RM-Ll readings, ORCS' sample . I 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 i reduce the radioiodine. source term by.a factor,of ten. However,- ', if.the source term is generated from a gas channel: reading, the :l iodine will be reduced but the noble gas will be increased'due to ) a constant meter reading and a change in the isotopic ration. l Application of the. filter reduction for lodines in the Aux /FHB'is I available for source terms generated from RMA-4,'RNA-6,' and the l contingency default' source terms from a sample result,' damaged 1 fuel rods, a fuel cask,'or a waste gas release. 'It is important l to remember that filter application can significantly change the I noble gas to iodine ratios.
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65.0 2042c
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""* r Radiological controls Department '6610-PLN-4200.02 Title Revision No.
~TMI Rmeroency Dose Calculation Manual-(EDCM) o
-Anytime a filter correction factor is applied to'the iodines downstream of the~ filters, the noble-gas sour:e term must be-
-increased, while iodines obtained upstream of a filter-will!
reduce the lodine source' term without'changieg the noble gases. Y
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"==bar f Radio?.ogical' Controls Department 6610-PLN-4200.02 ,
{' , Revision No. Title
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THI Emergency Dose calculation Manual (EDCM) 0 5.12. Noteorology Inquiry - Upon' initiation of the meteorol'ogical data input section of the RAC program, the computer places a telephone call to_ the IBM ~ PC located at the base of the TMI' met tower and requests from it the most' recent 15 minute average of,the met data stored in the met tower PC. The RAC program is able to. centinue. if the ' telephone call is not completed, the - met tower phone is busyi;or,the met' tower fails to respond'after the call has begun.- If :any of the r above ' conditions occur, the data is marked 'as _ missing by'the RAC program., After the call is terminated,-a new screen-is. .. _ presented to the user. listing the collected data in a tabular 1 format and-asking the operator for-tie 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 is -
allowed to default to the "B" sensor value.f If both "A" and "a"-sensor
- values are missing, no delault value'is presented or allowed. To obtain-the default value the operator presses the Return key while the cursor is:
l i on the first character of the input field.' 'The operator-is free to' enter' his own value.- i Af ter the Return key is prossed, the RAC program subjects .the inputted . value to limit checking. !!he limits ares l Leyet Limit Upper Limit , i
\,,j/ v nd Speed 0.5 '99 W ' .- Direction- 0- 360 Delta T -30 +30 If the entered value is lower than the lower limit or higher.than the upper limit, the operator is requested to 8.nputca new-value. An' exception is the1 ,
wind speed valuo. If the et tered value le less than 0.5 mph, the RAC program uses 0.5 mph and inf orms too user of this fact. f f i t-l
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67.0 2042c 1 (
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' Radiological' controls Department -6610-PLN-4200.02 i l Title Revision No._ '
?
'TMI Emeroency Dose Calculation Manuel-(EDCMI O- ,
?
l 5.13 . Dispersion Model - The TMI-1 RAC model computes both Whole Body Dose using ,
- s. finite dose model and thyroid dose using a semi-infinite model. Many ,
subroutines ~are called by.both models. The results of the two models are:
~
presented on a single output page.- 3 5.13.1 Finite (Whole. Body) Dose Model' , TheTMI-l'RACmodelcalculatesexternalwholebodygamma[ dose' rate using a finite model'for both ground and elevated releasee.. The finite gansna dose algorithm is-licensed from Dr. John Hamawis of Entech 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" _ multiplying.the dose rate by the expected duration'of release.' The finita gamma dose algorithm in the TMI-l' RAC model has the.
-same structure as-Pickard,;Lowc, & Garrick's~ MIDAS. finite gamma dose algorithm.- - The basis for the algorithm'is a four dimensional array -of finite gansna f actors. These finite gamma-
~
factors are pre-computed three dimensional. numerical integrations which appear in the theory of.the finite cloud model and i represent the spatial distribution of the radioactive material in y the finite plume. These' factors depend upon the plume dimensions. ; at the downwind distance of interest, the crosswind distance, the plume elevation and the average gamma energy of the.nuclide mix: in the' cloud..-They are sometimes= referred to as " gamma X/Q" in 1 the literature although they are not derived from typical X/Q-_ _; calculations. 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,'1000, 1250,'1500i -i 1750, 2000, 2250, 2500, 3000,' E00, 4000, 4500, ' 5000, 5500, . 6000, i 6500, 7000, 7500, 8000, 9000, 10000, 15000, and 20000 meters. The 6 crosswind distances-aces' 0,-50,;100, 150,1250 and 500 i meters.' The 6, heights above ground are 0, 30, 60,'100, 150,, lI and 300 meters. The ? energy groups are: . 032i 081E .15, 25, !
.53, and 1.0 Nov. TN, abundances of the noble gases; for the six':
')
energy. groups were taken from MIDAS. For offoctive release-heights'other than the 6lf1xed. heights, the-l' finite gamma factors are. extrapolated to that height by the .
- l. subroutine (Interpolate). For downwind distances other than the-28 fixed downwind dia.tances, ' tihe finite. ganuna f actor of the .
nearest fixed distance is assigned to that distance, i.e., no horizontal interpolation is done, as is consistent-with MIDAS. , The THI-1 RAC model explicitly includes the contribution of I-131, I-132, I-133, I-134, and I-135 to the external whole body y gamma dose. This method-of handling the' contribution from the( ; j radiolodines is more accurate'than the method used:in MIDAS. "The , q j
- - 68.0 2042c l
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b lie & V ***b** Radiological Controls Department 6610-PLN-4200.02 Title Revition No. Ti.I Emergency Dose Calculation Manual (EDCM) 0 )
'l abundances of-the radioiodines were taken from the Radioactive Decay Data Tables, D.C. Kocher, 1981. All radionuclides are decayed during plume travel.
5.13.2 seel-Infinite Dose Model The TMI-1 RAC model calculates the thyroid dose rate due to inhalation of I-131, I-132, 1-133,'I-134, and I-135. -The thyroid dose rate is proportional to'X/Q. .The constant of proportionality is the product of the child breathing rate and' the child inhalation dose factors. .The program uses the child ;;
- breathing rate of 0.42 m8 /hr (f rom Table I-5, Reg. Guide 1.109) !
and the child inhalation dose factors are from Table E-9, Reg., j Guide 1.109 to compute the dose rate conversion f actors.s The , I dose,is computed by multiplying the dose rate by.the expected
' duration of release. .
u The radioiodines are decayed during plume travel' time. The decay constants for I-131 through I-135 are from the Radiological Henith Handbook. 5.13.2.1 X/Q Calculations
/ The basis for the X/Q calculation is the: Gaussian diffusion equation and a 10 x 7 array of sigma y's and sigma s's. The ;l array of values correspond to sigma y's and.sigmajz's for 7 3 I
stability, classes and at 10 fixed downwind distances. - For distances other than the fixed downwind distances, the sigma y's and sigma Z's are linearly interpolated before X/Q is computed .) for-that distance. The ten fixed distances ares 200,J500,1000, " 2000, 3000, 6000,'10000, 30000, 50000,'end'80000 meters.
~
5.13.2.2 Compute Building Effect Returns one of seven prs-computed virtual source distances, depending on stability. class.' The victual source distances for' 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! j nearest large building.- Building wake affects are simulated by 1 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 j without building wake effects at 800 meters downwind with : I stability class D.- ' X/Q' would then be computed at 800l 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 greatar.than the actual A downwind distance and is called only for groand level portion of release. 69.0 2042c-
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/ Radiological controls Department 6610-PLN-4200.02
\ Titla _Bevision No. 4 TMI Emergency Dose calculation Manual (EDCM) 0 ,
5.13.3 subroutines Used by Both rinite'and semi-infinite Models: 5.13.3.1 Compute THI-1 Emergency Action Level. . Declares the emerger.cy action level from highest dose whether whole body or thyroid.' . Emer. Action Level = Maximum Dose Rate (mrom/hr) within110'ailes e WHOLE BODY. THYROID-
=50 None 0 $ does rate -< 10 and 0; $ dose rate t <'
i Alert 10 $ dose' rate <' .50 or 50 . <, dose l rate < 250-i Site Area Emers.ency :50 5 dose rate <- 1000 or ,250' - <' dose, rate ' < 5000 General Emergency . dose rate-l3. 1000 dose rate - g ' 5000 The subroutine computes.the EAL for whol's body and thyroid dose-and then reports the more severe of the two.- , 5.13.3.2 Compute site Boundary The whole body an.1 thyroid doses are= computed at the site boundary. The distance to the site boundary varies with'the. compass sector that the wind is blowing to. Thistroutine returns this distance in meters.. 5.13.3.3 Compute Terrain Factor. j Computes terrain height in meters for a given downwind distance. L At downwind distances other 'than those in the subroutine, terrain height is computed by linear ihterpolation, except at distances closer than 610 meters. Between the plant'a.J 610 meters i downwind, the terrain height is set equal to t'ae terrain height. at 610 meters. Terrain further from the plant is never lower: i l than terrain closer to the plant due to mathematical l- approximations. 7 L 5.13.3.4 Compute Stebility class i As' measured by,the TMI Meteorological Tower from the 150'ft minus .t 33 ft temperature difference. Table 5.13-1 relates the temperature difference to the stability class. The equivalent f temperature dif ference per 100 ft .17 shown in the last colusm of . the table. With. regard to the temperature difference, 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 determined by the' measured temperature lapse rate.per Reg. Guide 1.21. V ' 70.0 2042c l-l-
ENuclear ml Radiological Controls Department
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6610-PLN-4200.0fi.
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Title Revision No. TMI Emergency Dose Calculation Manual (EDCM)- O
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5.13.3.5 Adjust Wind speed Adjusts wind speed from the anemometer' height to.the release I height. _The wind speed is adjusted according to the following equation: u.= u,(h/h o)P-where the subscript "0"-denotes the an - ter height andL"u" and "h" are the wind speed and height above ground,' respectively.- The exponent p is a function of stability: 0.25, 0.33 and 0.50 for unstable,. neutral and stable cases, respectively. . If the'-
. adjusted wind speed is less than 0.5 mph, the adjusted wind speed, t !
is. bet equal to 0.5 mph.- 5.13.3.6 compute Exit Velocity l l Computes exit velocity of the released material'in feet per second by dividing. cubic feet:per minute by the' stack ! cross-sectional area.. 5.13.3.7 Compute Plume Rise ; i computes the plume rise in meters-for the. elevated portion of.a spilt wake release. 1Nn) formulas are used-to calculate the s plume i rise; for unstable and neutral conditions jet plume rise,=. momentum dominated, is calculated from Briggs' Plume' Rise, Eq. -* 4.33; for stable stability,'it is calculated using Eq.E4.28 from Briggs' Plume Rise. ; i I 71.0 2042c , 1 1 a
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y g. 'l Radiological controls Department ~ 6610-PLN-4200.02- ~ Title ' Revision No. s
'TMI Emergency Dose celeulation Manual (EDCM)- 0 l
1 5.13.3.8 acompute Entrainment Factor' . l l Computes entrainment factor for wake split flows. .A mixed mode. ; release is assumed whens (1) the release point is at the'1evel-ofl l or above. adjacent' solid structures but lower than olevated- i release points,' (2)ithe- ratio of - plume exit velocity to hork.ontal' wind' speed is between one and five.: _specifically, the- - entrainmentL factor,EEg , is computed according to the following , formulas: ' E, = 1.0. f or w,/u le 1, E, = 2.58' - 1.58(wo /u) . foril gt w o/u le 1.5 L i Eg = 0.30 - 0.06(w,/u) for '1.5 gt ' w n/u le - 5.0 { Eg = 0.for w /u: o gt- 5.0 - where wo is the stack gas exat velocity and u is the wind speed-at stack height in miles per hour. ) Note that the entrainment factor does not address the case of J two
, adjacent plumes mixing with each other, as would be the case in TMI-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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Radiological Controls Department =- 6610-PLN-4200.02 Title Revision No.- ,
=
q o TM! Emergency' Dose calculation Manual (EDCM) O c i t. l TABLE 5.13-l'.
. classification of Atmospheric stability stability- Pasquill; _
Delta T Delta T
-Classification Categories ~- (150' - 33')- ' (
- F/100 ' ) --
(*F) (*F) 1 -{ ' Extremely Unstable- ..A. < -1.22 ,< -1.04 Moderately Unsta' ale B 3 -1.22 to yt -1.09 3:-1.04 to < -0.93 , l
'C, g -0.93 to < -0.82.
~
- Slightly Unstable. t_-1.09 to;< -0.96(
Neutral ;D_ 1 -0.96 to < -0.32 't'-0.82.to -0.27 q Slightly Stable 'E g_-0.32,.to < +0.96_ 0.27.to < +0.82 .) Moderately Stable .F 3:+0.96.to < +2.56 _1 +0.82 to < +2.19 Extremely Stable G > +2.56 > +2.19 j I 1 i
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<F g [ ' Number j Radioloolcal Controls Department 6610-PLN-4200.02_
Title -Revision No.,
'TMI Emergency Dose Calculation Manual (EDCM) =0-5.14E offsite Air Sample Analysis' i j
5.14.1 Introduction The "of fsite Air sample Ar.stysis"f portion of the RAC code is used [ in' conjunction with resulto per eided from field teams to assess :
)
thyroid dose; commitment. ; I The method involves ^ collection of an air. sample using a low flow j _(about 50' LPM) sampler with both a particulate filter and an- 1 1 iodine adsorber cartridge. The flow rate of the sampler, the '{ duration of sample collection, the background of the- frisker 'used 4 ito count the sample, the gross counts on the particulate' filter, ! and the gross counts on-the iodine cartridge are called-into the < RAC or EACC from the field teams. The RAC or EACC_ staff then' '-j uses the RAC code to estimate the offsite dose commitment-based on the sampls. 5.14.2 Assumptions A calibrated faco 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 i and ratemeters showed a consistent 0.0039 (0.39%)- counting _t efficiency. (Reference 7.7, 7.10, 7.11). Since I-131 has a l fairly strong beta (0.6 MeV max.), the usual particulate filter l
counting efficiency of 0.1 (10%) is;'903. =The collection: efficiency for both filters for theA, calculations are assumed to -l be 1.0. 5.14.3 Calculation j s The method first calculates the not' counts.per minute for the particulate and iodine cartridge. Then, using the given efficiencies separately, it calculates the air concentration of gaseous and particulate iodines. These are then combined for a total air concentration. A child breathing rate and: dose-conversion factor is then applied along with the estimated
~
duration of exposure to obtain the offsite dose, commitment. i Since the plant RAC code normally accounts for five different a iodine isotopes, the dose conversion factor (DCF) used is a j weighted average of the child DCFs based on the relative abundances of the five isotopes at damage classes of one and five-with 100 minutes decay. This accounts for counts on the samples which will be caused by isotopes other than I-131. 74.0 2042c l
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g-i l' I'WINuclear , , . Radiological Controls Department' 6610-PLN-4200.02 Title Revision No.= TMI Emergency Dose-Calculation Manual (EDCM) 0 5.14.4- . Example Given an offsite air sample was.taken with the~following results: Background = 100 eps gross cartridge countrate = 200 eps gross particulate countrate'= 200 cym-flow rate through~ sampler a-50 LPH
' sample duration = 10 min.
exposure duration =.1' hour-
-DRCF = 4.038 mees /hr pC1/cc.
The RAC program follows the logic below to calculate an cff-site thyroid dose commitment'for'this sample. a.. not particulate countrate = 200. 100I='100 cpm-b.- not particulate activity.= 100/.1 = 1000 dpm. ; i
- c. not cartridge count rate = 200 - 100_= 100 cpm-
- d. not cartridge activity = 100/0.0039'a 25600 dpm
- e. total activity in sample = 1000 + 25600 = 26600 dpm !
- f. total microcuries = 26600/2.22E6 = 0.012 pC1'
- g. sample volume = 50
- 1000 *-10 = SES cc'
- h. air concentration = 0'.012/5E6 =,2.4E-8 pC1/cc
- 1. dose commitment = 2'.4E-8pci/cc*1 hour *4E8 mrom M =L1.6 mrom'
. pol /cc 5.14.5 The. items listed below appear while performing this section of the RAC program. Once all inputLis entered, the resultant. ,
thyroid dose coannitment is displayed in mrem, or mrom/hr 'if .a 'j duration of one~(1)~ hour is entered. ; OFFSITE AIR SAMPLE ANALYSIS' EXAMPLE SAMPLE TIME (military clock) SAMPLE LOCATION FIELD TEAM DESIGNATION BACKGROUND COUNTRATE 100 Cpm GROSS PARTICULATE COUNTRATE 200 Cpm i 75.0 2042C_ j
I- .: } }. Nuclear ,,, M-- ! Radiological controls ' Departeent - 6610-PLW-4200.02' Title' Revision No. THI Emergency Dose' Calculation Manual (EDCM) 0-GROS 5 CHARCOAL COUNTRATE 200 Cpm FLONRATE Tlut0 UGH SAMPLER 60 LPM. DURATION OP 5 AMPLE COLLECTION 10 MIN-
^'
EXPECTED DURATION OF RELEASE 1 HOURS For Dose Rate Enter Duration of One Hour. THYROID DOSB COMMITMENT 9.668+00 MREM (using weighted child DRCF):
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Ed I & ( *** pg . l Radiological Controls Department 6610-PLN-4200.02-Title Revision No.' TMI Emergency Dose Calculation Man al (EDCM) O 5.15 Liquid Release Calculation - In this section of the program calculations are performed for liquid source term determination, (see Figures 5.15-1, 5.15-2, and; Tables 5.15-1, : 5.15-2 ) . NPC's.in'the river, travel time to , downstream users, and ingestion dose commitment calculations. The methods -j used to perform the calculations are as follows: )
- 1. The concentrations of the liquid ef fluents are determined by one of -
the following methods.' Bach method uses only the four usual' iodine
-isotopes and Co-134 and Co-137. All selections are menu driven.
a.- Normal Miscellaneous Liquids -'An ' average'Lisotopic content'for; '} miscellaneous plant-liquid wastes le called if this option is. j selected. The isotopic concentrations used are typical values- i for default use only... The values.cannot be changed. If other
~
=isotopics are known to be present, other options should bacused.
5 The. discharge rate'is used with the concentrations to calculate .; the source, term. l i
- b. Known Isotopic Concentrations. If the concentrations of the five i lodines and two cesiums are known from gamma analysis of the, liquid, then the option for use of actual isotopic concentrations can be used. The program will prompt the user for each isotope.
The discharge rate is prompted for in order to calculate the source term from the concentration.
- c. Primary to Secondary Leakage with Secondary Liquid Release. If primary to secondary leakage occurs and the secondary'liquide are +
released to the river, this optionuis appropriate'if the actual
~
concentrations are not known.- The program will prompt for the RCS temperature and pressure in order to calculate. the NRC damage class and associated isotopic percentages.1.The. isotope percentages for the five iodiner sre normalised to a sum of 1:and j-the cesium activit.es are calculated using ration for the co-134- 'i and Co-137 to the 1-131 based on the NRC damage class. 'The ratios are 0.0075 for Cs-137 and 0.035 for Co-134. These are i assumed to represent a. damage. class.7 or greater' accident. .To , adjust for other damage classes, 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, -; while at lower damage. classes, the' ratio may be adjusted down<by'
.s,-.1 or 0.
The leakrate and duration of primary to secondary-is used to s estimate the maximum activity'in the secondary system. The i 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 te secondary leakage. The isotopic concentrations and the di: a rate are used to calculate the source-term. ;
77.0 2042c
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i W o jg { Number' , I # Radie cical Controls Department' '6610-PLN-4200.02
\s_, Title + .
-Revision No.s g TMI Emergency Dose Calcalation Manual (EDCM) O' e.. ' The:actavity in the turbine building sump:following a primary.to'
; secondary-leak is calculated using the' isotopic percentages based -'
on damage class and the count rate of.RMLL10. The response of-RML 10 in cpa/pci/mliis.used.to estimate the total concentration > cand the isotopic percentages are used to calculate individual centrations. : 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 l
' flow rate and~ inputting the value.in the program.1. Instructions:are
-provided for obtal'ning the flow rate.1 :The river flow rate is then ;
- used'along with the discharge flow rate to calculate the concentration
~
in the= river.- The concentration in the riverLis'then divided by,the ~
, water MPC to' determine the MPCs in the river t'o downstream users.< :
The river concentration isLused'along with:the, total discharge; time,L j; 3.
?to calculate the- dose -commitment: to ,an ' individual from drinking the s q river water from one of the downstream intakes. . The river '
concentration is multiplied by the duration of;the1 release, the. usage factor, the ingestion dose commitment factor foriinfants,7 and'the' , inf ant usage f actor (3301/yr) to obtainL an- estimated. doce commitment .
- for 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 a11'seven isotopes.
( _,/
- 4. A fiume arrival time is estimated for each known downstream user.and= 1 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 TMI discharge. , y
- 5. If the concentration of any nuclide exceeds le-6 in the river, l downstream users must be informed and recommended"to curtail ~ usage., }
- 6. If'any MPC fraction exceeds 500 MPCs,,the NRC must'be notified within- [
, 24 hours per 10 CFR 20.403.. If the MPCs: exceed 5000, immediate .
'f L notification of the NRC is required.
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I 2 g Figure 5.15-2 6610-PLN-4200.02
- REVISION 0 Unit 2LiquidPathways Control Control + ;
Building = Service Sump Area Sump l Turbine , Diesel j Building :: Gen. 'A'
! Sump j- ItSump }
Tendon l i , Diesel l { Access , ' Gen.'B' l Gallery Sump l Jump
.\ !
+
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/RML) -
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@ -> COMPOSffE SAMPLER TO RIVER ,
80.0 2042c
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=~r Radiological controls Department 6610-PLN-4200.02 Title Revision No.
TMI Emereeney Dose calculation Manual (EDCM) 0 TABLE 5.15-1 TMI-2 Sump Capacity sugg Total capacity gallons / Inch Turbine Bldg. sump 1346 gals 22.43-Circulating Wate. Pump Hour' sump 572 gals 10.59. Control Bldg. Area sump 718 gals 9.96 Tendon Access Galley 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 qals 14.29 Chlorinator House sump ---- ----- Water Trectuant sump 1615 gals 22.43 J Air Intake Tunnel ,' Normal Sump 700 gals ----- I Emergency sump 100,000 gals 766.00 Condensate Polisher sump 2617 gals 62.31 03udge tolloc; eau =p 1106 gals 26.33 Heat $r Drain .ute ---- ----- solid Waete S.*;tng Facility Sump 1476 gals 24. n Aux. D1dg. sump 10,102 gals -202 Decay Beas v6 ult Swip 418.5 gals or 957 gals (total) - 10 Bul& dang Spray Vault samp 478.5 gals or 957 gals (total) - 10 h j y 01.6 2042C i
ENuclear m Radiological Controls Department 6610-PLN-4200.02 Title Revision No. TWI BmerQoney Dose Calculation Manual (EDCN) 0 TABLE B.15-2 THI-1 Sump / Tank Capacities Sugg Capacity (Gallons) Turbine Building sump (TBS) 10,000 Auxiliary Building sump (abs) 10,000 Aeactor Building' sump (RBS) 10,000 Intermediate' Building sump West 1,000 Tendon Access Gallery sump 1,000 Intarmodiate Building sump East 1,000 Auxiliary Boiler sump 2,000 Powdex Sump 40,00 ) Industrial Waste Treatment system sump (IWTS) 300,000 Industrial Waste .'iltration System Sump (IWFS) 80,000 Tanks THI-2 Condensate B Tank 250,000 THI-1 OTSO A or 8 (secondary) 25,000 TMI-1 WECST A or B 8,000 Neutraliser Tank 100,000 BW8T 350,000 i condensate Tank A/B 260,000 l C2.0 *2042c
[<41 Nuclear ,I
"-r Radiological controls Department I 6610-PLN-4200.02 Title Revision No.
TMI Emeroency Dose cateulation Manual itDcW) 0
$.16 Protective Action Recomendation Logic - The Logic Ciaoram is designed to enable the user to develop protective actions based upon plant conditions, release duration and dose assessments. The logic is diagramed in Figure 5.16-1 for THI-1 and Figure 5.16-2 for TMI-2.
l
- i 83.0 2042c' 1
{t" { j j g g Number Radiological controls Department 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 0 TIME IS.E 88.ApS IAST 1 Wet asaaman FIGURE 5.16-1 m y g gygg m g pgg larf Ama oemos, a a:ars I leem osmoose a maamm b arvav paa mas a m m a a naam rea was maanan in mas TptiliTsiit7 4 FREE E M peR vtfMD8 ,18 Ilpt#gs, y a Mclanal (MEEtERABkT8 585E54 CAN 007 K IthlE VITMDI -18 IGNm8 TMDI AC M N TDi DI AGERBA8CE WITN IK BBB48EE. A8 A leStat 1581B8010 DELTUOMB IRA fD 8 BEES F W ORD fue N WAmalffD nagsg Ase 3 00Lg8 30VgMan h A8 & 19 9884 MIDGEBS 3 DELTUBIB FW 8188 R4518 7 4p AidB 8 ;&E8 EMlWDe AstD CGITDAR 480539Cff O l C
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I i JR & V TMI I Radiological Controls Department 6610-PLN.4200.02 ' ( Title Revision No. 7981 Emergency Dose Calculation Manual (EDCM) O j r L FIGURE 5.16-l' (Cont'd) l PROTECTIVE ACTION RECOMMENDATIONS (PAR) LOGIC DIAGRAM NOTE la AS INDICATED BY ONE OF THE FOLLONING 1
- 1. CAT-PAS SAMPLE RESULTS OF TOTAL NOBLE GAS CONCENTRATION > 2300 MICRO CI/CC l I
TOTAL RADICIODINE CONCENTRATION > 420 MICRO CI/CC
- 2. HIGH RANGE CONTAINMENT AREA MONITOR READINGS OF RNG-22 OR RMG-23 > 400 R/HR (HIGH ALAAM) !
NOTE 2: AS INDICATED BY ONE OF THE FOLLOWING
- 1. RCS POST ACCIDENT SAMPJ.E ANALYSIS INDICATES GREATER THAN NRC DAMAGE CODE 2.
- 2. RAC SOFTWARE COD 3 CALCULATES GREATER THAN N8tC DAMAGE CODE 2. (THIS IS BASED ON RCS PRESSURE AND THE AVERAGE OF THE FIVE HIGHEST INCORE THERMOCOUPLES - POINT C40D6 ON COMPUTER) .
l
- 3. LETDOWN MONITOR READINGS RML 1' LOW AND RML-1 HIGH ARE s OFFSCAt.E HIGH ('tHE ISOLATION INTERLOCKS MUST BE BYPASSED TO GET THIS READING).
t NOTE 3: AS INDICATED BY ONE OF THE FOLLOWING:
.1 - REACTOR AUILDING PRESSURE > 30 PSIG t 2 - REACTOR BUILDING HYDROGEN CONCENTRATION > 3% BY VOLUME 3 - SIGNIFICANT OTSG LEAKAGE INVOLVING MULTIPLE TUEE FAILURES.
4 - A DIRECT REACTOh BUILDING TO ATMOSPHERE RELEASE PATH-WAY SUCH A8 RB PLitGE VALVES FAILURE TO CLOSE.- f f i I V 85.0 2042c-9 _w- -.-- .,--w .-w,
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I
' A N TMI I Radiologleal controls Department 6610-PLN-4200.02 Title Revision No. 1
. i
.TMI Emergency Doeo Calculation Manual (EDCM) 0 1
FIGURE 5.16-1 (Cont'd) 1 PROTECTIVE ACTION RECO90tENDATIONS (PAR) LOGIC DIAGRAM 1 NOTE 4: TMI EVACUATION TIME ESTIMATES i i LOWER (HOURS) UPPER (HOURS) DEST ESTIMATE (NIGHT) 5.75 9.50 I TYPICAL WEEKDAY (NORMAL) 6.25- 10.25 ADVERSE WEATHER 10.00 12.25 j i LOWER - 000D STATE OF EMERGENCY READINESS (SLOW SCENARIO) , UPPER - LACK OF ADEQUATE PREPARATION TIMI (FAST SCENARIO) l t 1
.........................................................a...................
l NOTE 5: CONSIDERATION SHOULD BE GIVEN TO THE PROJECTED EXPOSURE TO
- SE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATESj IN ;
SO DOING, YOU NUST FACTOR RELEASE DURATION, RELEASE. MAGNITUDE AND ASSUME A PROTECTION FACTOR OF 2 FOR UP TO THE
> F7RST 2 HOURS OF RELEASE ~ DURATION AND A PF OF 1 FOR > l i- \ 2 HOURS RELEASE DURATION.- THE PATHWAY OF LEAST EXPOSURE
- SHOULD BE CHOSEN. . IF THE DOSE RATE IS 400 NREM/HR; SHELTERING FOR 3 HOURS WOULD RESULT IN AN EXPOSURE OF ,
800 MREM. , ! NOTE 6: PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICA-f TION OF THE KEYHOLE CONCEPT. ' (CONSIDER USING 2 MILE RADIUS ANT 10 MILE DOWNNIND). s s I f- , I, 86.0 2042c
b* @ M7 THI Radiological controls Department 6610-PLN-4200.02 Title Revision No. TWI Smereeney Dose Calculation Manual (EDCM) 0 FIGURE 6.16-2 1855882 M M 8 ana mesur y IECBeceafEBS - 888D Isu manocups,asawso I (- oneserer a suuss i dr svav na Wuns umas a e 1esame om uuss amm to suas WES E &M M pas tgnes as egang& F A artieses N MNRS M telIER R 8581499 6 ISBRB 1 DEN 3,33 . . .c n ' , ,,,,,,,,,, g,,,, g ggg,g ,, g ammaan ammasma gg,1ging gan 13 3 gags FSOREB848B M mens me 3 saEB EhAIWBS as a egens summens S&W W 8188 SSB 7
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J e N I TMI Radiological Controls Department 6610-PLN-4200.02 , Title Revision No. l l TMI Emereeney Dose Calculation Manual (EDCM) 0 I I i FIGURE 5.16-2 (Cont'd) ) THREE MILE ISLAND UNIT 2 PAR LOGIC DIAGRAM NOTES NOTE 1: AS INDICATED BY ONE OF THE FOLLOWING: 1 - REACTOR BUILDING PRESSURE > 4 PSIG 2 - REACTOR BUILDING NYDROGEN CONCENTRATION > 3% BY VOLUME 3 - A DIRECT REAC1CR SUILDING TO ATHOSPHERE RELEASE PATH ' t WAY SUCH AS R5 PUNGE VALVES FAILURE:TO CLOSE. t 4 - A DIRECT FUEL HANDLING JUILDING TO ATNOSPHERE RELEASE " PATHWAY SUCH AS A FILTER TRAIN FAILURE. ( NOTE 2: TMI EVACUATION TIME ESTIMATES LOWER (HOURS) UPPER (HOURS) l BEST ESTIMATE (NIGHT) 5.75 9.50 , TYPICAL WEEKDAY (NORMAL) 6.25 10.25 ADVERSE WEATHER 10.00 12.25 LOWER - OJOD STATE OF EMERGENCY READINESS-(SLOW 5'WNARIO) UPPER - LACK OF ADEQUATE PREPARATION TIME (FAST SCENARIO) [ , i l t NOTE 3: CONSIDERATION SHOULD BE GIVEN TO THE PROJECTED EXPOSURE TO ; SE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATES, IN SO ; I DOING, YOU MUST FACTOR RELEASE-DURATION, RELEASE MAGNITUL.4 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 PATHhhY 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. l l l l l NOTE 4: PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICATION l OF THE KEYHOLE CONCEPT I (CONSIDER USING 2 MILE RADIUS AND 10 MILES DOWNWIND) )
..__ ___ ...._____........___........______..________________...____... __... J l
-.1 1
1 1 1 1 88.0 -2042c - _ _ . - _ . _ c .. . ..-_;__.. ,
a 4 lan E & ( *** Radio 14eical Controla Department 6610-PLN-4200.02-Title Revision No. TMI Roeroency Dose Calculation Manual (EDCM)- 0 5.17 Dose Projection Model Overview THI-1 5.17.1 The dose projection model may be regarded as three distinct sections. Certain variables are passed between those sectiones
- a. 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 the operator, upon learning that the met conditions have changed and l 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'est of source term and met parameters.
]
5.17.2 source term parameters that are used in the dose projection portion of the model are:
- 1. An array of the 15 isotopes of interest in pci/sec. I
- 2. The release flow rate in efm.
- 3. The release point 6 tack height of 48.6 meters. . ]
- 4. The release point stack diametet in meters - 1.7 meters ~for the Station Vent and 1.1 meters for the Reactor Building. i
- 5. A single letter code for the type of release, I
- a. G - Ground level
- b. 8 - Split wake
- c. E - Elevated l
- 6. A character string describing the source, i.e., RNA-6 300 cpm.
- 7. The date and time of the source term parLmeter entry.
The met conditions section requires as input the type of release, i.e., G, 5, or I. If this variable is not present the operator is required to specify it before the program continues. 89.0 2042c- !
. . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _.~.__ _ _ _
l N ggg Radiological Centrole Department 6610-PLN+4200.02 ! Title - Revision No.
]
TMI Emergency Dose calculation Manual (EDCM) 0 i i 5.16 THI-2 Source Term Calculation ! NOTE: . Follow Figurs 5.18-1 TMI-2 RAC Program Flowchart following this-section. ,
......................._.................. ......... __........ .___....... i 5.18.1 Source Term Calculations - The source term portion of'the TMI-2 dose assessment program is used to generate the quantity and :
radionuclide make up of the radioactive material released (or
' available for release) to the environment. Once the source term is measured or estimated, meteorological and dosimetry _models are ,
applied to the assessment. Some specific accident scenarios are ased to calculate radionuclide release factors and assess the accident consequences. These assessments are documented in ; Technical Evaluation sports (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 J Radiation Monitoring System readings, 2) Actual sample results, i J or 3) Contingency calculations. l
-? 5.18.2 The following f acilities are considered as oeing radioactive . ^
material release pathways for the TMI-2 RAC program. , i i 4 l DEFAULT ; VENTILATION CURIES FACILITY RMS FLOW RATE (CFM) AVAILABLE I J 1 - WHPF EBERLINE PING 7100 100 I 2 - RLM EBERLINE PING 900 100 3 - CACE AMS-3 (2) 2000 100 4 - EPICOR II EDERLINE PING 9000 100 ,
.1 5 - B&W TRAILER AMS-3 1300 100 l
6 - STATION VENT HPR219 VICTOREEN PING 120K - 130K 100 j 7 - STATION VENT HPR219A IBERLINE PING 120K - 130K 100 I E - ISWSF/ PAINT SHED NONE NONE 100 ; i 9 - RAD. INST. SHOP NONE 4000 100 4 The above list shows the associated default data concerning each ; release point in the THI-2 RAC program. The station vent release f 90.0 2042c
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, Nuclear TMI
== ~ r !
Radiological Controls Department 6610-PLN-4200.02-_
~
Titit Revision No. T'd! Emergency Dose Calculation Manual (EDCM) 0 pathway also includes an optinn 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 urer. The following parameters are used to calculate a THI-2 source ters:
- 1. RMs reading (CPM, pC1)
- 2. RMS Channel Efficiency (CPM /pC1/cc, CPM / min /pci/cc).
n
- 3. The release flow rate (CFM) j i
- 4. Co-137/8r-90 Ratio.
5.18.4 Post Accident sample source Term Calculation This option for a particular release pathway is to use actual sample results to develope a source term. This option would be l the most preferable method for calculating - THI-2 source term - since this method eliminates built in conse vatisms.from RMs or
\s_- contingency calculations. In this option the sample results or defaults, if required, will be used in conjunction with the release flow rete (CFM) to calculate a source term.
5.18.5 contingency calculation Source Term calculation This option utilizes 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 f acility along with the associated release - flow rate are used to . calculate a conservative source term. 5.18.5.1 source Term Filtration (Contingency Cale only) During the final calculation of a source term a filtration i efficiency will be prompted to determine the fraction of ths I source term that will not reach the' environment due-to I flitration. I 1 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-1 dose projection process. 'A meteorology I. option will gather meteorology data and combine it wita the f, source term information to complete a dose projectior. .
91.0 2042c t l
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l; Radiological controls Department 6610-PLN-4200.02 ks Title hevision No. ; ! 'TNI Emergency Dose Calculation Manual (EDcM) 0 ; l l i ! ! '5.18.7 Dose Projection Model Overview THI-2 l 5.18.7.1 The dose projection model may be regarded as three distinct l sections. Certain variables are passed between those sections . ,
- s. source term generation 3
- 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 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 i 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 are t g
- 1. The radionuclides ar-90 and co-137 in their. chosen ratio.
- 2. The release flow tate in efm.
- 3. The choice of fuel related source term or not.
- 4. A single letter code for the type of release.
- a. G - Ground level
- b. S - Split wake
- c. E - Elevated
- 5. The date and time of the source term paraester entry.
The met conditions section requires as input the type of rslease, t i.e., G, Se or E. If this variable is not present the opetator is required to specify it before the program continees. . P z 92.0 2042c t t
-- . . .- . . -- - .- . ~ _ _ ._ . - - - -- - . . - . . - . . - - .
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( Radiological controls Department 6610-PLN-4200.02 ; Title Revision No. l TMI Emergency Dose calculation Manual (EDCM) 0 i 5.18.7.3 Met parameters that are passed to the dose projection portion of the model are:
- 1. Delta T in or per 117 feet.
- 2. Wind speed in mph.
1
- 3. Wind direction in degrees, from. I
' 4. The date and time that met conditions were entered.
Upon being invoked, the dose calculation segment checks to ese if the 15 isotopic values are defined. If they are not, the source term segment is invoked. After ..his the dose calculation.eegment 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. 5.18.7.4 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 dase projection segment is called. ! The dose projection segment does not retain any variables for the i (
\ orogram's use elsewhere in the program but sends all its output
.o the screen. l t
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Number 5 TN1 Radioloalcal controls Department 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDOf) 0 FIGURE 5.18-1 THREE MILE ISLAND UNIT 2 RAC PROGRAM FLOWCHART 1 Wa.pr ' 2 .- R' mo START 3 - cact 4 - IPiCOR !! 5 - B&w ?RAr.to 6 $? A?:0N vtN? w#*2'$ V 7 - ST At:0N VEN* -**: H gg,gg?
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Title Revision No.
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Tit! Emer0ency Dose Calculation Manual (EDCN) 0 6.0 RESPON82BILITIES 6.1 The Rhc is respor.sible to ensure that dose assessments using the l methodology in the EDCM are performed upen implementation of the Emergency > Plan. . 6.2 The RASE have the responsibility to support the RAC in performance of. radiological controle and dose essessment using the methodology-in the ' EDCN. 6.3 The Chemistry Coordinator has the responsibility to supF rt, t% RAC in the procurement and analysis of in-plant samples required to W ar.cffy the accident. I 6.4 Radiological Contr0ls has the responsibility of proper review,
. documentation, and Listribution of the EDCM and the RAC program software.
Radiological contro;s is responsible for ensuring that the EDCM and the RAC program software are current and compatible. l 1
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7.0 REFERENCES
7.1 American National Standard (ANS), ANSI /ANS-18.1 - 1984, Radioactive source l Term for Normal Operations of Light Water Reactors l 7.2 APS Source Term Report - Report to the American Physical Society of the Study Group on Radionuclide Releases From severe Accidents at Nuclear Power Plants, February 1985 j l 7.3 Dose Assessment Manual'for Emergency Preparedness Coordinators, February J 1986, INPO 86-008 7.4 EDCM Flowchart Block Diagram 7.5 Efficiency Check using an Air 1-131 Source Cartridge and a Ba-133 Source Cartridge, Memorandum 9502.-88-0139, September 28, 1988 j 7.6 Emergency Dose Calculation Manual.(EDCM) Source Code Listing j l 7.7 IPA 520/1-75-001 - Manual of Protective Action Guides and Protective i Actions for Nuclear Incidents 1 7.8 EPIP 6415-1MP-1300.07 - Off-site /On-site Dose Projections- i 5 ) 7.9 Evaluation of a Front Loaded Iodine Cartridge using Various Survey Equipment, Memorandum 9100-88-0194, May 12, 1988 f 7.10 Field Measurements of Airborne Releases of Radioactive Material, Memorandum l 9502-88-0098, May 25, 1986 7.11 FSAR, THI-1 Chapter 11, Radioactive Waste and Radiation Protection 7.12 FSAR, TMI-1 Chapter 14 - Safety Analysin i i 7.13 FSAR, THI-2, Volume 10, Section 15, Accident Analysis l I 7.14 GPUNC Emergency Plan, 2000-PLN-1300.01 7.15 GPUNC Emergency Preparedness Program, 1000-ADM-1319.01 , 7.16 ICRP Report of the Task Group on Reference Man, 1981 7.17 INPO 80ER 83-2 " Steam Generator Tube Ruptures" 7.18 Introduction to Health Physics, Herman Comber, 2nd Edition, 1985 7.19 NRC-BKL Source Term Report 7 7.20 NUREG-0017 Rev.1 - PWR - GALE Code; Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWR, April 1976 101.0 2042c
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=~r Radiological controls Department 6610-PLN-4200.02 , k Title .
Revision No. TMI Emereeney Dose Calculation Manual (EDCM) 0 7.21 NUREG-0133 - Preparation of Radiological Effluent Technical Specifications , for Nuclear Power Plants, October 1978. 7.22 NUREG-0591 - Environmental Assessment for use of EPICOR II at Three Mile Island Unit 2, October 3, 1979' 7.23 NUREG-0654 - Revision 1 - Criteria for Preparation and Evaluation of i Radiological Emergency Response Plar.s and Preparedness in Support of l
. Nuclear Power Plants t 7.24 NUREG-0737 - Clarification of TMI Action Plant Requirements, U.S. Nuclear
' Regulatory Commission, November'1980, Generic Letter 82-33, supplement 1. to ~
NUREG-0737 - Requirements for Emergency Response Capability, U.S. Nuclear Regulatory Commission, Washington, D.C., December 1982 { 7.25 NUREG-1228 - Source Term Estimation during Incident Response to Severs ! Nuclear Power Plant Accidents, October 1988 7.26 NUREG/CR-3011 - Dose Projection Considerations for Emergency Conditions at i Nuclear Power Plants
. 7.27 N1830 - Post Accident Reactor Coolant Sampling 7.23 N1831 - Post Accident Atmospheric Sampling 7.29 N1832 - Post Accident Sample Analysis 7.30 N1833 - Post Accident Core Damage Calculations l 7.31 OP1202 Abnormal Transients Rules, Guides and Graphs 7.32 OP1202 RCS Super Heated -!
7.33 OP1210 Excessive Radiation Levels 7.34 operational Quality Assurance Plan, 1000-PLN-7200.01 7.35 Proprietary Midas User Documentation, Pickard, 1.,we, and Garrick 7.36 Radioactive Decay Data Tables, David C. Kocher, ORNL, DOE / TIC-11026, 1981. 7.37 Radiological Health Handbook, Revised Edition Jan. 1970, US Dept. HEW.- 7.38 Reg. Guide 1.21 Measuring, Evaluating, and Reporting Radioactivity'in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants, Rev.-1, June 1974. . 7.39 Reg. Guide 1.109 - Calculation of Annual. Doses to Man From Routine Releases i of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix 1, October 1977, Rev. 1 l l l 102.0 2042c i I
7 Nuclear ,, Radiological Controls Department 6610-PLN-4200.02 l Title Revision No. l TM5EmergencyDoescalculationManual(EDCM) 0 7.40 SER-15737-2-007-108, Rev. 4, March 5, 1985. Safety Evaluation Report for a l 1981-2 Fuel Canister Accident . l 7.41 SER-419628-003, Rev. 7, Instrument Calibration Facility, Feb. 12,.1988 7.42 TDR-390 - TNI-1; Primary-to-Secondary OTSG Leakage and its On-site /Off-site Radiological Impact, April 1983 7.43 TDR-405.- TMI-1; Evaluation of Plant Radiation Release and its 10CFR50, l Appendix I conformance for Different Operating Conditione ; 0 ' 7.44 TDR-431 - Method for Estimating Extent f.f Core Damage Under Severe Accident conditione
]
7.45 TER-13587-02-G03 015, Rev. 6, January 21, 1986. Interim Solid Waste Staging Facility Technical Evaluation Report 7.46 TER-15737-2-G00-104, Rev. 5, May 8, 1985. Technical Evaluation Report for the CACE 7.47 TER-15737-2-003-107, Rev. 6, Feb. 2, 1987. Technical Evaluation Report for I the WHPF. 7.48 WASH-1400 - 1975 Nuclear Safety Study WASH-1400 (also known se Rasmussen Report) 8.0 EXHIBITS 8.1 Exhibit 1, RAC Data Collection (Example) l 4 i f r b a l . O
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'TWI Emergency Dose Calculation Manual (EDCN) 0 _
ERHIBIT 1 -] 1 RAC Data Collection (Example) l 1 rege 1 ; RAC DATA COLLLCTION ! Primeer to 1stenaery nelseng- ] TIMt 0F READING l l l$Y$ TEM l PARANTtt l INDICATION l i 1 1 1 1 I l I 1 I
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1MI Basroency Dose calculation Manual (EDCM) 0 EKHIBIT 1 (Cont'd)- ; i l Page 2 RAC DATA COLLtCTION ;
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RAC DATA COLLECT!DN Aust14ary and Fuel Han4ites hutidinu nelegt (Coet's.) PARAMETit l INDICATION I t!Mt 0F READING l -~ 840NITOR l L l ! I i 1 1 I I l 1 l W G1 I teeteel noas I d/he i I I I I I I , l W-G2 i Radiothesister tab I et/he 1 I I I I I l l W G3 i Nucleae tenoitna Room i mR/h* 1 1 I I I i l l W G4 I slet Machine lhes I d/he ! I I I I I l ' l W G9 1 FHS Seemt Fuel lettaa f '4/he i I I I I I 1 l W G10 i Aux. Bida. 30$' I e4/Fe i 1 1 l t l l W G11 i MU/P Domin Area I mR/he l 1 1 1 1 l l l W G12 i Sol 9d had. Waste Area I d/me I l 1 1 i I j. k I mA/h, I I I i l l W G13 i Aux. Blea. 201' 1 1 l ' l W G14 I Waste Evasorater Area I mA/he - I I I I I I l
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