ML20248D407
| ML20248D407 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/27/1989 |
| From: | Shaw R, William S GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| 9100-PLN-4200.0, NUDOCS 8910040238 | |
| Download: ML20248D407 (105) | |
Text
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l- THI-l EMERGENCY DOSE CALCULATION MANUAL (EDCM) l INSTRUCTION MEMO g, 8 c b RETURN'TO: Betty Nash Procedure Distribution Control Unit 2 Admin. Bldg. n, TMI h'f y( -
-Please update your Copy of the THI-l Emergency Dose Calculation Manual (EDCM) as E
}- instructed below. Also, please sign the acknowledgement at the bottom of this memo and return to Betty Nash as shown above.
REMOVE INSERT 9/on-?L M- N'A 00. OA 9/00 -?L N- Ala 00. d2
- h. O c'o . DA L.f.(h D & <
o k-A s-E9." _ d 9-h NY AR %. /-Rd-f9-dA#Y ADDITIONAL INSTRUCTIONS / COMMENTS I hereby acknowledge receipt of the item above and have complied with the instructions.
(Signature) (Ext. No.) (Date) 2175B/0033b DC 891004o238 890927 ros noocx osoog28_
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La Nuclear TMI-1 Number
/ ' 'N, Radiological Controls Department 9100-PLN-4200.02 d ) Title Revision No.
v TMI Emergency Dose Calc'ulation Manual (EDCM) 8 Applicability / Scope The EDCM is applicable to all qualified Responsible Office Radiological Assessment Coordinators. This manual provides the methods used to perform dose projections durino emergencies 9100 This document is within QA plan scope X Yes No Effective Date Safety Reviews Required _ X_ Yes __
No List of Effective Pages Page Revision Page Revision Page Revision Page Revision 1.0 3 27.0 8 53.0 8 79.0 8 2.0 8 28.0 8 54.0 8 80.0 8 3.0 8 29.0 8 55.0 8 81.0 4.0 7 30.0 8 56.0 8 8 5.0 7 31.0 8 57.0 8 8 6.0 7 32.0 8 58.0 8 .0 8 7.0 7 33.0 8 9.0 8 8 8.0 8 34.0 8 Q 8 .0 8 9.0 7 35.0 .0 7.0 8 10.0 7 36.0 2.0 88.0 8 11.0 8 37.0 63.0 xV 8 89.0 8 12.0 7 38.0 64. 90.0 8 13.0 7 39. ci 8 6 91.0 8
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71.0 97.0 8 20.0 7 46.0 72.0 98.0 8 21.0 8 47 7 99.0 8 22.0 7 4 - 8 0 8 100.0 8 23.0 7 9. 8 8 101.0 8 24.0 8 0 8 6 8 102.0 8 25.0 8 . 8 103.0 8 26.0 2.0 8 78.0 8 104.0 8 11// Signature l Concurring Organizational Elcment I Date Oriainatoril 6 f u/d:'*5*" i Radiological Enoineer I h 2 T- TY Concurred ll6 C d d f u r d a l Environmental Scientist ld 7 4 4 By ll 7 fle vb l Emergency Preparedness l 7 dG-89 ll b Mr. e l Radiological Engineering Mgr.THI-11 9.-24,-21 ll //[bd Environ. Pams. Mgr.. Envir. Cont. L2 }b F1 ll h l Emergency Prep. Manager. TMI l 9[26[f9 Il M M /dfM M l Radiological Enaineerino Mar.TMI-21 '9/zh'/99 Approved llEP Mb l Rad Con Director, TMI-1 l 9--27~ 8/
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TMI Emergency Dose Calculation Manual (EDCM) 8 Table of Contents
.Section Page 1.0 PURPOSE 4.0 2.0 APPLICABILITY / SCOPE 5.0 3.0 DEFINITIONS -6.0 4.0 PREREQUISITES 16.0 5.0 PROCEDURE 17.0 5.1 Source Term Calculations 19.0 5.2 Selection of Release Pathways and Characteristics 20.0 5.3 Calculation of NRC Damage Class and Isotopic Percentages 24.0 5.4 Radiation Monitoring System (RMS) Source Term Calculation 35.0
! 5.5 Post Accident Samples Source Term Calculation 39.0
'5.6 Contingency Calculations Source Term Generation 41.0 5.7 Decay Scheme c a lculation 46.0 5.8 Noble Gas to Iodine Ratio Calculations 48.0 5.9 Effluent Release Flow Rates 51.0 5.10 Two-Phase Steam 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 5.15 Liquid Release Calculation /7.0 5.16 Protective Action Recommendation Logit 83.0 5.17 Dose Projection Model Overview THI-l 89.0 5.18 TMI-2 Source Term Calculation 90.0 L
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M2l$87 TMI-1 Number c'~'] Radiological Controls Department 9100-PLN 4200.02
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TMI Emergency Dose Calculation Manual (EDCM) 8 Table of Contents (Cont'd)
Section Page 6.0 RESPONSIBILITIES 100.0
7.0 REFERENCES
101.0 8.0 EXHIBITS 104.0 None
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THI Emergency Dose Calculation Manual (EDCM) 7 l'.0PURDOSE'
-The purpose of this manual is to provide'a document that describes the assumptions and methodology used in the current TMI-l and TMI-2 Radiological Assessment Coordinator (RAC) programs. This includes calculating projected on-site and off-site doses.from releases of radioactive material-to the environment 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 off-site sample results, or' contingency ca;ralations, if RMS and sample results are not available. These-dose projections are performed by computer using the current version of the TMI-l or THI-2 RAC. programs. The Radiological Assessment Coordinator is responsible for implementing the dose projection process for THI-l and THI-2.
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TMI Emergency Dose Calculation Manual (EDCM) 7 2.0 APPLICABILITY / SCOPE The EDCM is applicable to all qualified TMI Radiological Assessment Coordinator l (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 dose projections during emergency situations where radioactive material has been or is predicted to be released to the environment.
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TMI Emergency Dose Calculation Manual (EDCM) 7 3.0 DEFINITIONS 3.1 BABCOCK AND WILCOX TRAILER (B&W TRAILER) - that was used for TMI-2 sample analyses after the THI-2 accident. This facility has:its own ventilation system including installed HEPA filters. This facility has an AMS-3 installed as a ventilation radiation monitor. The ventilation system is run at 1300 CFM.
3.2 BUILDING WAKE EFFECTS - When an atmospheric release occurs at, near, or below the top of a building (or any structure) the dispersion of the release is affected by the wake effect of the building. Air flow over and around the structure from the prevailing wind tends to drive the release 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 concentrations are affected very little by building wake. Building wake effects are most noticeable for ground level or low flow stack releases 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.
b 5 3.3 " CHI over Q" (X/Q) - is the dispersion of a gaseous release in the environme.it calculated by the split wake dispersion model. Normal units of X/Q are sec/ cubic meter. X/Q is used to determine environmental atmospheric concentrations by multiplying the source term represented by Q. Thus dispersion, X/Q (sec/ cubic meter) times source term, Q (pCi/sec) yields environmental concentration X (pCi/ cubic meter).. X/Q.
l 1s a function of many parameters including wind speed, delta T (change in temperature with height), release point height, building size, and release velocity, among others. The release model takes all these into account when calculating atmospheric dispersion.
3.4 CONTAINMENT AIR CONTROL ENVELOPE (CACE) - This facility provides a l containment outside of the TMI-2 equipment hatch. This containment uses
- l. two AMS-3 radiation monitors to monitor releases, at 2000 CFM, from this
! facility when the facility is in use.
3.5 CONTAINMENT ATMOSPHERIC POST-ACCIDENT SAMPLING SYSTEM (CATPASS) - Post accident sampling system capable of providing sample (s) following an accident condition, coincident with a blackout, with limited personnel-exposure. The sampilng 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 L 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 damage and provide good data for the Reactor Building source term if a keactor Building release is possible.
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7 3.6 CONTINGENCY CALCULATION - A source term calculation performed in the absence of sufficient effluent 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.
3.7 CORE DAMAGE - (Note: This definition to be used in lieu of defective /fal.ed fuel.) A set of core classifications used to address the requirements of the NRC NUREG 0737 Criterion 2(a) upon implementation of the Emergency Plan. Based upon RCS pressure and incore thermocouple readings, an assessment is made of the degree of cladding failure, fuel overheat, and fuel melt.
3.8 DEFECTIVE FUEL / FAILED FUEL - See definition of core damage.
3.9 DOSE RATE CONVERSION FACTOR (DRCF) - A parameter calculated by the methods and models of internal dosimetry, which indicates the committed dose equivalent (to the whole body or an organ) per unit activity inhaled or ingested. This parameter is specific to the radionuclides and the dose pathway. Dose conversion factors are commonly tabulated in units of mrem /hr per curie /m3 inhaled or ingested.
3.10 ELEVATED PELEASE - An airborne effluent plume which is well above any (Q building wake effects so as to be essentially unentrained is termed an f.e,ated release. ,he source of the plume may be elevated either by virtue of the physical height of the source above the ground elevation and buildings or by a combination of the physical height and the jet plume rise. Semi infinite modeling of elevated releases generally will not produce any significant ground level concentrations within the first few hundred yards of the source. Semi infinite modeling of elevated releases generally have less dose consequence to the public due to the greater downwind distance to the ground concentration maximum compared to ground releases. Elevated releases as used in the EDCM actually means "not at ground" in the split wake plume model. Other definitions of " elevated" 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.
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.
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TMI Emergency Dose Calculation Mar.ual (EDCM) 8 3.13 EMERGENCY DOSE CALCULATION MANUAL (EDCM) - This controlled dose calculation manual is the documentation describing the content and calculational methods of the Radiological Assessment Coordinator (RAC) program.
3.14 EMERGENCY OPERATIONS FACILITY (EOF) - The Emergency Operations facilities serve as the primary locations for management of the Corporation's overall emergency response. These fr Gities are equipped for and staffed by the Emergency Support Organizaf.on to coordinate emergency response with
, off-site support agencies and to assess the environmental impact of the emergency. The EOF participates in accident assessment and transmits appropriate data and recommended protective actions to Federal, State and Local agencies.
3.15 EMERGENCY PLANNING ZONE (EPZ) - There are two Emergency Planning Zones.
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 gi'an 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 call 4 the Plume Exposure Pathway EPZ. The second is an area approximately 50 miles in radias around the site, for which emergency jm planning consideration of the ingestion exposure pathway has been given.
This is called the Ingestion Exposure Pathway EPZ.
(v} 3.16 ENTRAINMENT - When a release is treated as a wake split release an entrainment factor is applied to specify how much of the release is to be considered elevated and how much is to be considered a ground release.
Entrainment factor is related to the building wake effect. The entrainment 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 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 :enditions cause entrainment factors which will split the release between ground and elevated. The general form for the application of the entrainment factor (Ef) is:
X/Q(splitwake).X/Q(ground)*Ef + X/Q(elevated)*(1-Ef).
As can be seen from the formula, when the entrainment factor is one, the release is entiraly ground and when the entrainment factor is zero, the release is entirely elevated. When 0 < Ef < 1 then the release is split.
3.17 EPICOR II - Radioactive Liquid Waste Processing Facility 1ocated on the east side of TMI-2. This facility is used to procest, THI-2 radioactive waste. An Eberline PING radiation monitor is located on the ventilation I
system of this facility. The ventilation system average flow rate is 9000 CFM.
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7 3.18 EXCLUSION AREA (EA) - As defined in 10CFR100.3; "that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area". At TMI this is an area with a 2000 ft. radius from the point equidistant between the centers of the iMI-l and TMI-2 reactor buildings.
3.19 EXIT VELOCITY AND PLUME RISE - Atmospheric dispersion and ground concentrations are in part dependent on release height. Higher release heights will cause lower maximum concentrations at 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 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 P'.UME MODEL - Atmospheric dispersion and dose assessment model which is based on the assumption that the horizontal 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, and the air attenuation of the p}
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ks gamma rays coming from the cloud. This model can estimate the dose to a 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 the effluent source.
3.21 FUEL HANDLING BUILDING ENGINEERED SAFETY FEATURE VENTILATION SYSTEM - The Fuel Handling Building ESF Ventilation System, is being added to TMI-I in accordance with a commitment to the NRC. This commitment has been included in the NRC THI-l restart report. The Fuel Handling Building ESF Ventilation System is installed to contain, confine, control, mitigate, monitor and record radiation release resulting from a THI-l pcstulated spent fuel accident in the Fuel Handling Building as described in FSAR, Section 14.2.2.1, Update 1, 7/82. Normal operation of the Fuel Handling Building ESF Ventilation System will be during THI-l spent fuel movements in the Fuel Handling Building. The system design shall include adeq, ate air filtration and exhaust capacity to ensure that no uncontrolled radioactive release to atmosphere occurs. The System shall include effluent radiation monitoring capability.
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- l. 3.22 GAUSSIAN PLUME EQUATION - An equation which takes input parameters of l plume height, sigma-Y, sigma-Z, and wind speed, which explicitly I calculates the straight line Gaussian Plume Dispersion. The Gaussian Plume equation actually averages short term variations to produce a mean effective plume, so short term measurements of the plume may not be l duplicated by the Gaussian Plume Model.
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(y. Title TMI Emergency Dose Calculation Manual (EDCM) '7 3.23 GROUND RELEASE - An airborne effluent plume which contacts the ground essentially at the point of release either from a source actually located at the ground elevation or from a source well above the ground elevation which has significant building wake effects- to cause the plume to be entrained in the wake and driven to the ground elevation is termed a ground level release. Ground level releases are treated differently than elevated releases in that the X/Q calculation results in significantly higher concentrations at the ground elevation near the release point.
Ground releases also have generally lower X/Qs all the way downwind.
3.24 HYDROGEN PURGE SYSTEM - Post-accident containment purge system is designed to maintain the hydrogen concentration of the post-accident containment atmosphere below the lower flammability limit. The system does this by introducing outside air into the Reactor 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 rties exist which can be additive to give flow from 5 to 1250 CFM.
3.25 INTERIM SOLID HASTE STAGING FACILITY (ISHSF/ PAINT SHED) and the TMI-2 PAINT SHED OR RADWASTE MATERIAL STORAGE FACILITY (RMSF) - These facilities have no ventilation system or radiation monitor, but have the potential to release radioactive material to the environment.
V 3.26 LOW POPULATION ZONE (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 event of a serious accident.
3.27 -METEOROLOGICAL INFORMATION ANP DOSE ASSESSMENT SYSTEM (MIDAS) - This is the acronym for the computer program that can be used by the Environmental Assessment Command Center (EACC) to project off-site doses for routine effluents 4nd releases during emergencies. 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 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 derived from the calculated radionuclides concentrations 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".
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~' v ) Title TMI Emergency Dose Calculation Maaual (EDCM) 8 3.29 0FF-CENTERLINE DOSE CALCULATIONS - Dose calculations that are calculated i ' at various distances from the calculated plume centerline (0, 50, 100, 150, 250, and 500 meters). These calculations are performed at 28
. distances from the plant.
I 3.30 0FFSITE AIR SAMPLE ANALYSIS SYSTEM - An air sampling and analysis system specifically designed for iodine air sampling and.. thyroid dose F assessment. The system consists of an air pump unit which draws air f through a canister containing a highly efficient iodine adsorbing material and a Geiger Mueller detector for canister evaluation.
3.31 PARTITION FACTOR - (Condenser), see NUREG-0017 3.32 POST-ACCIDENT SAMPLING SYSTEM (PASS) - System used for acquiring a pressurized liquid sample of the RCS during emergency conditions. The post-accident reactor coolant sampling system provides a means of obtaining a representative sample of reactor coolant, including dissolved gases, reactor coolant bleed tank contents and reactor containmer.t sump
. contents, within one hour after the decision to acquire the sample, without excessive operttor exposure or compromise of interfacing safety-related systems.
a 3.33 PROTECTIVE ACTION GUIDE (PAG) - Projected radiological dose or dose V commitment values to individuals of the general population and to emergency workers that warrant protective action before or after a release of radioactive material. Protective actions would be warranted provided the reduction in individual dose expected to be achieved by carrying out the protective action is not offset by excessive risks to individual safety in taking the protective action. The protective action guide does not include the dose that has unavoidably occurred prior to the assessment.
3.34 PROTECTIVE ACTION RECOMMENDATION (PAR) - Those actions taken during or after an 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.
3.35 RADIATION INSTRUMENT SHOP - Facility used to repair / calibrate / maintain instrumentation used by Radiological and Environmental Controls Departments. This facility includes calibration sources that may possibly be released in a wc$rst case accident (e.g., fire). This facility has a ventilation system, (rated at 4000 CFM), but no installed radiation monitor, or filtration.
3.36 RADIATION HONITORING SYSTEM (RMS) - The RMS detects, indicates, annunciated, and records the radiation level at selected locations inside and outside the plant to verify compliance with applicable Code of Federal Regulations (CFR) limits. The RMS consists of the following subsystems:
area monitoring, atmospheric monitoring, and liquid monitoring.
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THI Emergency Dose Calculation Manual (EDCH) 7 3.37 RADI0 IODINE PLATE 0VT - Iodines are very chemically reactive, being members j of the halogen family. As such, iodines have a very high probability of reacting with almost any other material they come in contact with.
Radioiodine plateout is a generic term for the mechanisms by which radioactive iodines are removed from a waste stream by contact with materials not specifically designed or engineered for radioiodine removal. Examples of potential radioiodine plateout reactions are the removal of todine from gaseous wastes by adsorption onto interior surfaces of ductwork and piping and on any exposed surfaces of the room or building originating the release.
3.38 RADIOI0 DINE PROCESSOR STATIONS (MAP-5) - System used for acquiring particulate and iodine samples from the Reactor Building Exhaust,.
Auxiliary and Fuel Handling Building Exhaust or the Condenser Off-gas Exhaust during emergency conditions. The stations are controlled by solenoid valves which activate on high alarm indications on the low gas channels of the effluent stream. Flow is actuated through (3) parallel filter cartridges per station. The sampling times are adjustable on each local control panel. The filter cartridges must be removed manually for analysis, n 3.39 RADIOLOGICAL ASSESSMENT COORDINATOR (RAC) - The RAC is responsible for all
( on-site radiological assessment activities. Initially, the RAC is i responsible for directing the en-site and off-site survey teams. The RAC is relieved of off-site radiological monitoring responsibilities by the Environmental Assessment Coordinator. The RAC performs dose projections, based upon source term estimates and provides information to the EAC.
Initially the Group Radiological Control Supervisor assumes the role of the RAC until relieved by the Initial Response Emergency Organization RAC, and RASE.
3.40 RADIOLOGICAL ASSESSMENT SUPPORT ENGINEER (RASE) - Individuals assigned to asstst the RAC in performing dose calculations, source term calculations, and overall assessment and control of ra Q logical hazards. Normally one RAC and one RASE are on duty at all times.
3.41 REACTOR COOLANT SYSTEM (RCS) - This system contains the necessary piping and components to provide sufficient water flow to cool the reactor. This system provides for the transfer of thermal energy from the reactor core to the once through steam generators (OTSG) to make steam, acts as a moderator for thermal fission, and provides a boundary to separate fission products from the atmosphere.
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3.42 RELEASE DURATION - Release duration refers to the time interval during w'ich radionuclides are released from the nuclear facility. Releases may .
be monitored, unmonitored, actual, or projected. The time interval used I to estimate a release of unknown duration should reflect best estimates of '
the plant technical staff. In the absence of other information, use two
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TMI Emergency Dose Calculation Manual (EDCM) 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> as the expected release duration. 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 time the dose projection is done.
3.43 RELEASE RATE - This term refers to the rate at which radionuclides are released to the environment. Normally, it will be expressed in curies per second (C1/sec) or microcuries per second (pC1/sec).
3.44 RESPIRATOR AND LAUNDRY MAINTENANCE FACILITY (RLM) - This facility is used to process clean and maintain laundry and respirators for THI-1 and TMI-2. This facility's 900 CFM ventilation system is monitored using a Eberline PING radiation monitor.
3.45 RMS RESPONSE FACTOR - Parameter which is used to convert RMS monitor count rates to total microcurie per cubic centimeter of the assumed or measured radionuclides spectrum passing by the monitor. This is different from a meter calibration factor which does the same thing for a single calibration nuclide. These factors are adjusted for changes in mixture decay.
O 3.46 SEMI-INFINITE PLUME MODEL - Dose assessment model which is based on the
(' assumption that the dimensions of an effluent plume are large compared to j
the distance that gamma rays can travel in air. If the plume dimensions i are larger than the gamma ray range, then the radius of the plume might just as well be infinite since radiation emitted from beyond a certain distance will not reach the receptor. The ground is considered to be an infinitely large flat plate and the receptor is assumed to be standing at the center of a hemispherical cloud of infinite radius. The radioactive cloud is limited to the space above the ground plane. This is the origin of the name SEMI-INFINITE 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 diffusion equation which determine horizontal and vertical diffusion. Sigma-Y and Sigma-Z varies by stability class and distance from release point.
i 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
" 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" is that portion of the equation which addresses the activity released.
Although the term " Source Term" is used loosely to mean almost any activity for airborne, 11gulds, and even dose rate calculations in plant, strictly speaking " Source Term" applies only to radioactive material p actually released.
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%d TMI Emergency Dose Calculation Manual (EDCM) 7 3.49 SPLIT HAKE RELEASE' .Altborne releases, for purposes of assessing off-site dispersion, must address the elevation of the release since wind speed changes with height, buildings affect. dispersion for low releases and even wind direction.can t,e different. Many release points are actually at a height where, given different conditions of release flow rate and meteorology, could either be most accurately described as ground or elevated releases, or some mixture between the two. The purpose of
. treating a release'as a split wake release is to address this problem.
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 reletse. In intermediate conditions, the model will " split" the release between ground and elevated as appropriate, so that a release might be 25% ground and 75% elevated from the same release point.
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 Pre p) . stable. Pasquill stability class is a breakdown of the relative (V atmospheric stability into seven groups, denoted as A through G, from most unstable to most stable. In the pasquill stability class system, stability is related to the relative change in temperature with height, delta T. The more negative the change in temperature with increasing height, the more unstable the atmosphere. The RAC program uses sensors on the Meteorological tower at 33 feet and 150 feet to determine the delta T. Once the delta T is determined, a stability class is selected based on the delta T and the atmospheric dispersion (X/Q) is calculated based on the selected stability class.
3.51 STATION VENT HPR-219 - This radiation monitor and release pathway is the main release point for TMI-2. All ventilation from the Reactor Building, Auxiliary Building, and fuel Handling Building are 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 * (T HPR-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-2 Turbine Building. This unit is used in conjunction with HPR-219 to monitcr the main TMI-2 release pathway.
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v TMI Emergency Dose Calculation Manual (EDCM) 7 3.53 TERRAIN FACTOR - The terrain factor is the terrain height in meters above 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 Following discharge to the environment the steam fraction relief valves.
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's 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 dependent on the stability class. 1 f% ~
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Radiological Controls Department 9100-PLN 4200.02 Revision No.
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4.0 PREREQUISITES 1
, 4.1 The following are the prerequisites for performance of TMI projected doses I i using the methods in the EDCM, and the current TMI-1 or TMI-2 RAC Program.
4.1.1 The Emergency Plan is being implemented.
4.1.2 The RAC station is manned and functional.
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TMI Emergency Dose Calculation Manual (EDCM) 8 5.0 PROCEDURE This section"of the EDCM is divided into the programs that are contained in the
~
RAC computer programs for THI-1 and TMI-2. Listed below is a table of contents for the procedure section of the EDCM:
5.1 THI-l Source Term Calculations 5.2 Selection of Release Pathways and Characteristics
, 5.3 Calculation of NRC Damage Class and Isotopic Percentages 5.4 Radiation Monitoring System (RMS) Source Term Calculation 5.5 Post Accident Samples Source Term Calculation 5.6 Contingency Calculations Source Term Generation 5.7 Decay Scheme Calculation j 5.8 Noble Gas to Iodine Ratio Calculations L( 5.9 Effluent Release Flow Rates 5.10 Two-Phase Steam Flow Determination j 5.11 Source Term Flitration 1
5.12 Meteorology Inquiry 5.13 Dispersion Model 5.14 Offsite Air Sample Analysis.
i 5.15 -Liquid Release Calculation j 5.16 Protective Action Recommendation Logic j i
5.17 Dose Projection Model Overview, THI-l j 5.18 THI-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 "0N" switch. Adjust brightness / contrast on monitor appropriately.
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fN Radiological Controls Department 9100-PLN-4200.02-
-Title- Revision No.
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- 2. Check that modem is on (for microcom modem - DTR and mail mode carrots on).
3 .' Turn on printer (ON/0FF switch).
- 4. Insert THI-l or THI-2 Disk into the A disk drive 1(3.5"~ disks),
- a. Turn on computer (right side ON/0FF switch).
-b. ~RAC.' program will load within 1 minute.
....--------- = ---.-------...--____------ ___ .....---- -
NOTE: To reload or interrupt program with the computer ON - hit the Ctrl, Alt, Del keys at the same time, with disk in the PC.
- 5. For computers with RAC program on a hard drive:
.- a . Turn on computer.
- b. Menu will appear, choose appropriate RAC Program, or r
{
.c. If no menu, use directory RAC1 for THI-1 or RAC2 for TMI-2.
A
- d. Type.RAC.
- 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 1-2 minutes and will print when l- completed.
- 9. When done: Remove all disks, turn off CRT, computer, and printer. Store disks appropriately.
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'TMI Emeroency Dose Calculation Manual (EDCM) 7 t ..
5.1 Source Term Calculations - The source term portion of the TMI-l dose
! assessment program is used to generate the quantity and radionuclides 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 used to calculate radionuclides release factors and assess the accident consequences.
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7 5.2 Selection of Release Pathways and Characteristics - The TMI-1/ computers l-program will prompt for the release pathway and the release characteristics.
l- 5.2.1 The.following are-the Release Pathways:
- l. OTSG Tube Rupture 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.
~
- 1. OTSG Tube Rupture via condenser off-gas -l
- r
- 2. 0TSG Tube Rupture directly to atmosphere via the Main Steam
(- Reliefs or Atmospheric Dump Valves
- 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 Re. leases
- 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 genr.ation:
- 1. Use RMS
- 2. Use Post Accident Sample Result 4
- 3. Use Contingency Calculation O 20.0 1572c X
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Muaer 7'N '. Radiological Controls Department 9100-PLN-4200.02-1 I.= Title- Revision No.
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'5.2.4- The THI-1 RAC computer program accommodates airborne' releases from the following pathways (See Figure 5.2-1):
A. The OTSG Tube Rupture.such as:
.l.- RM-A5 Condenser Off-gas-
- 2. RM-A5 High-Condenser Off-gas
- 3. RM-G25 Condenser Off-gas
-4. RM-G26 Main Steam Reliefs and Atmospheric Dump Valve's'
'5. RM-G27 Maia Steam Reliefs and Atmospheric Dump Valves
- 6. RM-A5 MAP-5 Samples
- 7. Main Steam Release directly to the atmosphere'
- 8. Contingency Calculations without RMS'or Samples n B. The Reactor Building such as:
t f b 1. RM-A9 Reactor.Bu'1 ding 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
- 2. RM-A6 Auxiliary Building Exhaust
- 3. RM-A8 Station Vent (Auxiliary and Fuel Handling Buildings)
- 4. RM-A8 High-Auxiliary Building Exhaust A
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- 5. RM-A14 - ESF Fuel Handling Building Exhaust
- 6. MAP-5 Samples
- 7. Contingency Calculations without RMS or Samples O) t
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TMI Emergency Dose Calculation Manual (EDCM) 7 FIGURE 5.2-1 l CD 1
ATMO5PHERIC RADI ATION MONITORING STAfl0N k[ ACTOR ini.
VENT BL M. PURGE
-t Auz!LI arf & f ufL .
' 2A HANDLlhG BUILDlhG l 0 30 CFM 80,000 (f M kM A. RMA.
8 NI 9 HI REACION - l CAIIA'85 gg, gMA. RMA. bu!LDlhG WDGT's 6 8 9 RMG.8 l 70 ATMO5PHERE - Mg 170.000 UM 50.000 W u gr r-v[hil[A110N p A g C A P "'
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iALvt5 t-RMA-A MAlh 5H AP P-9> 10 TURBihi pyppg RMA+ &MA m IURSlh[
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- Number-jf% Radiological Controls Department 9100-PLN 4200.02 Title. Revision No.
TMI Emeraency Dose Calculation Manual (EDCM) -8 5.3 Calculation of NRC Damage Class ano Isotopic Percentages _ This
' calculation will determine the mix or percentages of the following fifteen rr onuclide s .
' 5. Ten Noble Gases Five Radiciodines
- 1. Kr-85m 1. I-131
- 2. Kr-85 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
- 8. Xe_135m
) 9. Xe-135
- 10. Xe-138
.5.3.2 The NRC Damage Class Determination The determination of the NRC Damage Class is performed using various core temperature regions from Operations Procedure 1210-8, see Figure 5.3-1. The core temperatures used in this section of the program come from operations pt C4006, which is the average of the five highest incore thermocouple.
The curves relating to saturation, and cladding failures are approximated by straight line equations. NRC damage classes 1 -
l- 10 are based on the different pressure and temperature regions of Figure 5.3-1.
l NOTE: A 57 allowance is made for the accuracy of the average of l 5 incore temperatures from the C4006 reading from Operations.
___ - _ _ _ _ _ _ _ = - - _ _ _ _ _ _ _ . _ . _ _ ----____-- - - - - _ _ _ _ _ - - -_
5.3.2.1 Core Temperature Regicos - Figure 5.3-1 The region to the left of curve C represents normal RCS activity, NRC Class-1. The region between curves C and D
-- Q represents RCS plus a percentage of gap activity, NRC Class 2 _
- 4. The region between Curve D and Curve E represents RCS plus l
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Nm-TMI Emergency Dose Calculation Manual (EDCM) 8 all gap activity plus a percentage of noble and volatile fission product release from fuel grain boundaries, (CS, I, Rb), NRC Class 5 - 7. The region between Curve E and 2550*F incore temperature represents RCS activity plus 100% of the gap activity and 100% of the in vessel melt release assuming NUREG-1228 release fractions, NRC Class 8 - 10.
5.3.2.1.1 The matrix below shows the theory of fuel damage based on TDR-431.
NRC DAMAGE CLASS DEGREE OF MINOR INTERMEDIATE MAJOR DEGRADATION <10% 10 - 50% >50%
No Fuel Damage (RCS) -No Damage : Class 1 -
Cladding Failure (GAP) 2 3 4 Fuel Overheat (Fuel Matrix) 5 6 7 Fuel Melt (Fuel Matrix) 8 9 10 5.3.3 Calculation of Radionuclides Mix Percentages Based on NRC Damage Classification. Once the determination of NRC Damage Class,
,m 1 - 10 has been determined from the Core Temperature Regions the l
\ various radionuclides mix percentages can be calculated based on
( ,/ i the distribution of the RCS Activity, GAP Activity, and/or fuel g Matrix Activity. The program models the various combinations of activities for each NRC Damage Class as follows:
RCS GAP FUEL NRC DAMAGE ACTIVITY ACTIVITY MATRIX CLASS FRACTION FRACTION FRACTION 1 1 0.0 0.0 2 1 0.1 0.0 3 1 0.5 0.0 4 1 1 0.0 5 1 1 0.1 6 1 1 0.5 7 1 1 1 8 1 1 1 9 1 1 1
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TMI Emergency Dose Calculation Manual (EDCM) 8 j Therefore, as an example: NRC Damage Class-6 would consist of 100% RCS Activity, plus 100% of the GAP Activity, plus 50% of the fuel Matrix Activity.
5.3.3.1 THI-1 Normal RCS Activity - THI-1 normal RCS Activity (NRC Damage Class-1) and the rapid power transient RCS Activity (NRC Damage Class-1A) are listed below. The rapid power transient RCS activity represents the " spiking" of the iodines and noble gases during a rapid transient. (Defined as a 1 0% 1 power change over one minute). This activity is 100 X the normal RCS activity.
TMI-1 NORMAL RCS ACTIVITY
- NORMAL RCS RAPID POWER TRANSIENT RCS DAMAGE DAMAGE CLASS 1 CLASS 1A UC1/cc Percent Curies ** UCt/cc Percent Curtes**
/N I-131 4.83E-02 0.56 1.04E+01 4.83E+00 0.56 1.04E+03
( ) I-132 1.91E-01 2.21 4.09E+0! 1.91E+01 2.21 4.09E+03
'\ '
I-133 1.47E-01 1.71 3.16E+01 1.47E+01 1.71 3.16E+03 I-134 3.01E-01 3.48 6.44E+01 3.01E+01 3.48 6.44E+03 I-135 2.16E-01 2.50 4.62E+01 2.16E+01 2.50 4.62E+03 SUBTOTAL 9.03E-01 10.46 193.48 9.03E+01 10.46 1.93E+04 KR-85M 1.54E-01 1.78 3.30E+01 1.54E+01 1.78 3.30E+03 KR-85 0.00E+00 0.00 0.00E+00 0.00E+00 0.00 0.00E+00 KR-87 1.65E-01 1.91 3.53E+01 1.65E+01 1.91 3.53E+03 KR-88 2.91E-01 3.37 6.23E+01 2.91E+01 3.37 6.23E+03 XE-131M 0.00E+00 0.00 0.00E+00 0.00E+00 0.00 0.00E+00 XE-133M 1.14E-01 1.32 2.44E+01 1.14E+01 1.32 2.44E+03 XE-133 5.66E+00 63.57 1.21E+03 5.66E+02 65.57 1.21E+05 XE-135M 1.04E-01 1.20 2.23E+01 1.04E+01 1.20 2.23E+03 XE-135 1.07E+00 12.42 2.30E+02 1.07E+02 12.42 2.30E+04 XE-138 1.70E-01 1.97 3.64E+01 1.70E+01 1.97 3.64E+03 SUBTOTAL 7.73E+00 89.54 1656.80 7.73E+02 89.54 1.66E+05 TOTAL 8.64E+00 100.00 1.85E+03 863.763 100.00 1.85E+05 NOBLE GAS 8.56 8.56 8.56 8.56 8.56 8.56 I TO IODINE
,_s RATIO
('~,) * " NORMAL" RCS CONCENTRATION AND PERCENTAGES ARE FROM NORMAL RCS OPERATIONAL DATA i
- THESE CURIE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS 26.0 1572c
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5.3.3.2 Gap Activity and fuel Matrix Activity - Gap activity and Fuel Matrix activity used in the program are determined from TDR-431. These curie activities are based on irradiation of the entire core at full power, 2535 MW ,t for 930 days.
TMI Unit 1 GAP Fuel Matrix Act. Act.
Isotope Curies Curies Kr-85m 4.84E+04 2.13E+07 Kr-85 7.48E+04 8.59E+04 Kr-87 2.63E+04 3.90E+07 Kr-88 6.67E+04 5.91E+07 Xe-131m 7.96E+04 5.40E+05 Xe-133m 9.30E+04 3.09E+06 Xe-133 8.34E+06 1.28E+08 Xe-135m 2.72E+04 3.37E+07
, Xe-135 3.45E+04 1.59E+07 l' Xe-138 0.00E+00 0.00E+00 I-131 1;29E+06 6.37EA7 (m'x_-) 1-132 I-133 I-134 1.85E+05 2.79E+05 1.74E+04 9.70E+07 1.43E+08 1.67E+08 I-135 8.83E+04 1.30E+08 Sum 1.13E+07 Sum 9.02E+08 G
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TM1 Emergency Dose Calculation Manual (EDCM) 8 l j
E.3.3.3 The following tables represent tne NRC Damage Class 2 - 10 mixes, percentages, curies, and concentrations used in the RAC .
code, i a
DAMAGE DAMAGE CLASS 2 CLASS 3 uC1/ct Percent Curies"* UCi/cc Percent Curies **
I-131 6.03E+02 12.09 1.29E+05 3.01E+03 12.11 6.45E+05 I-132 8.66E+01 1.74 1.85E+04 4.32E+02 1.74 9.25E+04 I-133 1.31E+02 2.62 2.79E+04 6.52E+02 2.62 1.40E+05 I-134 8.43E+00 0.17 1.80E+03 4.10E+01 0.16 8.76E+03 I-135 4.15E+01 0.83 8.88E+03 2.07E+02 0.83 4.42E+04 SUBTOTAL 8.70E+02 17.45 1.86E+05 4.35E+03 17.46 9.30E+05 KR-85M 2.28E+01 0./6 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 KR-88 3.15E+01 0.63 6.73E+03 1.55E+02 0.63 3.34E+04 XE-131M 3.72E+01 0.75 7.96E+03 1.86E+02 0.75 3.98E+04 fm XE-133M 4.36E+01 0.87 9.32E+03 2.17E+02 0.87 4.65E+04 (d
I XE-133 XE-135M XE-135 3.90E+03 1.28E+01 1.72E+01 78.29 0.26 0.34 8.35E+05 2.74E+03 3.68E+03 1.95E+04 6.37E+01 8.17E+01 78.30 0.26 0.33 4.17E+06 1.36E+04 1.75E+04 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 SUBTOTAL 4.12E+03 82.55 8.81E+05 2.05E+04 82.54 4.40E+06 TOTAL 4.99E+03 100.00 1.07E+06 2.49E+04 100.00 5.33E+06 NOBLE GAS 4.73 4.73 4.73 4.73 4.73 4.73 TO IODINE RATIO
- THESE CURE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS 28.0 1572c
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Number i- f Radiological 1' Controls Department 9100-PLN-4200.02
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. - \s ,
TMI Emergency Dese Calculation iual (EDCM) 8 DAMAGE DAMAGE CLASS 4 CLASS 5 UC1/cc Percent Curies ** aCi /cc- Percent Curies **
I-131 6.03E+03 12.11 1.29E+06 3.58E+04 7.60 7.66E+06 I-132 8.65E+02 1.74 1.85E+05 4.62E+04 9.81 9.89E+06 I-133 1.30E+03 2.62 2.79E+05 6.81E+04 14.46 1.46E+07 I-134 8.16E+01 0.16 1.75E+04 7.81E+04- 16.59 1.67E+07 I-135 4.13E+07 0.83 8.83E+04 6.12E+04 12.99 1.31E+07 5UBTOTAL 8-69E+03
. 17.46 1.86E+06 2.89E+05 61.44 6.19E+07 KR-85M 2.26E+02 0.45 4.84E+04 1.02E+04 2.16 2.18E+06 KR-85 3.50E+02 0.70 7.48E+04 3.90E+02 0.08 8.34E+04 KR-87 1.23E+02 ~0.25 2.63E+04 1.83E+04 3.90 3.93E+06 KR-88 3.12E+02 0.63 6.68E+04 2.79E+04 5.93 5.98E+06 XE-131M 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
, XE-133 3.90E+04 78.31 8.34E+06 9.88E+04 20.97 2.11E+07
-XE-135M 1.27E+02 0.26 2.72E+04 1.59E+04 3.37 3.40E+06 XE-135 1.62E+02 0.33 3.47E+04 7.59E+03 1.61 1.62E+06
,, XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01
( SUBTOTAL 4.11E+04 82.54 8.79E+06 1.82E+05 38.56 3.89E+07 TOIAL 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 CURE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS
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en ex Radiological Controls Department 9100-PLN-4200.02 l
( ) Title Revision No.
L.J TMI Emergency Dose Calculation Manual (EDCM) 8 DAMAGE DAMAGE CLA:S6 CLASS 7 uCi/cc Percent Curles** UCi/cc Percent Curies **
l-131 1.55E+05 7.18 3.31E+07 3.04E+05 7.13 6.50E+07 I-132 2.28E+05 10.55 4.87E+07 4.54E+05 10.66 9.72E+07 I-133 3.35E+05 15.56 7.18E+07 6.70E+05 15.71 1.43E+08 I-134 3.90E+05 18.10 8.35E+07 7.80E+05 18.31 1.67E+08 I-135 3.04E+05 14.11 6.51E+07 6.08E+05 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 1.07E+07 9.98E+04 2.34 2.13E+07 KR-85 5.50E+02 0.03 1.18E+05 7.51E+02 0.02 1.61E+05 KR-87 9.12E+04 4.23 1.95E+07 1.82E+05 4.28 3.90E+07 KR-88 1.38E+05 6.42 2.96E+07 2.76E+05 6.49 5.92E+07 XE-131M 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 3.37E+07 XE-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. 645:+01 1.70E-01 0.00 3.64E+01 b SUBTOTA'. 7.44E+05 34.50 1.59E+08 1.45E+06 33.93 3.10E+08 TOTAL 2.16E+06 100.00 4.61E+08 4.26E+06 100.00 9.12E+08 NOBLE GAS 0.53 0.53 0.53 0.51 0.51 0.51 IODINE RATIO
- THESE CURE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS
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f3 Radiological Controls Department 9100-PLN-4200.02 j
( . Title- Revision No.
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LTMI Emeraency Dose Calculation Manual (EDCM) 8 1
DAMAGE DAMAGE CLASS 8 CLASS 9 UCi/cc Percent Curies ** UCi/cc Percent Curies **
. 1-131 3.04E+05 7.13 6.50E+07 3.04E+05 7.13 6.50E+07 I-132 4.54E+05 '10.66 9.72E+07 4.54E+05 10.66 '9.72E+07 I-133 6.70E+05 .15.71 1.43E+08 6.70E+05 15.71 1.43E+08 I-134 7.80E+05 18.31 1.67E+08 7.80E+05 18.31 1.67E+08 I-135 6.08E+05 14.26 1.30E+08 6.08E+05 14.26 '1.30E408 SUBTOTAL 2.82E+06 66.07 6.03E+08 2.82E+06 66.07 6.03E+08 KR-85M 9.98E+04 2.34 2.13E+07 9.98E+04. 2.34 2.13E+07 KR-85 7.51E+02 0.02 1.61E+05 7.51E+02 0.02 1.61E+05 KR-87 1.82E+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+05 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 1.36E+08 6.37E+05 14.95 1.36E+08
'XE-135M. 1.58E+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 1.59E+07
- i. - XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01
( SUBTOTAL 1.45E+06 33.93 3.10E+08 1.45E+06 33.93 3.10E+08 TOTAL 4.26E+06 100.00 9.12E+08 4.26E+06 100.00 9.12E+08 NOBLE GAS 0.51 0.51 0.51 0.51 0.51 0.51 IODINE RATIO
- THESE CURE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56.595 GALLONS d
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~'g Radiological Controls Department 9100-PLN-4200.02 L j Title Revision No.
v TMI Emergency Dose Calculation Manual (EDCM) 8 DAMAGE CLASS 10 uC1/cc Percent Curies **
I-131 3.04E+05 7.13 6.50E+07 I-132 4.54E+05 10.66 9.72E+07 I-133 6.70E+05 15.71 1.43E+08 I-134 7.80E+05 18.31 1.67E+08 I-135 6.08E+05 14.26 1.30E+08 SUBTOTAL 2.82E+06 66.07 6.03E+08 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 XE-133 6.37E+05 14.95 1.36E+08 XE-135M 1.58E+05 3.70 3.37E+07 XE-135 7.45E+04- 1.75 1.59E+07 (3 XE-138 1.70E-01 0.00 3.64E+01
's SUBTOTAL 1.45E+06 33.93 3.10E+08 i
TOTAL 4.26E+06 100.00 9.12E+08 NOBLE GAS 0.51 0.51 0.51 IODINE !
RATIO
- THESE CURE VALUES ARE BASED ON NORMAL RCS VOLUME OF 56,595 GALLONS 5.3.3.4 Calculation of Radionuclides Mix Percentages - Once the computer hus determined the total amount of combined activities or curies for a certain NRC Damage Class, these curies are then normalized to 100%, i.e., the percentage of each radionuclides in the total mix is calculated. These percentages are then displayed on the screen along with the NRC Damage Class, 1 - 10. I Exceptions - The above percentages are replaced in cases where an assumed mix is more appropriate. These cases are:
- 1. Contingency calculations for:
l
- a. Spent Fuel Accident in the Fuel Handling Building -
FSAR mix is assumed.
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,'~'s: Radiological Controls Department 9100-PLN 4200.02 Revision No.
(n) ~ Title TMI Emergency Dose Calculation Manual (EDCM) 8
- b. Fuel Cask Accident in the Fuel Handling Building -
FSAR mix is assumed.
- c. Spent Fuel Accident in the Reactor Building - FSAR mix is assumed.
- d. Haste Gas Decay Tank - FSAR mix is assumed.
- e. LOCA in Reactor Building using NRC Damage Class Default concentration.
- f. OTSG tube rupture directly to atmosphere using NRC r Damage Class Default concentration.
- g. OTSG tube rupture via condenser off gas Jsing NRC Damage Class Default concentration.
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] Title Radiological Controls Department Revision No.
b'( f TMI Emergency Dose Calculation Manual (EDCM) 8 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-l Radiation Monitoring System.
5.4.1 Only those RMS channels available for a selected release pathway '
- are offered to the user. These are listed.in Section 5.2. To calculate a source term from a RMS reading the following parameters are used
- 2. RMS CHANNEL EFFICIENCY RELATING TO THE CALIBRATION NUCLIDE: CPM /pCI/CC
- 3. THE METER RESPONSE FACTOR
- 4. THE NRC DAMAGE CLASS MIXTURE
. S '. .THE RELEASE FLOW RATE l 5.4.2 In order to gather the above information the program will proceed in the following manner.
l
\ 1. Once a release pathway has been chosen the first option-to l the user is whether or not to decay the mixture from the j time of reactor shutdown. If "yes" is chosen the program !
decays the eventual mixture based on the hours input by the i user.
- 2. The appropriate radiation monitors for the pathway chosen 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 l monitor reading in counts per minute, mR/ hour, or counts l per minute per minute (based on a rise time). i
- 4. The next prompt is for the flow rate for the release pathway. The applicable flow rates are discussed in the effluent flow rate section. Appropriate default values are i also listed.
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- 5. Once the above data has been entered,'the program wil);use.
1the RMS reading, the particular radiation ~ monitor's efficiency, the moniter r9sponse factor, the nuclide fraction from the isotopic perc9ntage section of the program relating to NRC damage' class' determination, .
radionuclides 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 rad!olodines.
5.4.3 The~ calculations are performed in the following fashion:
- 1. First the total monitor response factor is calculated by multiplying the' individual.nuclide percentages from the NRC damage clas's determination by the individual nuclide-monitor re';ponse factors.
15 M= I P.* In 1
^im Nhere: M - total monitor' response factor
{
P - individual nuclide percentages from NRC damage class In - individual nuclide monitor response factors The In's for the'various RMS detectors are listed as follows:
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-N, Radiological Controls Department 9100 DLN-4200.02 (sj ) Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 INDIVIDUAL MIXTURE RESPONSE FACTORS (In)
Beta Scintillation Detectors Scintillation Ion GH Tubes RM-A2Lo, RMA4Lo, Detectors Chamber RM-A5Hi, RM-A8Hi(3), RM-A6Lo, RM-A5Lo, NUCLIDE RM-G26 & RM-G27(1) RM-G24(2) RM-A9Hi, RM-G25 RM-A8Lo. RM-A9Lo(4)
Kr-85m 75.5 212.2 2.35 1.49 Kr-85 1 1 0.011 2.36 Kr-87 687 324.39 3.59 10.98 Kr-38 872 334.15 3.70 8.09 Xe-131m 1.5 4.88 0.054 0 Xe-133m 16.5 35.15 0.378 0 Xe-133 0 90.24 1 1 gy Xe-135m 211 195.12 2.16 0 N Xe-135 123 221.95 2.54 2.66 Xe-138 1434.5 939.02 10.41 6.94 I-131 185 240 2.66 0 I-132 1194.5 747.8 8.286 0 I-133 1433 219.51 2.432 0 I-134 1008 542.44 6.011 0 1-135m 887.5 341.46 3.784 0 (1) MeV/ Dis Iso.
MeV/ Dis cal iso. (calibration isotope is Kr-85, threshold set to exclude Xe-133 at 80 kev.)
(2) I% Probability Iso.
I% Probability Cal. Iso. (calibration isotope is Kr-85)
(3) I% Probability 1so.
fi Probability Cal. Iso. (calibration isotope is Xe-133; values for RM-ASHi, except Xe-133, will be multiplied by 4)
,7 '} (4) _I Beta decay probability
- Bete end-pt. energy isotope i I Beta decay probability
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fT - Radiological Controls Department 9100-PLN 4200.02
,iKj):' . Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8
- 2. The noble gas source term.in pC1/sec is now calculated using the following equation and input data:
Ngst - ( 1
- ACT * 'l
- Ng/100) * (Flow)
- 472 M Me Where: Ngst - Noble Gas source term in pCi/sec.
M = Total monito: response factor.
Act - cpm, mR/hr, cpm / min reading from the monitor.
He - Monitor sensitivity in cpm /pC1/cc, mR/Hr/pC1/cc, or cpm / min /pC1/cc.
F!ow'. Flow rate in CFM 472 = cc/sec/CFM.
Ng - Sum of Noble gas percentages from thel selected NRC classification.
t' 3. The radiciodine source term in pC1/sec is then calculated using the
( noble gas source term and the noble gas to iodine ratio, as discussed in Section 5.8.
Rist - Ngst
- Ri *l-Ng Tfcf (if applicable)
Where: Rist - Radioiodine source term in pCi/sec.
R1 The radiolodine to noble gas ratio.
Ng Tfcf - Two phase steam factor - If this is a steam release with water; the water will tend to keep the radioiodines in solution.
- 4. The noble gas and the radiolodine source terms are then multiplied by the individual isotopic percentages of the NRC damage class mixture to determine the pC1/sec of each of the 15 nuclides, 10 noble gas and 5 radiolodines.
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THI Emergency Dose Calculation Manual (EDCM) 8 5.5 Post Accident Samples Source Term Calculation - One option of the RAC program is to use actual plant effluent sample results to develop the release source terms. This is in fact the preferable method for estimating release quantitles if the sample results are available since the method eliminates some of the built in conservatism of using monitor readings, or contingency calculations. Using sample results also eliminates errors in the source term when the actual release mixture is different from the assumed mix. The routines which allow use of the post accident samples contained in the RAC programs provide the. menu selectors to call the different subroutines for each type of post accident sample, 5.5.1 One menu selection is the sample station / method to be used. For the Reactor Building three options are presented: 1) CATPASS (Contal u nt Atmosphere Post Accident Sample System),
- 2) Marinelli/prefilter (marinelli with a particulate and iodine filter upstream), or 3) MAP-5, Radiolodine Processor Station.
For the condenser off-gas or the Aux /FHB release pathways, only the Marinelli/prefilter and MAP-5 samples are available. A menu will appear on the computer screen which will list the two or three available sample methods and prompt with ' enter choice'.
When a choice - 1, 2, or 3 as appropriate - is entered the gm program will continue.
\ 5.5.2- The MAP-5 program will prompt for each of the identified radiolodine species in the silver zeolite sample from the MAP-5 Processor Station. It will place the value for each of the five radiolodine species into one of the elements of the five element array in microcuries per cc. It will then sum the results and 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 CORRECTION WOULD NOT BE APPLIED FROM TIME OF SHUTDOWN SINCE THE ANALYSIS ITSELF ACCOUNTS FOR THAT. If'the release is from the Reactor Building, an option selection is provided to determine if the Reactor Building is isolated or not and if the release is proposed or in progress. It then will adjust the noble gases. Since the MAP-5 only provides information on the radiolodines, the expected ratio between the lodines and noble gases is used to approximate the noble gas activity. Sinct' the MAP-5 is downstream of the charcoal filters, if the charcoal filters are effective, the noble gas activity will be increased by a factor of ten to account for the filter reduction of the lodines. Based on the release rate in CFM the isotopic concentrations in pCi/cc are converted to a release rate in pC1/sec for the final source term.
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v TMI Emergency Dose Calculation Manual (EDCM) 8 5.5.3 The CATPASS program prompts for the 10 noble gas and the five radiolodine nuclides identified from air sampling. The noble gas and iodine activities in microcuries per cc are put into an 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 whether the containment is isolated or not and if the release is i proposed or in progress. Calculations are made using the input activities to develop new isotopic percentages and the activities are changed from pC1/cc to pC1/sec based on the release rate defined to arrive at the final source term. ;
5.5.4 The marinelli program is called if the marinelli/prefilter 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 CATPASS program, the marinelli program proceeds in an identical manner, m
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5.6 Contingency Calculations Source Term Generation - The contingency ,
calculations attempt to determine a source term based upon a prioritized !
set of plant conditions. The user is guided through a set of questions in order to model the contingency calculation with the best obtainable information. In this way, credible conservative assumptions, as defined in the FSAR default parameters, are replaced with real-time accident conditions as indicated by plant instrumentation. This will make the calculated source terms more realistic.
5.6.1 In the contingency program the previously determined release pathway is utilized to select:
- 1. Secondary Side Release
- 2. Reactor Building Release
- 3. Station Ventilation 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 itne rupture.
,s The " Reactor Building Release" includes accidents that result in
[ a release from the Reactor Building via: the purge duct, when ;
l v j the purge valves are open, or design basis leakage, when the purge valves are closed.
The " Station Ventilation Release" includes accidents that result in a release from the Auxiliary or Fuel Handling Buildings. i 5.6.2 A " Secondary Side Release" is calculated by identifying four parameters:
- 1. RCS Activity [Dl] pCi/cc
- 2. Primary to Secondary Leakage [D2] gpm l l
- 3. Transport Fraction [D3]
- 4. Two Phase Release [Tfcf]
5.6.2.1 The "RCS Activity" is determined utilizing:
- 1. RM-L1 High [All cpm, D1 - Al/22.2 pC1/cc ;
- 2. RM-L1 Lo [All cpm, D1 - Al/1270 pC1/cc
- 3. Most recent RCS sample, in pCi/cc, or
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Number f~'s Radiological Controls Department 9100-plN-4200.02 (q,t [ Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8
< 4. Default to a RCS concentration dependent on the NRC Damage Class:
RCS Default NRC Damage Class uCi/cc 1 8.64E+00 1A 8.64E+02 2 4.99E+03 3 2.49E+04 4 4.98E+04 5 4.71E+05 6 2.16E+06 7 4.26E+06 8 4.26E+06 9 4.26E+06 10 4.26E+06 5.6.2.2 The " Primary to Secondary Leakage" is determined utilizing:
- 1. RCS identified leakage [D2] gpm n
- 2. Default to 400 gpm for a double-ended tube shear [D2]
(V)- The " Transport Fraction" [03] is a function of the release 5.6.2.3 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 condenser off-gas the noble gas transport is 1.00, the radioiodine transport fraction is 0.0075. The radiciodine transport fraction is a product of: The fraction of radioiodine entering the OTSG 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 zero in the condenser off-gas. For a release direct to atmosphere the noble gas and radiciodine transport fractions use the fraction of steam released to total steam flow from the 3
OTSG. An additional partition factor ITfcf] is applicable for l a two phase direct release. The resultant source terms in pct /sec are calculated by:
Ngst D1
- D2
- Ng/100
- 63.09 Rist = 01
- D2
- D3
- RI/100
- 63.09
- 1/Tfcf.
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. ,y-( jL . Title Revision No.
' TMI Emercency Dose Calculation Manual (EDCM)
-j i 8 5.6.3 A " Reactor Building Release" is calculated by one of two methods. If the accident type is a LOCA then four parame urs are identified:
- 1. RCS Activity [A2] pC1/cc
- 3. Transport Fraction, E4 - 0.1 or 1.0 depending on the F
Reactor Building Spray status
- 4. Release Flow Rate CFM; E3 - flow
- 1. .RM-L1 High Channel [A1] cpm; [A2] = [Al]/22.2 pC1/cc
- 2. RM-L1 Low Channel [A1] cpm; [A2] = [Al]/1270 pC1/cc
- 3. Representative RCS sample results [A2] in pC1/cc
(~.,
4 Default to a RCS concentration dependent on the NRC Damage
.(
R) Class:
RCS Default NRC Damage Class UC1/cc 1 8.64E+00 1A 8.64E+02 2 4.99E+03 3 2.49E+04 4 4.98E+04 5 4.71E+05 6 2.16E+06 7 4.26E+06 8 4.26E+06 9 4.26E+06 10 4.26E+06
- 4. Default mix according to core condition as previously identified. The noble gas and radiotodine released to the Reactor Building is calculated as follows:
El - A2
- A3
- Ng/100
- 3785/5.6E10 pC1 E2 - A2
- A3
- Ri/100
- 3785/5.6E10 pCi A
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TMI Emergency Dose Calculation Manual (EDCM) 8
[
- 5. The RCS leakage to the Reactor Building is determined by requesting the " total gallons of RCS leakage into the RB".
The transport fraction is determined on the basis of the status of Reactor Building Spray. The Noble gas transport ,
fraction is assumed to be 1.00. Radiotodine concentration
[ is reduced as a result of plateout of elemental iodine.
r TM transport fraction for instantaneous radioiodine plateout is 0.5. An additional adjustment of the L
' radiolodine concentration in the Reactor Building is
- necessary when Reactor Building Spray is activated. The spray reduces the iodine concentration in the Reactor Building atmosphere by an assumed 907..
- 6. The release flow rate is determined via flow rate recorder FR-146 if the purge valves are open. If the purge valves are closed the release flow rate is determined via the design basis RB leakrate adjusted for actual RB internal pressure as indicated on PT-291.
5.6.3.2 If the accident type is a Fuel Handling Accident in the Reactor Building, then the number of damaged fuel rods is identified by e the " user" or an FSAR default condition is used.
(s' El - 1.7
- Num rod /208 E2 - 0.05
- Num rod /208 5.6.4 A " Station Ventilation Release" is calculated for:
- 1. LOCA in Auxiliary Building )
- 2. Haste Gas Accident
- 3. Fuel Handling Accident in the Fuel Handling Building, including ESF Fuel Handling Building releases.
5.6.4.1 If a grab sample is available from the affected area, it is used i to determine the source term. If not, the accident is modeled by:
- 1. Determining the number of damaged fuel rods for a fuel handling accident in the Fuel Handling Building.
Ngst . 4.2E6
- Num rods /56 pC1/Sec Rist - 750
- Num rods /56 pC1/Sec l
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TMI Emergency Dose Calculation Manual (EDCM) 8
- 2. Using FSAR assumed conditions for a cask drop accident.
Ngst - 1200 pCi/Sec Rist - 450 pC1/Sec
- 3. Determining the curies released for a Waste Gas Accident.
Typical source term based on a typical inventory of:
Ngst - 1.0E9/Dr pCi/sec Rist - 1.0E5/Dr pCi/sec or FSAR worst case:
Ngst - 1.0E10/Dr pC1/sec Rist - SE6/Dr pCi/sec
,fg Where: Dr - duration of release.
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TMI Emeroency Dose Calculation Manual (EDCM) 8 5.7 Decay Scheme Calculation - The user has the option to (1) decay the postulated mixture from the time of reactor shutdown to time of the dose projection or (2) decay sample data from the time the sample is obtained to the time of dose projection. Subroutine (decay) only decays forward in time. This subroutine (decay) adjusts the individual nuclide percentages according to the conventional exponential decay equation:
A-A exp o (- t) 5.7.1 Fifteen isotopes are decayed according to the equation N(w) - I(w)
- EXP (-decay time
- f(w))
where:
I(w) - postulated isotopic percentage decay time - user input time f(w) - isotopic decay constants read from data files r-~s 5.7.1.1 The adjusted isotopic percentages N(w) are corrected for Xenon
( ) buildup due to iodine decay. For Xe-131m the equations are:
\_/
S1 - I(11) - N(11)
N(5) - 0.88
- S1 + N(5) where:
51 - amount of I-131 decayed 0.88
- S1 - amount of Xe-131M buildup 5.7.1.2 For Xe-133M the equations are:
51 - I(13) - N(13)
N(6) = 0.02
- S1 + N(6) where:
51 - amount of I-133 decayed 0.02
- S1 - amount of Xe-133m buildup N(7) - 0.98
- S1 + N(7) calculates the amount of Xe-133 buildup from I-133 N
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Nuclear TMI au*er Radiological Controls Department 9100-PLN-4200.02 l!(Py- Title. ,
Revision No.
w TMI Emeroency Dose Calculation Manual (EDCM) 8 L. 5.7.'1.3 :For'Xe-135m and Xe-135.the equations are:
51 - I(15) - N(15)
N(8)'- 0.3 *.51'+ N(8)..
N(9) 0.7
- S1 + N(9)
. here:
.w 51 - amount of.1-135 decayed t.
0.3
- S1 - amount of Xe-135m buildup 0.7
- S1 - amount of Xe-135 bullaup 5.7.1.4~ The isot'pic o percentages are recalculated as:
I(w) - N(w)/ Sum.(N)
- 100 where:
N(w) - adjusted / corrected postulated isotopic percentages
.(
s Sum (N) - sum of the fifteen isotopic percentages I(v) - final isotopic percentages based upon 100.
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g- b Radiological Controls Department 9100-PLN-4200.02 Revision No, i Title
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-TMI Emeroency Dose Calculation Manual (EDCM) 8 5.8- Noble Gas to Iodine Ratio Calculations 5.8.1 Whether performing dose'proje'ctions based upon RMS readings, post accident samples or contingency calculations, it may be necessary to compute the NOBLE GAS TO IODINE RATIO. The uses of this ratio are discussed below.
An airborne release from a nuclear power plant will primarily consist of noble gases and radiolodines. Except in the most severe and improbable accident scenarios, radioactive particulate are not expected to be important dose contributors. The RAC program was designed to incorporate ten noble gases and five radiolodines.
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 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
_[m that are specific to the isotopic mixture being released. These
( calculated quantitles can be adjusted to account 'or radioactive decay during the accident sequence.
5.8.2 The THI-I RAC model always projects both thyroid and whole body 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 circumstances where the release rates of specific isotopes may be zero or negligibly small. But, in general, the program accounts for the fifteen isotopes listed above.
In certain circumstances it is not possible to obtain release rates for all fifteen isotopes individually. For example, some plant effluent monitors have only noble gas channels while others have particulate, iodine and gas channels. The MAP-5 sampling system yleids only lodine information, where the CATPASS and the Marinelli gas sampling systems yield information on all fifteen isotopes. For release pathways where information 48.0 1572c I
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' Radiological Controls Department 9100-PLN-4200.02 N
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TMI Emergency Dose Calculation Manual (EDCM) 8 e_ .
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b on.both noble gases and iodines is not available, the RAC program'uses the noble gas to iodine ratio to fill-in the missing information. The following example illustrates the use
.of this ratio:
A certain type of' reactor accident has occurred. Based on an assessment of the degree of core damage and the accident type, the computer selects a default mixture of 15 nuclides and
' calculates the fraction of the mix that each isotope
' represents. The noble gas to lodine ratio is also calculated Assume that the ratio was equal to 5/1 in this case. Also assume that an todine sample was taken which indicated a total radioiodine release rate of 5000 pCi/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 pCi/sec. It would also calculate individual noble gas release rates by using the isotopic fractions from the default mix. I 5.8.3 To summarize, the highest quality information available is a quantitative measurement of each nuclide. This type of information is available from RCS, gas Marinelli and prefilter, and CATPASS samples. So there is no need to invoke the noble
[ gas to iodine ratio in these cases. The second best measurement y"
would be one that yielded gross noble gas and gross iodine readings. This situation occurs in the low range radiation monitors which have individual noble gas and iodine channels.
Based upon the default mixture fractions, the release is apportioned among the fifteen 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.
5.8.4 There are some refinements and subtlettes 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 this manual, the program also corrects the Dose Rate Conversion Factor (DRCF) for decay of the isotopes in the mix. Wnen performing dose projections several hours or more after the reactor has tripped, these two decay corrections can significantly alter the resultant projections. The computer operator is given the option of whether or not to account for decay between reactor trip and dose projection.
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TMI-l number Radiological Controls Department 9100 PLN 4200.02 Revision No.
t(Wv [ TMI Emergency Dose Calculation Manual (EDCM)
Title 8
5.8.5 For dose projections based upon RMS readings,' the: decay correction to'the noble gas to iodine ratio is straightforward.
A default mixture of the fifteen noble gases and iodines is selected based upon an assessment of core damage. If the computer operator elects to decay the mix, he is prompted for the decay time between reactor trip and dose projection. All o fifteen isotopes are decayed by the standard exponential decay law and the noble gases and todines are totaled separately so that their ratio at dose projection time can be calculated.
(Ingrowth of xenon isotopes from decay of lodine is accounted for.) The decay adjusted ratio ran then be used to fill in the missing noble gas or iodine incarnation, as explained above.
5.8.6 When iodine samples are taken at the MAP-5 stations a two step decay process is used. As above, a default mixture is chosen, based upon the NRC core damage classification. For a dose calculation based upon a radioiodine processor saraple, if de e.y correction is desired, the user is prompted for two decay time intervals:
- 1. The time between sampling and dose projection
' [O} 2. The time between reactor trip and_ dose projectis QJ 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 todine ratio is computed for the decayed default mix. This ratio, along with the gross radiciodine sample result, is used to compute a gross noble gas source term. Isotopic source terms are calculated from the decayed mixture noble gas fractions. Note that the final source term is a combination of noble gases from a cefault mix and radiciodines from a sample. Each has been decayed to the dose projection time.
A word of cautien should be added at this point. The iodine l
released in certain types of accidents may be reduced by various chemical and physical processes such as lodine plateout or formation of water soluble iodide salts. The noble gas to l iodine 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 underestimating the noble gas portion of the source term. RAC personnel should be aware of this possibility. A comparison of CN field team data and the source term dose projections would reveal agreement for thyroid doses, but not for whole body doses.
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' Radiological Controls Department 9100-PLN-4200.02'
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TMI Emercency Dose Calculation Manual (EOCM)- 8 5.9 Effluent Release Flow Rates - Flow rates for effluent releases to'the
" environment are divided into four categories:
~
7 .l. Normal ventilation flow rates.
- 2. Reactor Building leakage. flow rates.
- 3. Adjacent momentum plume rise (station vent,and reactor purge concurrently releasing).
a- ..
- 4. Flow rates-for OTSG. tube rupture release-directly to atmosphere.
. Buoyant plume rise'
- Source term calculation using RMG-26 or RMG-27
- Source term calttlation using a contingency calculation
, These flow rates for accident source ~ terms to the environment are calculated as follows:
m -5.9.1 THI-I 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-default flow rate (s). Each normal plant flow path has predetermined. flow rate ranges, and assigned flow recorders as ,
follows:
- 1. Reactor Building Purge
-FR909 0-20,000 CFM; Low' Range.
-FR148B 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
-FI284 100-1000 CFM
-Total 5-1250 CFM ,
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' . i. m Radiological Controls Department 9100-PLN-4200.02
.lj Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 E 4 .. Kidney filter System p -AHE-101, AH-F-12 P/I filters 20,200 SCFM
- 5. Auxiliary Building Exhaust
-FRISO 0-100,000 CFM
- 6. Fuel Handling Building Exhaust
-FR149 0-50,000 CFM
- 7. Auxiliary and Fuel Handling Building Exhausts
.-FR-151 0-150,000 CFM
- 8. Condenser Off-Gas Exhaust
-RMR15 Recorder FT-ill3 Ch. A 0-200 CFM
- 9. ESF Fuel Handling Building Exhaust
- bf -No flow Recorder at this time 0-8000 CFM Default values are used in the RAC Program when a small value or-an unknown value is required as input to a dose projection. The default values are:
5000 CFM - Reactor Building Purge
- Reactor Building Purge and Make-up Exhaust
- Auxiliary Building Exhaust
- Fuel 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 projections even though a value is small or unknown. Therefore, once the dose projection is complete, the results may be ratioed up or down depending on the situation. For example, if the default value of 5000 CFM was used for a Reactor Building Purge and subsequently a Tech. Functions calculation was performed 52.0 1572c l
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Nuslear TMI-I-wu e r Radiological Controls Department 9100-PLN-4200.02 f Revision No.
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-TMI Emergency Dose Calculation Manual (EDCM) 8' indicating 1000 CFM flow. -The dose projection could be ratioed down by-one fifth (1/5). Therefore, a dose projection of 10 mrem would then be approximately 2 mrem based on the reduced-
-flow calculation, realizing that the X/Q.will also be affected by reducing flow.-
5.9.2 Reactor Building Leakage Flow Rate Another'section of this program caictlates a leakage flow rate out of the Reactor Building based on Reactor Building pressure.
The Reactor Building pressure indicator-is PT-291, 0-100 psig, located 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 Reactor Building Leak Rate:
LT
= LA*SQRT(PT /PA )
where: Li - - Reactor Building Leak Rate in CFM LA - Maximum allowable integrated leakage rate
'r at PA LA = 6.14 CFM PA - Peak Reactor Building internal' pressure at design basis accident, PA = 50.6 psig PT
= . Actual Reactor Building internal pressure in psig Therefore, the maximum leakage' allowed at a design basis accident pressure of 50.6 ps b is 6.14 CFM. Leak rates at 0-60 psig can be calculated from the above formula. The default value in this subroutine is 50.6 psig. A graphic representation follows in Figure 5.9-1.
5.9.3 Adjacent Momentum Plume Rise (Station Vent and Reactor Purge)
For an isolated stack, either the station vent or the Reactor Purge, the stack gas exit velocity can be calculated from the flow rate according to the following formula:
w1 I (1) wr where w = stack gas exit velocity V = flow rate or volume flux r - radius of stack Oj 53.0 1572c
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, . 1 Nuclear. ,,_,
numeer rN Radiological Controls Department 9100-PLN 4200.02
( ).. Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 a
The Station Vent and the Reactor Purge stack are situated close.
enough together that their plumes will mix as the plumes rise.
For two or more adjacent stacks that have different exit velocities, the effect of mixing on the exit velocities of non-buoyant plumes can be given by the following formula:
w E* (2)
IV Where U - exit velocity due to mixing If these adjacent stacks were modeled as a single stack, the radius of the stack would be given by:
r- U (3) tr w At TMI-1. the reactor building stack and station vent are adjacent stacks. For computing plume rise, the stack gas exit l' velocity and stack radius we e calculated acccrding to Eq. (2)
( '~
and (3) above. A comparison of the adjacent plume rise with the plume rise from the individual stacks is shown in Table 5.9-1 5.9.4 Flow Rate Calculations for OTSG Tube Rupture Release Directly to the Atmosphere (see Figure 5.9-2 and Figure 5.9-3).
THI-l has 22 main steam relief and atmospheric 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. The valve position openings along with the secondary system pressure relate to a release flow and plume height. The percent open for these two (2) valves can be read at the center control panel under the turbine bypass dump controller for MS-V-4A/B from 0 - 100%.
Each of the 22 valves at TMI-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
,o valve opens, the flow through it will be approximately equal to 54.0 1572c
w Nuclear 1,,_,
Nucer i
-Radiological' Controls Department 9100_Pl.N ?200.02
' Revision NC.
(M.
v Title TMI Emeroency Dose Calculation Manual (EDCM) 8 the rated flow, and the flow can be assumed to be approximately constant until the pressure in the steam line drops to the point where the valve reseats. For a stuck open valve, the pressure decreases rapidly with time, and the flow through the valve is only a small fraction of the rated flow. For either a normally ';
operating valve or stuck open valve, if the pressure and ,
temperature in the steam line 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 steam plume is initially controlled by its velocity, temperature and cross-sectional area. Depending on these variables and atmospheric conditions, the plume rise can vary from hundreds to :
thousands of feet. Plume rise is a very important factor in determining maximum ground level doses. For a PWR, plume rise can increase the effective stack height by a factor of 5 to 50.
Since maximum ground level dose is roughly proportional to the inverse square of the effective stack height, a plume rise of 200 feet, for example, gives a ground level concentration ;
p) 100 times higher than that from a plume rise of 2000 feet. i
} ' (V Modeling of plume rise begins with modeling the steam condition ,
at the valve inlet. Table 5.9-3 outlines the calculational '
steps required to compute buoyant plume rise, beginning with the ,
valve inlet. The far left side of the table identifies the area for which the calculation applies: valve inlet, top of stack, jet 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 different temperature air can cause thermal boundaries resistant to plume vertical travel. In some conditions, a buoyant plume may penetrate these layers and ;
not come down to the surface as predicted. In other cases 1 the plume may be unable to penetrate the layer and the effective stack height will be reduced to the height of the layer. This may cause ground concentrations to be higher 1 and closer to the plant than predicted. In these conditions
! (i.e., highly stable meteorology with a buoyant plume) off_
site monitoring will provide an indication of the magnitude of the effect. It may also be possible to estimate the effect through visual observation of the plume.
I (O) v -_-_---__________-----_-..-___ =-= -_-___-____--__--_-____-___-----_-___-_-
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e~n Radioloolcal Controls Department 9100 f'LN-4200.02 4
) -Title Revision No.
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TMI Emergency Dose Calculation Manual (EDCM) 8 5.9.4.2 Source Term Calculation Using RMG-26 or RMB-27 (see figure 5.9-3)
RMG-26 and RMG-27 are effective in calculating a primary to secondary release source term direct to the atmosphere when:
- 1. Atmospheric Dump Values (ADV) MSV-4A or MSV-45 are open from 0-100%, as indicated on Control Room Panel "CC", and releasing radioactive steam to the environment, 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,
- 4. Steam bypass dump to the condenser through MSV-8A/B.
When the u*.er chooses a release from an OTSG tube rupture directly to the atmosphere and is using RM-G26 or RM-G27 readings, the calculation Steam Flow Computation is used to f-~s determine a release flow rate, depending on which of the valves I i are open (Table 5.9-2). The mass flow rate from each open valve
- ( ,s/ ir added up to give a total flow rate to the environment. A source term is calculated using the flow rates in CFM and the RM-G26/27 readings converted to concentration using the monitor efficiencies to give pC1/second.
=. . ---------------------------------------------------
NOTE: Calculation of a source term using RMS (RHG-26/27) is dependent on the Atmospheric Dump Valves (ADV) status. If the ADV is open, the calculation is appropriate. If the ADV is closed but plant conditions (OTSG 1eakrate and core 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-2A/B, then the contingency calculation applies.
--_-_------=- =---_----_---------_--------------------------------
i 5.9.4.3 Source Term Calculation Using a Contingency Calculation When the user performs a Contingency Calculation due t' the lack of sample results or RM-G26/27 readings, the flow rate corresponding to the set point pressure is used, if the valve operates normally. This flow rate is given in Table 5.9-2. If, however, the valve sticks open, and the steam generator pressure
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tig ' Radiological Controls Department' 9100-PLN-4200.02 -
? Title Revision No.
g TMI Emeroency Dose Calculation' Manual (EDCM)- 8 is less than'the set point pressure,Ithen the flow rate is based'
- - on 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 modeled as a release to the'-
atmosphere.
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TMI-1 Radiological Controls Department 9100-PLN-4200.02 !
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( ,/ Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 TABLE 5.9-1 Adjacent Plume Rise at TMI-1 Reactor Bldg Stack and Station Vent Stack Actual Flow Characteristics RAC Model MIDAS Station Stack Flow Stack Plume Flow Stack Plume Stability ' Reac Bld Stack Vent Diameter Rate Diameter Rise Rate Diameter Rise Stack Diameter (m) (cfm) (m) (ft) (:fm) (m) (ft)
(cfm) (m) (cfm) 20.000 1.847 30.3 10,000 1.1 25.4 Unstabhp 10,000 1.1 10.000 1.7 10,000 1.7 16.4
/NeEEAd (C d d) 199.2 65.000 1.1 165.4 1'),000 1.1 120.000 1.7 130,000 1.827 107.0 l
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65.000 1.7 f
24.8 22.1 5Mk 10,000 1.1 10.000 1.7 20,000 1.847 10,000 10,000 1.1 1.7 16.5 (Class;F}l 1.7 130.000 1.827 85.8 65.000 1.1 75.9
< 10,000 1.1 120.000 65.000 1.7 57.0
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TMI-1 E~~'s . Radiological Controls Department 9100 PLN 4200.02
' t, ) Title Revision No.
v.-
TMI EmergencV Dose Calculation _.",anual (EDCM) 8 TABLE 5.9-2 TMI-1 STEAM GENERATOR RELIEF VALVES Valve Stack Vane Vane Discharge Set Point Stack Number Function Manufacturer Area Pressure l Flow Rate Diameter (MS-V) (sq h) (psig) (Ibs/hr) (inches) 17A-D Relief VsNes, Bank 1 Dresser / 16 1050 792,617 10.02 Consolidated 18A-D Relief Vanes, Bank 2 Dresser / 16 1060 800.065 10.02 Consolidated 19A D Relief Vanes, Bank 3 Dresser / 16 1080 814.960 10.02 f-Consolidated
%./ 20A&D Relief Vanes, Bank 1 Dresser / 16 1050 792.617 10.02 Consolidated 20B&C Relief Vanes, Bank 4 Dresser / 16 1092.5 824269 10.02 C, consolidated 21A&B Relief Vanes. Dresser / 3.97 1040 194.B20 10.02 Small Safety Consolidated 22A Safety Rehef, Lonergan 6.38 200 70211 13.13 Emergency Feed Pump 22B Safety Vane, Emerg. Lonergan 6.38 220 76,795 13.13 (F.W.P.T. Steam inlet) 4A&B Relief Vakes (manual) Fisher Variable 1010 402,792 13.13 to Atmosphere 4
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- / T. Radiological Controls Department 9100-PLN 4200.02 3
( ,) Title Revision No. i TMI Emergency Dose Calculation Manual (EDCM) 8 1 TABLE 5.9-3 Calculational Steps for Computing Plume Rise l Source Step Quantity Computed input Values Needed of input Valve inlet 1 Steam flow rate thru valve pressure of steam operator Top 2 Pressure of steam at top a. steem flow rate step 1 of of stack, below chamfer b. intamal radius of stack constant Stack. (if choked flow) c. enthalpy <ccatant>
Below Chamfor 3 specific volume of steam at pressure at top of stack step 2 top of stack, below chamfer I Below 4 Temperature of ste2m at pressure at top of stack step 2
~ Chamfer top of stack, below chamfer Also 5 Velocity of steam at top of a. specmc voiume of steam step 3 Jet stack, below chamfer t. flow rate of steam step 1 Origin c. intemal radius of ctack constant 6 Density of steam at a. temperature of steam step 4 jet orirgin (ambient pressure) b. pressure of ambient ak < constant >
Jet 7 ' Jet radius at ongin a. density of steam step 6 Origin (Needed for MIDAS ony; b. velocity of steam step 5 used in RAC but not c. flow rate of steam step 1 really needed) 8 Plume rise a. jet radius at origin step 7 Plume b. density of steam at origin step 6 i
l
- c. velocity of steam at ongin step 5 l
Rise d. density of ambient air < constant >
l-
' e. wind speed at origin operator
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8 f FIGURE 5.9-1
, TMI-1 RB LEAK RATE VS RB PRESSURE ,
~
PSG 8 I I I I
la AX Al DE BLE LEAK RAfE M Ph=6.1m 50 ik=CFM e
1 E Ps=PEAX RB IN1rERNAL PRESSURE $50Apsig) # DESDI tJ BASIS AE! DENT
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Number r"~'s Radiological Controls Department 9100-PLN-4200.02 i ) Title Revision No.
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TMI Emercency Dose Calculation Manual (EDCM) 8 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 33-2 (Recommendation #12). INPO SOER 83-2 " Steam Generator Tube Ruptures" was developed based upon the steam generator tube rupture events at R. E. Ginna, Oconee and Rancho Seco. Recommendation #12 states
" Emergency Plan Implementation Procedures should . . . ensure that estimates of doses can be made for two-phase or liquid releases through the steam generator safety relief valves." GPUN Corporation is required to respond to all SOER recommendations. The calculational method used to implement this recommendation is based upon the assumptions that the valve inlet fluid condition is either pure liquid or steam (as indicated by the OTSG wide range level instrumentation) and following discharge, the steam fraction is described by assuming that there is no change in total energy content. If the 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, SPGA PTl and 2 or SPGB PT 1 and 2.
5.10.1 The program determines a two-phase correction factor [Tfcf]
f which is a function of OTSG pressure in psia. Thie factor is f s\ only calculated if the OTSG water level is indicating a liquid
\sl ~ release (greater than 600 inches on the wide range level instrument reading). The correction factors are used to account for the radiciodine that would remain in the liquid portion of the resultant two-phase release to the environment.
Upon input of the OTSG pressure in Psia the code selects a correction factor which is subsequently used in the radiciodine source term equation to correct the radiolodine source term.
RIST - D1
- D2
- D3
- RI/100
- 63.09
- 1/Tfcf
. ..- ------------------= - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
NOTE: Increasing OTSG A/B water level will possibly help cut down the release of radiolodine due to the partitioning effect of the iodine in water. Increasing OTSG level should be discussed with the Emergency Director as a means of reducing offsite doses.
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Number-
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J TMI Emeroency Dose Calculation Manual (EDCM) 8 r
5.11 Source. Term Filtration - The TMI-1 RAC Program provides the option to include or disregard reduction of source term through filtration. The
- sc'urce term reduction is' applied to radiolodine species only.
5,1141 Radiolodine normally exists'in chemical forms which are highly y
-reactive. They readily adsorb onto surfaces and can be scrubbed chemically from the atmosphere. The radioic, dine removal methods available in the plant by design are the charcoal filter banks in the ventilation system. .As applied in the RAC tape, anytime a sample is obtained upstream of a charcoal filter a filter-reduction can be applied. Anytime a default source term is used, a filter reduction and/or building spray reduction can be~
applied.
5.11.2 In any case where the RAC program will apply the source. term reductions a prompt is provided by the. program.. In a general form, the program will prompt "ARE THE CHARCOAL FILTERS.
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 radiolodine species. However, the RAC p*ogram takes a g
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 nodel. The-filters are either functioning at the full. capacity - 90% or from 0 - 90% due to degradation, 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 iind contingency default source terms generated from a spent fuel accident or LOCA in the Reactor Building, including RM-L1 readings, RCS sample results, cladding damage, or fuel melting scenarios all provide the opportunity to apply the filtration to the iodine source term. Normally, application of the filter fraction will simply reduce the radiotodine source term by a factor of ten.
However, if the source term is generated from a gas channel reading, the iodine will be reduced but the noble gas will be increased due to a constant meter reading and a change in the isotopic ratios. Application of the filter reduction for todines in the Aux /FHB is available for source terms generated from RMA-4, RMA-6, and the contingency default source terms from a sample result, damaged fuel rods, a fuel cask, or a waste gas release. It is important to remember that filter application can significantly change the noble gas to iodine ratios.
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MUClear TMI-l Number f
~
's Radiological Controls Department 9100-PLN-4200.C2
( Title Revision No.
)
TMI Emergency Dose Calculation Manual (EDCM) 8 Anytime a filter correction factor is applied to the iodines downstream of the filters, the noble gas source term must be increased, while iodines obtained upstream of a filter will gduce the iodine source term without changing the noble gases.
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"u=ber f~N Radiological Controls Department 9:00-PLN 4200.02
- i Title Revision No.
TMI Emercency Dose Calculation Manual (EDCM) _
8 5.12 Meteorology Inquiry - Upon initiation of the meteorological 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 end requests from it the most recent 15 minute average of the met data stord in the met tower PC.
The RAC program is abid to continue if the telephor:e call is not completed, the met tower phone is busy, or the met tower fails to respond after the call has begun. If any of the above conditions occur, the data is marked as 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 format and asking tne operator for the wind speed. If the "A" sensor value is non-missing, the operator is allowed to default to it. If the "A" sensor value'is missing and the."B" sensor value is non-missing, the operator is allowed to default to the "B" sensor value. If both "A" and "B" sensor values are missing, no default value is presented or allowed. To obtain the default value the operator presses the Return key while the cursor is on the first character of the input field. The operator is free to enter his own value.
After_the Return key is pressed, the RAC program subjects the inputted value to limit checking. The limits are:
Lower Limit Upper Limit
(% Wind Speed 0.5 99 l Wind 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 input a new value. An exception is the wind speed value. If the entered value is less than 0.5 mph, the RAC program uses 0.5 mph and informs the user of this fact.
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au=* r (N Radiological' Controls Department 9100-PLN-4200.02
. Title Revision No.
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TMI Emergency Dose Calculation Manual (EDCM) 8 5.13 Dispersion Model - The TMI-1 RAC model computes'both Whole Body Dose using a' 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.
5 13.1 Finib: (Whole Rody) Dose Model b
The TMI-1 RAC i.cdel calculates external whole body gamma dose rate using a finite model for both ground and elevated releases. The finite gamma dose algt,rithm is licensed from Dr. John Hamawi 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 finite gamma dose algorithm in the THI-1 RAC model has the same structare as Plckard, Lowe & Garrick's MIDAS finite gamma dose algorithm. The basis for the algorithm is a four dimensional array of finite gamma factors. These finite gamma factors are pre-computed three dimensional numerical integrations which appear in the theory of the finite cloud O model and represent the spatial distribution of the radioactive material in the finite plume. These factors depend upon the plume dimensions at the downwind disthnce 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 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 distancec, 6 heights above ground, and 6 energy groups. Specifically, the downwind distances are: 400, 500, 600, 700, 800, 900, 1000, 1250, 1500, 1750, 2000, 2250, 2500, 3000, 3500, 4000, 4500, 5000, 5500, 6000, 6500, 7000, 7500, 8000, 9000, 10000, 15000, and 20000 meters. The 6 crosswind distances are: 0, 50, 100, 150, 250 and 500 meters. The 6 heights above ground are: 0, 30, 60, 100, 150, and 300 meters.
The 6 energy groups are: .032, .081, .15, .25, .53, and 1.0 MeV. The abundances of the noble gases for the six energy groups were taken from MIDAS.
For cffective release heights other than the 6 fixed heights, the finite gamma factors are extrapolated to that height by the subroutine (Interpolate). For downwind distances other than the 28 fixed downwind distances, the finite gamma factor of the nearest fixed distance is assigned to that distance, i.e., no horizontal interpolation is done, as is consistent with MIDAS.
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L NuclearL 1s1.,
- be r i Radiological Cont rols Department 9100 Pt.N 4200.02 l :
{ tTitle Revision No.
L TMI Emergency Dose Calculation Manual (EDCM)-
8 The TMI-l'RAC model explicitly includes the contribution of I-131, I-132, 1-133, 1-134, and I-135 to the. external whole body.
l gamma dose. This method of handling the contribution from the j, .tadiolodines is more accurate than the method used in MIDAS.
The abundances of the radiolodines were taken from the Radioactive Decay Data Tables. D.C. Kocher, 1981-. All radionuclides are decayed during plume travel..
5.13.2 Semi-Infinite Dose Model The TMI-1 RAC model calculates the thyroid dose rate due to inhalation of I-131, :-132, 1-133, 1-134, 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 dost factors. The program uses the child breathing rate of 0.42 m3/hr (from Table E-5, Reg. Guide 1.109) and the child inhalation dose factors are from Table E-9, U Reg. Guide 1.109 to compute the dose rate conversion factors,
! The dose is computed by multiplying the dose rate by the expected duration of release.
The radiolodines are decayed during plume travel time. The decay constcnts for I-131 through I-135 are from the Radiological Health Handbook.
5.13.2.1 X/Q Calculation:
l The basis for the X/O calculation is the Gaussian diffusion equation and a 10 x 7 array of sigma y's and sigma z's. The array of values correspond to sigma y's and sigma z's for 7 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 are: 200, 500, 1000, 2000, 3000, 6000, 10000, 30000, 50000, and 80000 meters.
5.13.2.2 Compute Building Effect Returns one of seven pre-computed virtual source distances, depending on stability class. The virtual 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 nearest large building. Building wake effects are simulated by adding the virtual source distance, for a particular stability class, to the actual downwind distance for the purpose of computing X/Q. For example, suppose we wanted to know X/O without building wake effects at 800 meters downwind with O
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L/ "'s Radiological Controls Department 9100 PLN 4200.02 i I Title Revision No.
r)
'" TMI Emergency Dose Calculation Manual (EDCM) ,- 8 X/0 would then be' computed at 800 meters
_ stability class D.
L 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/O at'a distance greater than the actual downwind distance and is called only for ground level portion of; release.
l ,
5.13.3 Subroutines Used by Both Finite and Semi-infinite Models:
[ 5.13.3.1 Compute THI-1 Emergency Action Level. Declares the emergency' L
action level from highest dose whether whole body or thyroid.
Emer. Action Level Maximum Dose Rate (mrem /hr) within 10 miles WHOLE BODY THYROID None 0 1 dose rate < 10 and 0 1 dose rate < 50 Alert 10 1 dose rate < 50 or 50 1 dose rate < 250 Site Area Emergency 50 1 dose rate < 1000 or 250 1 dose rate < 5000 General Emergency dose rate 2 1000 dose rate 1 5000
(
The subroutine computes the EAL for whole body and thyroid dose and then reports the more severe of the two.
5.13.3.2 Compute Site Boundary The whole body and thyroid doses are computed at the site boundary. The distance to the site boundary varies with the compass sector that the wind is blowing to. This routine returns this distance in meters.
5.13.3.3 Compute Terrain factor-Computes terrain height in meters for a given downwind distance. At downwind distantes other than those in the subroutine, terrain height is computed by linear interpolation, except at distances closer than 610 meters. Between the plant and 610 meters downwind, the terrain height is set equal to the terrain height at 610 meters. Terrain further from the plant is never lower than terrain closer to the plant due to mathematical approximations.
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o Nuclear TMI-l numeer TN Radiological Controls Department. 9100-PLN-4200.02
Title:
Revision No. .;
[
TMI Emergency Dose Calculation Manual-(EDCM) 8 5.13.3.4 Compute Stability Class. l
}
As measured by the TMI. Meteorological Tower from the 150 ft !
minus 33 ft temperature difference. Table 5.13-1 relates the temperature difference to the stability class. The equivalent l temperature difference per 100 f t is shown in the last column of. !
the table. With regard to the temperature difference,-MIDAS can i be confusing because MIDAS expects the input in degrees per 117 )
feet, but prints it out in degrees per 100 f t. Stability class is determined by the measured temperature lapse' rate per Reg. !
. Guide 1.21.
5.13.3.5 Adjust _ Wind Speed Adjusts wind speed from the anemometer height to the release height. The wind speed is adjusted according to the following equation:
U = uo(h/ho)p where the subscript "0" denotes the anemometer height and "u" and "h" are the wind speed and height above ground,
. n)
?
respectively. The exponent p is a function of stability: 0.25, 0.33 and 0.50 for unstable, n9utral and stable cases,
' \d 1 respectively. If the adjusted wind speed is less than 0.5 mph, j the adjusted wind speed is set equal to 0.5 mph. (
I 5.13.3.6 Compute Exit Velocity )
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 Computes the plume rise in meters for the elevated portion of a i
spilt wake release. Two formulas-are used to calculate the
- 1. plume rise; for unstable and neutral conditions jet plume rise, momentum dominated, is calculated from Briggs' Plume Rise Eq.
l 4.33; for stable stability, it is calculated using Eq. 4.28 from l Briggs' Plume Rise.
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Nuclear 1,1_,
Number l Radiological Contro's Department 9100-PLN 4200.02
) Revision No.
/ Title s_.-
TMI Emeroency Dose Calculation Manual (EDCM) 8 5.13.3.8 Compute Entrainment Factor l Computes entrainment factor for wake split flows. A mixed mode release is assumed when: (1) the release point is at the level j of or above adjacent solid structures but lower than elevated release points, (2) the ratio of plume exit velocity to horizontal wind speed is between one and five. Specifically, j the entrainment factor, E t, is computed according to the following formulas:
Et - 1.0 for wo/u le 1 Et - 2.58 - 1.58(wo/u) for 1 gt wo/u le 1.5 i 1
Et - 0.30 - 0.06(wo/u) for 1.5 gt wo/u le 5.0 Et - O for wo/u gt 5.0 where wo is the stack gas exit velocity and u is the wind speed at stack height in miles per hour.
("'s, Note that the entrainment factor does not address the case of ;
two adjacent plumes mixing with each other, as would be the case
!,v) in TMI-1, where it is possible for a cle.in plume and a contaminated plume to be emitted from adjacent but separate ;
stacks. These plumes are examples of co-located adjacent jets- I little is known about the modeling of co-located adjacent jets such as the ones at THI-1.
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=m y Radiological Controls Department 9100-PLN-4200.02 Revision No.
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, .. W , TMI Emergency Dose Calculation Manual (EDCM) 8-F w ,-
TABLE 5.13-1
, Classification ~of Atmospheric Stabuity-
- g. Stability Pasquill. LDelta'T Delta T l;l Classification-- Categories (150'~- 33'). (*F/100')
P (*F) h '(*F)
.Extremel'y Unstable AL < -1.22 < -l'04 .
Moderately Unstable- B- 2 -1.22 to ( -1.09 2 -1.04 to < -0.93
.Slightly Unstable. .C 2, 't;09. n < -0.96 2 -0.93 to.< -0.82
' Neutral
'D 2 4 .96 to=c -0.32 2_-0.82 to.-0.27.
51ightly.. Stable E_ 2 -0.32-to < +0.96 'l -0.27 to-< +0.82-s Moderately' Stable F. 2 +0.96 to < +2.56 .
2 +0.82 to < +2.19
^
K . Extremely Stable G > +2.56 > +2.19
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i Nuclear THI-l number
/~ Y Radiological Controls Department 9100-9LN 4200.02 1
( Title Revision No. -I TMI Emergency Dose Calculation Manual (EDCM) 8 i 5.14 Offsite Air Sample Analysis 5.14.1- Introduction
-The "Offsite Air Sample Analysis" portion of the RAC code is used in conjunction with results provided from field teams to assess thyroid dose commitment.
The method involves collection of an air. sample using a low flow
-(about 50. LPM) sampler with both a particulate filter.and an iodine adsorber cartridge. The flow rate of the sampler, the duration of sample collection, the background of the frisker used to count the sample, the gross counts on the particulate filter, and the gross counts on the lodine cartridge are called into the RAC or EACC from the field teams. The RAC or EACC staff then uses the RAC code'to estimate the offsite dose commitment based on the sample.
5.14.2 Assumptions A calibrated face loaded iodine cartridge was obtained and was
[( used to determine the actual efficiency of a Eberline E140N with.
a HP-210/260 type probe to be used for counting in the field.
The results of several tests on combinations of different probes and ratemeters showed a consistent 0.0039 (0.39%) counting efficiency. (Reference 7.7, 7.10, 7.11). Since I-131 has a fairly strong beta (0.6 MeV max.), the usual particulate filter counting efficiency of 0.1 (10%) is used. The collection efficiency for both filters for these calculations are assumed to be 1.0.
5.14.3 Calculation The method first calculates the net 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.
Since the plant RAC code normally accounts for five different iodine isotopes, the dose conversion factor (DCF) used is a 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.
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Nuclear THI-I Number 7
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Radiological Controls Department' 9100-PLN-4200.02 Revision No.
(J)
Title TMI Emergency Dose Calculation Manual (EDCM) 8 5.14.4 Example Given an offsite air sample was taken with the following results:
Background - 100 cpm gross cartridge countrate - 200 cpm gross particulate countrate - 200 cpm flow rate through sampler - 50' LPM sample duration - 10 min. '
exposure duration - I hour DRCF = 4.0E8 mrem /hr uCi/cc ,
The RAC program follows the logic below to calculate an off-site thyroid' dose commitment for this sample.
- a. net particulate countrate - 200 - 100 - 100 cpm 4
. b. net particulate activity - 100/.1 - 1000 dpm
- c. net cartridge count rate - 200 - 100 - 100 cpm
) d. net cartridge activity - 100/0.0039 - 25600 dpm O .
- e. total activity in sample - 1000 + 25600 - 26600 dpm
- f. total microcuries - 26600/2.22E6 - 0.012 uCi
- g. sample volume - 50
- 1000
- 10 - SES cc
- h. air concentration - 0.012/5E6 - 2.4E-8 UCi/cc 1
- 1. dose commitment - 2.4E-8uC1/cc*1 hour *4E8 mrem /hr - 9.6 mrem i uCi/cc I 5.14.5 The items listed below appear while performing this section of the RAC program. Once all input is entered, the resultant thyroid dose commitment is displayed in mrem, or mrem /hr if a q duration of one (1) hour is entered. l 0FFSITE AIR SAMPLE ANALYSIS EXAMPLE SAMPLE TIME (military clock)
SAMPLE LOCATION FIELD TEAM DESIGNATION f'5 75.0 1572c l l
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Nuilear . .
-TMI-1 Number 95 Radio'ogical Controls Department 9100-PLN 4200.02
' Title. >
-Revision No.
'TMI Emeraency Dose Calculation Manual (EDCM)-
8-l-
BACKGROUND'COUNTRATE-100 Cpm GROSS PARTICULATE COUNTRATE 200 Cpm-e GROSS CHARCOAL COUNTRATE'200 Cpm.
FLONRATE THROUGH SAMPLER 50 LPM DURATION OF SAMPLE COLLECTION.10 MIN
' EXPECTED DURATION OF RELEASE 1 HOURS For Dose Rate Enter Duration of One Hour THYROID DOSE COMMITMENT 9.66E+00 MREM (using weighted' child DRCF)
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Number gy Radiological Controls Department 9100-PLN-4200.02 Title Revision No.
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.TMI Emergency Dose Calculation Manual (EDCM) 8 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, 5.13-2) MPC's in the river, travel time to downstream users, and ingestion dose commitment calculations. The methods used to perform the calculations are as follows:
- 1. The concentrations of the liquid effluents are determined by one of the following methods. Each method uses only the four usual iodine
. isotopes and Cs-134 and Cs-137. All selections are menu driven,
- a. Normal Miscellaneous Liquids. An ' average' isotopic content for miscellaneous plant liquid wastes is called if this option is selected. The isotopic concentrations used are typical values for default use only. The values cannot be changed. If other isotopics are known to be present, other options should be used. The discharge rate is used with the concentrations to calculate the source term.
- b. Known Isotopic Concentrations. If the concentrations of the five iodines 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 The discharge rate is prompted for in order
- (m)
(/
for each isotope.
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 liquids are released to the river, this option is 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. The isotope percentages for the five iodines are normalized to a sum of 1 and the cesium activities are calculated using ratios for the Cs-134 and Cs-137 to the I-131 based on the NRC damage class. The ratios are 0.0075 for Cs-137 and 0.035 for Cs-134.
These are 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 .5, .1 or 0.
The leakrate and duration of primary to secondary is used to estimate the maximum activity in the secondary system. The discharge rate of secondary to the river is used to estimate the source term.
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Nuclear THI-l Numeer fi Radioloolcal Controls Department 9100-PLN-4200.02 Title Revision No,
'(w [. TMI Emeroency Dose Calculation Manual (EDCM) 8 1'
- d. RCS Direct to River. The damage class and isotopic percentages are used in a manner similar tc, that for primary to secondary leakage. The isotopic concentrations and the discharge rate are used to calculate the source term.
- e. The activity.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 RML 10. The response of RML 10 in cpm / pct /ml is used to estimate the total concentration and the isotopic percentages are used to calculate individual concentrations. The discharge rate from the turbine building sump is then used to calculate the source term.
- 2. The dilution in the river is calculated by first obtaining the river flow rate and~1nputting the value in the program. Instructions are provided for obtaining the flow rate. The river flow rate.is then used along with the discharge flow rate-to calculate the concentration in the river. The concentration in the river.is then divided by the water MPC to determine the MPCs in the river to downstream users.
- 3. The river concentration is used along with the total discharge time, to calculate the dose commitment to an' individual from drinking the V(n( river water from one of the downstream intakes. The river concentration is multiplied by the duration of the release, the usage factor, the ingestion dose commitment factor for infants, and the infant usage factor (3301/yr) to obtain an estimated dose 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 all seven isotopes.
- 4. A fiume arrival time is estimated for each known downstream user and printed out. The river volume flow is used in an algorithm based on a model derived from river dye dilution and flow studies conducted from the TMI discharge.
- 5. If the concentration of any nuclide exceeds le-6 in the river, 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per 10 CFR 20.403. If the MPCs exceed 5000, immediate notification of the NRC is required.
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TMI Emergency Dose Calculation Manual (EDCM) 8 !
TABLE 5.15-1 TMI-2 Sump Capacity Sump Total Capacity Gallons / Inch Turbine Bldg. Sump 1346 gals 22.43 Circulating Water Pump House 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 gals 14.29 Chlorinator House Sump ---- -----
Water Treatment Sump 1615 gals 22.43
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Air Intake Tunnel Normal Sump 700 gals -----
Emergency Sump 100,000 gals 766.00 Condensate Polisher Sump 2617 gals 62.31 Sludge Collection Sump 1106 gals 26.33 Heater Drain Sump ---- -----
Solid Waste Staging Facility Sump 1476 gals 24.
Aux. Bldg. Sump 10,102 gals ~ 202 Decay Heat Vault Sump 478.5 gals or 957 gals (total) ~ 10 Building Spray Vault Sump 478.5 gals or 957 gals (total) ~ 10 4 1 w
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T Radiological Controls Department 9100-PLN 4200.02 i / Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 TABLE 5.15-2 TMI-l Sump / Tank Capacities Sump Capacity (Gallons)
Turbine Building Sump (TBS) 10,000 Auxiliary Building Sump (ABS) 10,000 Reactor Building Sump (RBS) 10,000 Intermediate Building Sump West 1,000 Tendon Access Gallery Sump 1,000 Intermediate Building Sump East 1,000 Auxiliary Boiler Sump 2,000 Powdex Sump 40,000
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300,000-() Industrial Waste Treatment System Sump (INTS)
Industrial Waste Filtration System Sump (IWFS) 80,000 Tanks TMI-2 Condensate B Tank 250,000 TMI-l OTSG A or B (secondary) 25,000 TMI-l HECST A or B 8,000 Neutralizer Tank 100,000 BWST 350,000 Condensate Tank A/B 265,000 m
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\g Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 5.16 Protective Action Recommendation Logic ,The Logic Diagram 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 TMI-1 and Figure.5.16-2 for THI-2.
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8 TIME M IEAND (MIT 1 tainC EsamtAN FIGURE 5.16-1 KVELDPMDfr F PMNECTIVE ACTEM aremmeATEINS CPAft)
SITE AfEA DOIIENCY IS MCUN3 j EMBAL DERENCY E KCLAKD
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TMI Emergency Dose Calculation Manual (EDCM) 8' FIGURE 5.16-1 (Cont'd)
PROTECTIVE ACTION RECOMMENDATIONS (PAR) LOGIC DIAGRAM
--- --. .=--------------- __ :_------- _ = - - - - - = - - - - - - -
NOTE 1: AS INDICATED BY ONE OF THE FOLLOHING:
- 1. CAT-PAS SAMPLE RESULTS OF:
TOTAL NOBLE GAS CONCENTRATION > 2300 MICR0 CI/CC TOTAL RADIOI0 DINE CONCENTRATION > 420 MICR0 CI/CC
- 2. HIGH RANGE CONTAINMENT AREA MONITOR READINGS OF:
RMG-22 OR RMG-23 > 400 R/HR (HIGH ALARM)
==__-__ --- -=.-------- _ ------==- -- ----------_== -____-- = ---
NOTE 2 kh5hbibkTEDBkONEbFTHEibLLbhkNG
- 1. RCS POST ACCIDENT SAMPLE ANALYSIS INDICATES GREATER THAN NRC DAMAGE CODE 2.
- 2. RAC SOFTWARE CODE CALCULATES GREATER THAN NRC DAMAGE CODE 2. (THIS IS BASED ON RCS PRESSURE AND THE AVERAGE
,s OF THE FIVE HIGHEST INCORE THERMOCOUPLE - POINT C4006 ON COMPUTER)
(
(s_, 3. LE1DOWN HONITOR READINGS RML-1 LOW AND RML-1 HIGH ARE OFFSCALE HIGH (THE ISOLATION INTERLOCKS MUST BE BYPASSED TO GET THIS READING).
_ ---------_------------- -- =------
NbEh khkNbkbkkbbYbhEbF HEhbLLbhkNb 1 - REACTOR BUILDING PRESSURE > 30 PSIG 2 - REACTOR BUILDING HYDROGEN CONCENTRATION > 3% BY VOLUME 3 - SIGNIFICANT OTSG LEAKAGE INVOLVING MULTIPLE TUBE FAILURES.
4 - A DIRECT REACTOR BUILDING TO ATMOSPHERE PELEASE PATH-HAY SUCH AS RB PURGE VALVES FAILURE TO CLOSE.
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TMI Emergency Dose Calculation Manual (EDCM) 8~ .
FIGURE 5.16-1. (Cont'd)
PROTECTIVE ACTION RECOMMENDATIONS (PAR) LOGIC DIAGRAM
==------------------------
f NOTE 4: 'THI EVACUATION TIME ESTIMATES
' LOWER (HOURS) UPPER (HOURS)
BEST ESTIMATE (NIGHT) 5.75 9.50 TYPICAL HEEKDAY (NORMAL) 6.25 10.25 12.25
. ADVERSE kEATHER 10.00.
LONER - GOOD STATE OF EMERGENCY READINESS (SLOW SCENARIO)
UPPER - LACK OF ADEQUATE PREPARATION TIME (FAST SCENARIO)
_ =_ _-- ----------------------- =- ------
NOTE 5: CONSIDERATION SHOULD BE GIVEN TO THE PROJECTED EXPOSURE TO' BE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATES, IN SO DOING, YOU MUST FACTOR RELEASE DURATION, RELEASE g si MAGNITUDE AND ASSUME A PROTECTION FACTOR OF 2 FOR UP TO THE l FIRST 2 HOURS OF RELEASE DURATION AND A PF OF 1 FOR >
\s_ 2 HOURS RELEASE DURATION. THE PATHWAY OF LEAST EXPOSUR:
SHOULD BE CHOSEN. IF THE DOSE RATE IS 400 MREM /HR; SHELTERING FOR 3 HOURS H0JLD RESULT IN AN EXPOSURE OF 800 MREM.
==- _==_------_-----_------------------------------_------------------------
-- ------------- ==---- -__ .-------------------------------..------------
NOTE 6: PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICA-TION OF THE KEYHOLE CONCEPT.
(CONSIDER USING 2 MILE RADIUS AND 10 MILE DOWNHIND).
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%J' TMI Emergency Dose Calculation Manual (EDCM) 8 t-F FIGURE 5.16-2 (Cont'd) 1HREE MILE ISLAND UNIT 2 PAR LOGIC DIAGRAM NOTES NOTE 1: AS INDICATED BY ONE OF THE FOLLOWING:
I - REACTOR BUILDING PRESSURE > 4 PSIG 2 - REACTOR BUILDING HYDROGEN CONCENTRATION > 3% BY VOLUME 3 - A DIRECT REACTOR BUILDING TO ATMOSPHERE RELEASE PATH-WAY SUCH AS RB PURGE VALVES FAILURE TO CLOSE.
4 - A DIRECT FUEL HANDLING BUILDING TO ATMOSPHERE RELEASE PATHWAY SUCH AS A FILTER TRAIN FAILURE.
=- -----= .----------- === __==_ __---------- _ ------ _---
. == ==- _ - - - - - - - - - - - - - - - - - = -
NOTE 2: TMI EVACUATION TIME ESTIMATES LOWER (HOURS) UPPER (HOURS)
BEST ESTIMATE (NIGHT) 5.75 9.50 TYPICAL WEEKDAY (NORMAL) 6.25 10.25 fg ADVERSE HEATHER 10.00 12.25 s_- LOWER - GOOD STATE OF EMERGENCY READINESS (SLOW SCENARIO) j UPPER - LACK OF ADEQUATE PREPARATION TIME (FAST SCENARIO) !
--_----_-------------- _=-__ _-------_--------_----------------------------_ .;
--_ - --==--------------------. - - -------------------------
NOTE 3: CONSIDERATION SHOULD BE GIVEN TO THE PROJECTED EXPOSURE TO BE RECEIVED TO A PERSON IF HE SHELTERS VICE EVACUATES, IN SO DOING, YOU MUST FACTOR RELEASE DURATION, RELEASE MAGNITUDE j 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 i RELEASE DURATION. THE PATHWAY OF LEAST EXPOSURE SHOULD BE CHOSEN. IF THE DOSE RATE IS 400 MREM /HR; SHELTERING FOR 3 HOURS WOULD RESULT IN AN EXPOSURE OF 800 MREM. !
_ _=_ _ ---------------------------------------------
l NOTE 4: PROTECTIVE ACTION RECOMMENDATIONS SHOULD INVOLVE APPLICATION 0F THE KEYHOLE CONCEPT (CONSIDER USING 2 MILE RADIUS AND 10 MILES DOWNWIND)
- -=_ _ - - - . -- -------------- ___------------------------------------ ;
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Radiological Controls Department 9100-PLN-4200.02
/ litle Revision No.
.s TMI Emergency Dose Calculation Manual (EDCM) 8 5.17 Dose Projection Model Overview TMI-l 5.17.1 The dose projection model may be regarded as three distinct sections. Certain variables are passad between those sections:
- a. source term generation
- b. met data input
- c. dose calculation model Indeed, the operstor 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 i source term parameters. The program retains the most recent set of source term and met parameters.
5.17.2 Source term parameters that are used in the dose projection portion of the model are:
p-An array of the 15 isotopes of interest in pCi/sec.
(v) 1.
The release flow rate in cfm.
- 2. ;
- 3. The release point stack height of 48.6 meters.
4 The release point stack diameter in meters - 1.7 meters for the Station Vent and 1.1 meters for the Reactor Building.
- 5. A single letter code for the type of release.
- a. G - Ground level
- b. 5 - Split wake
- c. E - Elevated
- 6. A character string describing the source, i.e., RMA-6 300 cpm.
- 7. The date and time of the source term parameter entry.
The met conditions section requires as input the type of I release, i.e., G, S, or E. If this variable is not present the operator is required to specify it before the program
,S continues.
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?^ Radiological Controls Department 9100-PLN-4200.02
- ) Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 5.18 TMI-2 Source Term Calculation
_ _ . = - - - -__. ---------------------------- _ _---.--------------
NOTO Follow Figure 5.18-1 THI-2 RAC Program Flowchart following this section.
_ _-_ __ _ .=---- ..-----------
5.18.1 Source Term Calculations - The source term portion of the THI-2 dose assessment program is used to generate the quantity and radionuclides 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 used to calculate radionuclides release factors and assess the accident consequences. These assessments are documented in Technical Evaluation Reports (TER's) or Safety Evaluation Reports (SER's). Source Term Calculations are performed by three methods, once the release pathway is chosen. These methods are
- 1) using Radiation Monitoring System readings, 2) Actual sample results, or 3) Contingency calculations.
l l 5.18.2 The following facilities are considered as being radioactive material release pathways for the TMI-2 RAC program.
,(x. dn)
DEFAULT VENTILATION CURIES FACILITY RMS FLOW RATE (CFM) AVAILABLE 1 - WHPF EBERLINE PING 7100 100 2 - RLM EBERLINE PING 900 100 3 - CACE AMS-3 (2) 2000 100 4 - EPICOR II EBERLINE PING 9000 100 5 - B&W TRAILER AMS-3 1300 100 6 - STATION VENT HPR219 VICTOREEN PING 120K - 130K 100 7 - STATION VENT HPR219A EBERLINE PING 120K - 130K 100 8 - ISHSF/ PAINT SHED NONE NONE 100 9 - RAD. INST. SHOP NONE 4000 100
/m.
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.. 7 i ) -Title Revision No.
J THI Emergency Dose Calculation Manual (EDCM) 8 The above list shows the associated default data concerning each release point in the TMI-2 RAC program. The station vent release pathway also includes an option for a dropped fuel canister accident. Figure 5.18-2 shows the main THI-2 ventilation.
5.18.3 Radiation Monitoring System (RMS) Source Term Calculation Only the RMS channel available for a selected release pathway is offered to the user. The following parameters are used to calculate a THI-2 source term:
- 1. RMS reading (CPM, pC1)
- 3. The release flow rate (CFM)
- 4. Cs-137/Sr-90 Ratio 5.18.4 Post Accident Sample Source Term Calculation
) This option for a particular reltase pathway is to use actual lJ k sample results to develope a source term. This option would be the most preferable method for calculating a TMI-2 source term since this method eliminates built in conservatism from RMS or contingency calculations. In this option the sample results or defaults, if required, will be used in conjunction with the release flow rate (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 facility along with the associated release flow rate are used to calculate a conservative source term.
5.18.5.1 Source Term Filtration (Contingency Calc only)
During the final calculation of a source term a filtration '
efficiency will be prompted to determine the fraction of the source term that will not reach the environment due to filtration.
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Number Radiological Controls Department 9100-PLN-4200.02 9 Title TMI Emergency Dose Calculation Manual (EDCfD Revision No.
8 ]
I 5.18.6 Dose Calculation Once the source term is established for a release from a particular pathway. This section of the RAC program will proceed very similar to the TMI-l dose projection process. A meteorology option will gather meteorology data and combine it with the source term information to complete a dose projection.
5.18.7 Dose Projection Model Overview TMI-2 5.18.7.1 The dose projection model may be regarded as three distinct sections. Certain variables are passed between.those sections:
- 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 changcd and G the source term has not, may update the met conditions and then update the dose calculation without having to re-enter the source term parameters. The program retains the most recent set of source term and met parameters.
5.18.7.2 Source term parameters that are used in the dose projection portion of the model are:
- 2. The release flow rate in cfm.
- 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 parameter entry.
The met conditions section requires as input the type of release, i.e., G, S, or E. If this variable is not present the operator is required to specify it before the program G continues.
92.0 1572c
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nueer r s Radiological Controls Department 9100-PLN-4200.02
- ) Title Revision No.
TMI Emergency Dc.se Calculation Manual (EDCM) 8 5.18.7.3 Met parameters that are passed to the dose projection portion of the.model are:
- 1. Delta T in F* per 117 feet.
- 2. Wind speed in mph.
- 3. Wind direction in degrees, from.
4 The date and time that met conditions were entered.
Upon being invoked, the dose calculation segment checks to see if the 15 isotopic values are defined. If they are not,.the source term segment is invoked. After this the dose calculation segment checks to see if the wind speed value is defined. If it is not, the met condition segment is called. Only after these two conditions are satisfied is the dose model started.
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 dose projection segment is called.
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\ Title Revision No.
i TMI Emergency Dose Calculation Manual (EDCM) 8 j i
FIGURE 5.18-1 l 1
THREE MILE ISLAND UNIT 2 RAC PROGF.AM FLOWCHART
_ wu=r 2 - R.M START 3 - cAeE 4- EP'COR !!
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8 FIGURE 5.18-1 (Cont'd) l THREE MILE ISLAND UNIT 2 !
RAC PROGRAM FLOWCHART A
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TMI Emergency Dose Calculation Manual (EDCM) B FIGURE 5.18-1 (Cont'd)
THREE MILE ISLAND UNIT 2 RAC PROGRAM FLOWCHART A
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CALCULATIONS PERFORMED E
(1) SOURCE TERM (2) METEOROLOGY (3) x GROUND RELE ASE (4) EMERGENCY ACTION LEVEL (5) DOSE PROJECTION (mrem /hr)
(6) INTEGR ATED DOSE (mremi (7) RELE ASE POINT (8) RMS/S AMPLE RESULT D AT A (9) CS/SR RATIO (10) NOTE: ALPHA EFFECT ON SR SOURCE TERM.
(IF PERFORMED) l (J
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(w-) Title TMI Emergency Dose Calculation Manual (EDCM) 8 FIGURE 5.18-1 (Cont'd)
THREE MILE ISLAND UNIT 2 RAC PROGRAM FLOWCHART B X T T
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Nuclear THI-l number
//'D Radioloalcal Controls Department' 9100-PLN-4200.02 k litle Revision No.
h TMI Emeroency Dose Calculation Manual (EDCM) 8 6.0.' RESPONSIBILITIES 6.1. The RAC is responsible to ensure that dose assessments using the
. methodology.in the EDCM are performed upon implementation of the Emergency Plan.
6.2 The RASE have the responsibility to support the RAC in performance of radiological controls and dose assessment using the methodology in the
. EDCM.
6.3 The Chemistry Coordinator has the responsibility to support the RAC in the procurement and analysis of-in-plant samples required to quantify the ;
accident.
6.4 Radiological Controls has the responsibility of proper review, l documentation, and distribution of the EDCM and the RAC program software. !
Radiological Controls is responsible for ensuring that the EDCM and the ;
RAC program software are current and compatible. )
l i
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1 1
0 100.0 1572c I
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.g Nuclear- THI-I Number
/ i Radiological Controls Department 9100-PLN-4200.02
) Title- Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8
7.0 REFERENCES
7.1 American National Standard (ANS), ANSI /ANS-18.1 - 1984, Radioactive Source Term for National Operations of 1.ight Water Reactors 7.2 APS Source Term Report - Report to the American Physical Society c,f the Study Group on Radionuclides Releases From Severe Accidents at Nuclear Power Plants, February 1985 I
7.3 Dose Assessment Manual for Emergency Preparedness Coordinators, February-1986, INPO 86-008 7.4 EDCM Flowchart Block Diagram
'7.5 Efficiency Check using an Air I-131 Source Cartridge and a Ba-133 Source Cartridge, Memorandum 9502-88-0139. September 28, 1988 7.6 Emergency Dose Calculation Manual (EDCM) Source Code Listing 7.7 EPA 520/1-75-001 - Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 7.8 EPIP 6415-IMP-1300.07 - Off-site /On-site Dose Projections 7.9 Evaluation of a Front Loaded Iodine Cartridge using Various Survey Equipment, Memorandum 9100-88-0194, May 12, 1988 7.10 Field Measurements of Airborne Releases of Radioactive Material, Memorandum 9502-88-0098, May 25, 1988 7.11 FSAR, TMI-1 Chapter 11, Radioactive Waste and Radiation Protection 7.12 FSAR, TMI-1 Chapter 14 - Safety Analysis 7.13 FSAR, THI-2, Volume 10 Section 15, Accident Analysis 7.14 GPUNC Emergency Plan, IC00-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 SOER 83-2 " Steam Generator Tube Ruptures" 7.18 Introduction to Health Physics, Herman Cember, 2nd Edition, 1985 -
A U 101.0 1572c
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l
. (N + Radiological Controls Depar_tment 9100-PLN-4200.02 j ) Title Revision No.
W TMI Emeraency Dose Calculation Manual (EDCM) 8 7.19 NRC-BNL Source Term Report 7.20 NUREG-0017 Rev.1 - PWR - GALE Code: Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PHR, April 1976 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 j Island Unit 2, October 3, 1979 7.23 NUREG-0654 - Revision I - Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 7.24 NUREG-0737 - Clarification of TMI Action Plant Requirements, U.S. Nuclear l Regulatory Commission, November 1980, Gentric Letter 82-33, Supplement I 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 Severe x Nuclear Power Plant Accidents, October.1988 1 7.26 NUREG/CR-3011 - Dose Projection Considerations for Emergency Conditions at Nuclear Power Plants 7.27 N1830 - Post Accident Reactor Coolant Sampling 7.28 N1831 - Post Accident Atmospheric Sampling 7.29 N1832 - Post Accident Sample Analysis 7.30 N1833 - Post Accident Core Damage CalculaV.ons 7.31 OP1202 Abnormal Transients Rules Guides and Graphs 7.32 OP1202 RCS S 0. Heated I
7.33 OP1210 Excessive Radiation Levels 7.34 Operational Quality Assurance Plan, 1000-PLN-7200.01 7.35 Proprietary Midas User Documentation, Pickard, Lowe, and Garrick I
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. HEH.
7.38 Reg. Guide 1.21 Measuring, Evaluating, and Reporting Radioactivity in l
I O Solid Hastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants, Rev. 1 June 1974.
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cc Nuclear. THI-1 wuer f 1- Radioloalcal Contrels Department 9100 PLN-4200.02 y/ Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 8 7.39 Reg. Guide 1.109 -' Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I, October.1977, Rev. I 7.40 SER-15737-2-G07-108, Rev. 4, March 5, 1985. Safety Evaluation Report for a TMI-2 Fuel Canister Accident 7.41 SER-419628-003, Rev. 7, Instrument Calibration Facility, Feb. 12, 1988 7.42 TDR-390 - TMI-1; Primary-to-Secondary OTSG Leakage and its On-site /Off-site Radiological Impact, April 1983 7.43 TDR 405 - THI-1; Evaluation of Plant Radiation Release and its 10CFR50, Appendix I Conformance for Different Operating Conditions 7.44 TDR-431 - Method for Estimating Extent of Core Damage Under Severe Accident Conditions 7.45 TER-13587-02-G03-015, Rev. 6, January 21, 1985. Interim Solid Waste Staging Facility Technical Evaluation Report 7.46. TER-15737-2-G03-104, Rev. 5, May 8, 1985. Technical Evaluation Report for Q the CACE 7.47 .TER-15737-2-G03-107, Rev. 6, Feb. 2, 1987. Technical Evaluation Report for the WHPF.
7.48 HASH-1400 - 1975 Nuclear Safety Study WASH-1400 (also known as Rasmussen Report)
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' TMI Emergency Dose Calculation Manual (EOCM) g-8.0 EXHIBITS i None O
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