ML20204B529
| ML20204B529 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/12/1999 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| 6610-PLN-4200.0, NUDOCS 9903220150 | |
| Download: ML20204B529 (200) | |
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{{#Wiki_filter:! TMi IRMC INSTRUCTION MEMO MANUALM EM6 OSE OALCtLL AThof)Jaut;111 (NM) I DATE -/M-N CORRECT NAME AND ADDRESS IF NECESSARY Document Control Desk l Office Nuc. Reactor Regulation U.S. Nuclear Regulatory Commision RETURN TO: DEBBIE MARSHBANK Washington, DC 20555 1RMC 2 INFO ROOM 135 U2 ADMIN. BLDG. TMI PLEASE UPDATE YOUR MANUAL WITH THE ATTACHMENTS AS INSTRUCTED BELOW AND DESTROY THE SUPERSEDED PAGES/ PROCEDURE (S). ALSO PLEASE SIGN THE ACKNOWLEDGMENT AT THE BOTTOM OF THIS MEMO AND RETURN TO THE ADDRESS SHOWN ABOVE. l l DOCUMENT NUMBEll llEV laki0-PLA-% DADA 9 7 l 1 I [ ] UPDATE INDEX/ TABLE OF CONTENTS ATTACliED. n if l 1 ilEREBY ACKNOWLEGE RECEIPT OF THIS MEMO AND llAVE COMPLIED WITil THE INSTRUCTIONS. SIGNATURE AND RETUltNED MEMO REOUIRED ONI,Y IF CONTitOLLE D. (SIGNATURE) (EXTENSION) (DATE) 9903220150 PDR 990312 ADOCK 05000289' PDR._m
FOR F RMATION ONLY Humber
]/ U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 Appbcabihty/ Scope Responsible Office Effective Date Rad /Env TMI Division Health & Safety Dir. 03/12/99 This document is within QA plan scope X Yes No Safety Reviews Required X Yes No List of Effective Pages Paae Revision Paae Revision Paae Revision Paae Revision 1 9 21 9 41 9 61 9 2 9 22 r. 42 9 S2 9 3 9 23 9 43 9 63 9 4 9 24 9 44 9 64 9 5 9 25 9 45 9 65 9 6 9 26 9 46 9 66 9 7 9 27 9 47 9 67 9 8 9 28 9 48 9 68 9 9 9 29 9 49 9 69 9 10 9 30 9 50 9 70 9 11 9 31 9 51 9 71 9 12 9 32 9 52 9 72 9 13 9 33 9 53 9 73 9 O 14 15 9 9 34 35 9 54 9 74 9 ( 9 55 9 75 9 16 9 36 9 56 9 76 9 17 9 37 9 57 9 77 9 18 9 38 9 58 9 78 9 19 9 39 9 59 9 79 9 20 9 40 9 60 9 80 9
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)) U NUCLEAR TMI Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 List of Effective Pages Pace Revision Paae Revision Pace Revision Paae Revision 81 9 126 9 171 9 216 9 82 9 127 9 172 9 217 9 83 9 123 9 173 9 218 9 84 9 129 9 174 9 219 9 85 9 130 9 175 9 220 9 86 9 131 9 176 9 221 9 87 9 132 9 177 9 222 9 88 9 133 9 178 9 223 9 89 9 134 9 179 9 224 9 90 9 135 9 180 9 225 9 91 9 136 9 181 9 226 9 92 9 137 9 182 9 227 9 93 9 138 9 183 9 228 9 94 9 139 9 184 9 229 9 95 9 140 9 185 9 230 9 96 9 141 9 186 9 231 9 97 9 142 9 187 9 232 9 b 98 9 143 9 188 9 233 9 ( ,/ 99 9 144 9 189 9 234 9 100 9 145 9 190 9 235 9 101 9 146 9 191 9 236 9 102 9 147 9 192 9 237 9 103 9 148 9 193 9 238 9 104 9 149 9 194 9 105 9 150 9 195 9 106 9 151 9 196 9 107 9 152 9 197 9 108 9 153 9 198 9 109 9 154 9 199 9 110 9 155 9 200 9 111 9 156 9 201 9 112 9 157 9 202 9 113 9 158 9 203 9 114 9 159 9 204 9 115 9 160 9 205 9 116 9 161 9 206 9 117 9 162 9 207 9 118 9 163 9 208 9 119 9 164 9 209 9 120 9 165 9 210 9 121 9 166 9 211 9 122 9 167 9 212 9 123 9 168 9 213 9 124 9 169 9 214 9 125 9 170 9 215 9 v 2
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Tdle "CE N Radiological Controls Procedure 6610-PLN-4200.02 Revis'on No. TMI Emergency Dose Calculation Manual (EDCM) 9 TABLE OF CONTENTS Section Paoe 1.0 PURPOSE 4 2.0 APPLICABILITY 4 1 3.0 DEFINITIONS 4 4.0 PREREQUISITES 10 5.0 PROCEDURES 11 5.1 Source Term Calculations 12 5.2 Release Pathways and Characteristics 12 5.3 Calculation of Fuel Damage Class and Isotopic Percentages 14 5.4 Radiation Monitoring System (RMS) Source Term Calculation 26 (g) 5.5 Post Accident Samples Source Term Calculation 31 5.6 Contingency Calculations Source Term Calculation 32 5.7 Decay Scheme Calculation 37 5.8 Noble Gas to lodine Ratio Calculations 39 5.9 Effluent Release Flow Rates 42 5.10 Two-Phase Steam Flow Determination 55 5.11 lodine Reduction Factors 56 5.12 Dispersion Model 59 5.13 Liquid Release Calculation 65 5.14 Off-Site Air Sample Analysis 69 5.15 Protective Action Recommendation Logic 70 6.0 RESPONSIBILITIES 71
7.0 REFERENCES
71 8.0 EXHIBITS 73 8.1 Emergency Dose Assessment User's Manual 75 3
T' Numtttr U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 1.0 PURPOSE ) The pu'. pose of this manualis to provide a summary of the mathematics and assumptions used to perform offsite dose projections during emergency situations. This indudr-s calculating projected on-site and off-site j doses from releases of radioactive material to the environment in accident conditions upon implementation j of the Emergency Plan. As such, this document describes rnethods 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 calculations, if RMS and sample results are not available. , t ! 2.0 APPLICABILITY / SCOPE l , l l The EDCM is applicable to all qualified TMl Radiological Assessment Coordinator (RAC) personnel involved l in the projection of on-site and off-site doses during an emergency. This manual provides the calculational i methodologies used in performance of dose projections during emergency situations where radioactive l material has been or is predicted to be released to the environment. The contents of the EDCM shall be used as the basis for the computations performed in the Emergency Plan RAC computer program. 3.0 DEFINITIONS l 3.1 BUILDING WAKE EFFECTS - When an atmospheric release occurs at, near, or below the top cf a l ) building (or any structure) the dispersion of the release is affected by the wake effect of the building. l \j 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 j significantly. Building wake is accounted for as part of the split wake release modeling. 3.2 " CHI over Q" (X/Q)-is the dispersion of a gaseous release in the environment 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 tem 1 represented by O. Thus dispersian, X/Q (sec/ cubic meter) times source term, Q ( Ci/sec) yields environmental concentration X ( Ci/ cubic meter). X/Q is 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. 3.3 CONTAINMENT ATMOSPHERIC POST-ACCIDENT SAMPLING SYST"M (CATPASS)- Post accident sampling system capable of providing sample (s) following an accident condition, coincident with a blackout, with limited personnel exposure. The sampling system, located in a post-accident accessible area, provides the capability for obtaining samples of the Reactor Building atmosphere, within one hour after the decision has been made to acquire the sample (s). The samples (s) are then used for radiological and hydrogen analysis. These results will provide an indication of the extent of core damage and provide good data for the Reactor Building source term if a Reactor Building release is possible. 3.4 CONTINGENCY CALCULATION - A source term calculation performed in the absence of sufficient p) (V effluent radiation monitcring system readings or post accident sample data, it is a mathematical calculation based upon the most representative physical model of actual accident plant conditions. 4 L
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I TMI Emergancy Dose Calculation Manual (EDCM) 9 ) 1 3.5 COREi lAGE - (Note: This definition to be used in lieu of defective / failed fuel.) A set of core ! classific uuns used to address the requirements of the NRC NUREG 0737 Criterion 2(a) upon f 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, end fuel melt. ) 3.6 DEFECTIVE FUEL / FAILED FUEL - See definition of core damage. 3.7 DOSE RATE CONVERSION FACTOR (DRCF)- A parameter calculated by the methods and models of intemal dosimetry, which indicates the Committed Dose Equivalent (CDE) to an organ or Total Effective Dose Equivalent (TEDE) to the whole body per unit activity inhaled or ingested This parameteris specific to the radionuclide and the dose pathway. Dose conversion factors are commonly tabulated in units of mrem /hr per curie /m' inhaled or ingested. 3.8 ELEVATED RELEASE - An airbome effluent plume which is well above any building wake effects so as to be essentially unentrained is termed an elevated release. The source of the plume may be elevated either by virtue of the physical height of the source above the ground elevation and buildings or by a combination of the physical heignt and the jet plume rise. Semiinfinite modeling of elevated releases generally will not produce any significant ground level concentrations within the first few hundred yards of the source. Semiinfinite modeling of elevated releases generally have less dose consequence to the public due to the greater downwind distance to the ground p 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 Werature. 3.9 EMERGENCY ACTION LEVEL (EAL)- Predetermined conditions or values, including radiation dose rates; specific levels of airbome; 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. 3.10 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. 3.11 EMERGENCY DOSE CALCULATION MANUAL (EDCM)- This controlled dose calculation manual ! is the documentation describing the content and calculational methods of the TMl Radiological Assessment Coordinator (RAC) model. 3.12 EMERGENCY OPERATIOhlS FACILITY (EOF)- The Emergency Operations Facilities serve as the primary locations for management of the Corporation's overall emergency response. These facilities are equipped for and staffed by the Emergency Support Organization to coordinate emergency response with off-site support agencies and to assess the environmentalimpact of the emergency. The EOF participates in accident assessment and transmits appropriate data and recommended protective actions to Federal, State and Local agencies. 3.13 EMERGENCY PLANNING ZONE (EPZ)- There are two Emergeacy 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 given in order to assure that prompt and effective actions can be taken to protect the public and property in the event of an accident. This is called the Plume
. Exposure Pathway EPZ. The second is an area approximately 50 miles in radius around the site, (O)v for which emergency planning consideration of the ingestion exposure pathway has been given.
This is called the Ingestion Exposure Pathway EPZ. 5
r Number ! b U TMI NUCLEAA Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 3.14 ENTRAINMENT- Whea a re! case 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 l ground release. Entrainment factor is related to the building wake effect. The entrainment factor is l computed on a case by case basis and is dependent on bo'h the stack exit velocity and the wind i l speed. At low wind speeds and high exit vdocities, tuilding effects are lowest and the entrainment I factor selects for elevated release At high veind speeds and/or low exit velocities the building effect is highest and the entrainment factor results in a ground level release. Intermediate conditions cause entrainment factors which will split the release between ground and elevated. The general form for the application of the entrainment factor (Ef)is: X/Q(splitwake)=X/Q(ground)*Ef + X/Q(elevated)*(1-Ef). i As can be seen from the formula, when the entrainment factor is one, the release is entirely ground l and when the entrainment factor is zero, the release is entirely elevated. When 0 < Ef < 1 then the l release is split. ( 3.15 EPICOR ll- Radioactive Liquid Waste Processing Facility located on the east side of TMI-2. This facility is used to process radioactive liquid waste. A Victoreen radiation monitor is located on the ventilation system of this facihty. The ventilation system average flow rate is 9000 CFM. 3.16 EXCLUSION AREA (EA)- As defined in 10CFR100.3;"that area surrounding the reactor, in which V) 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 TMI-1 and TMI-2 reactor buildings. 3.17 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 l ground and will cause that maximum to occur further downwind than would a lower release height. i 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 i geyser and rises in a jet flow above the stack. The height to which the jet flow rises becomes the effective stack height. 3.18 FINITE PLUME 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 j I large compared to the distance that gamma rays can travelin 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, 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 , usefulin evaluating doses from an elevated plume or when the receptor is near the effluent source. j l This modelis used by the MIDAS computer program. i 3.19 FUEL HANDLING BUILDING ENGINEERED SAFETY FEATURE VENTILATION SYSTEM - The I Fuel Handling Building ESF Ventilation System is installed to contain, confine, control, mitigate, i l monitor and record radiation release resulting from a TMI-1 postulated spent fuel accident in the Fuel Handling Building as described in FSAR, Section 14.2.2.1. Normal operation of the Fuel Handling Building ESF Ventilation System is during TMl-1 spent fuel movements in the Fuel ! Handling Building. The system design includes adequate air filtration and exhaust capacity to O i ensure that no uncontrolled radioactive release to atmosphere occurs. The System includes effluent J 6 radiation monitoring capabihty. l 1 6
Number U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 3.20 GAUSSIAN PLUME EQUATION - An equation which takes input parameters of plume height, sigma-Y, sigma-Z, and wind speed, which explicitly 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 duplicated by the Gaussian Plume Model. 3.21 GROUND RELEASE - An airbome effluent plume which coqtacts 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/O 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. 3.22 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 airinto the Reactor Building, which allows the displaced containment atmosphere to be discharged in a controlled manner into the normal Reactor Building exhaust duct. In the flow path three release rates exist which can be additive to give flow from 5 to 1250 CFM. 3.23 INTERIM SOLID WASTE STAGING FACILITY (ISWSF)- This facility has no ventilation system or k radiation monitor, but has the potential to release radioactive material to the environment. Releases would be assessed using field team data and/or MIDAS. 3.24 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. 3.25 METEOROLOGICAL INFORMATION AND 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 and releases during emergencies. 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. 3.26 FUEL DAMAGE CLASS - A method of estimating the extent of coro 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 denved from the average of the five highest incore thermocouples and RCS pressure. The assessment is performed utilizing a matrix that consists of ten (10) possible damage categories ranging from "no damage" to " major clad damage plus fuel melting". 3.27 PARTITION FACTOR - (Condenser), see NUREG-0017 3.28 RCS 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 and analyzing a representative sample of reactor coolant, D including dissolved gases, reactor coolant bleed tank contents and reactor containment sump
) contents, within three hours after the decision to acquire the sample, without excessive operator exposure or compromise of interfacing safety-related systems.
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Number TMI ( "C" Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 3.29 PLANT SHUTDOWN RADIOlODINE SPIKING - Radiciodine spiking occurs when the reactoris shutdown, and is caused when the fuel gap activity is washed out of fuel rods with cladding defects. 3.30 PROTECTIVE ACTION GUIDE (PAG) - Projected radiological dose or dose 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 wa- d provided the reduction in individual dose expected to be achieved by carrying out the pr ott/ctive action is not 1 offset by excessive risks to individual safety in taking the protective action. The protective action I guide does not include the dose that has unavoidably occurred prior to the assessment. ! 3.31 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. l 3.32 RADIATION MONITORING SYSTEM (RMS) - The RMS detects, indicates, annunciates, and records the radiation level at selected locations inside and outside the plant to verify compliance with applicable Code of Federal Regulations (CFR) limits. The RMS consists of the following subsystems: area monitoring, atmospheric monitoring, and liquid monitoring. 3.33 RADIOlODINE PLATEOUT -lodines are very chemically reactive, being members 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 radiolodine plateout reactions are the removal ofiodine 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. 3.34 RADIOlODINE 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 emargency 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. 3.35 RADIOLOGICAL ASSESSMENT COORDINATOR (RAC)- The RAC is responsible for all on-site radiological assessment activities. Initially, the RAC is responsible for directing the on-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. 3.36 RADIOLOGICAL ASSESSMENT SUPPORT ENGINEER (RASE)-Individuals assigned to assist the RAC in performing dose calculations, source term calculations, and overall assessment and control of radiological hazards. Normally one RAC and one RASE are on duty at all times. r ) I I l 8 i
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TMI Emergency Dose Calculation Manual (EDCM) 9 3.37 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. 3.38 RELEASE DURATION - Release duration refers to the time interval during which radionuclides are l released from the nuclear facility. Releases may be monitored, unmonitored, actual, or projected. The time interval used to estimate a release of unknown duration should reflect best estimates of the plant technical staff. In the absence of other information, use eight hours as the expected release duration. For purposes of determining whether to take a protective action on the basis of projected dose from an airbome plume, the projected dose should not include the dose that has already been received prior to the time the dose projection is done. 3.39 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 (Ci/sec) or microcuries per second ( Ci/sec). 3.40 RESPIRATOR AND LAUNDRY MAINTENANCE FACILITY (RLM)- This facility is used to process clean and maintain laundry and respirators for TMI. This facility's 900 CFM ventilation system is fo) N) monitored using an AMS-3 radiation monitor. The RLM has HEPA filters installed in the ventilation system. Releases would be assessed using field team data and/or MIDAS. 3.41 RMS RESPONSE FACTOR - Parameter which is used to convert RMS monitor count rates to total microcurie per cubic centimeter of the assumed or measured radionuclide spectrum 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. 3.42 SEMI-INFINITE PLUME MODEL - Dose assessment model which is based on the assumption that the dimensions of an effluent plume are large compared to the distance that gamma rays can travel in air. If the plume dimensions are larger than the gamma ray range, then the radius of the plume might just as well be infinite since radiation emitted from beyond a certain distance will not reach the receptor. The ground is considered to be an infinitely large flat plate and the receptor is assumed to be stand'ng 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 FLUME. The noble gas DAC's were dev* ed on the basis of the sumi-infinite plume model. 3.43 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. 3.44 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, liquids, and even dose rate calculations in plant, strictly speaking " Source Term" applies only to radioactive material actually released. 9
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TMl Emergency Dose Calculation Manual (EDCM) 9 3.45 SPLIT WAKE RELEASE - Airbome 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 be 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 telease may appear to be entirely an elevated release and at very low flow rates it may appear to ne entirely a ground level release. In intermediate conditions, the model will" split" the release between g ound and elevated as appropriate, so that a release might be 25% ground and 75% elevated from the same release point. 3.46 STABILITY CLASS - Dispersion of an effluent plume in the atmosphere is a function of the amount of mixing occurring between the plume and the atmosphere around the plume. The amount of mixing is related to what is referred to as the stability of the atmosphere. Conditions which create good mixing are unstable and conditions which create poorer mixing are stable. Pasquill stability class is a breakdown of the relative atmospheric 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
-s 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 detta 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. 3.47 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. 3.48 TWO PHASE RELEASE - Liquid and steam release from the main steam safety relief valves. Following discharge to the environment the steam fraction is calculated assuming there is no change in system entropy and that the OTSG wide range levelinstrument is indicating that the valve inlet fluid condition is either pure liquid or steam (greater than 600 inches as indicated on the Wide Range Level Indicators on PCL Panel, and the pressure indicators Pl-950A and PI-952A). 3 49 WASTE HANDLING AND PACKAGING FACILITY (WHPF)- This facility is used to handle and package radioactive waste. This facility's ventilation is monitored by an AMS-3 radiation monitor, and runs at 7100 CFM. Releases would be assessed using field team data and/or MIDAS. 3.50 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. 4.0 PREREQUISITES 4.1 The following are the prerequisites for performance of TMI projected dosas using the methods in the EDCM, and the current TMl RAC model. 4.1.1 The Emergency Plan is being implemented.
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( ) 4.1.2 The RAC station is manned and functional. 10
I Number N y' NUCLEAA TMi Radiological Controls Procedure 6610-PLN-4200.02 i Tdie Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.0 PROCEDURE l This section of the EDCM is divided into the processes that are contained in the RAC model. Listed below is l a table of contents for the procedure section of the EDCM: ! 5.1 Source Term Calculations 5.2 Release Pathways and Characteristics l 5.3 Calculation of Fuel Damage Class and Isotopic Percentages 5.4 Radiation Monitoring System (RMS) Source Term Calculation 5.5 Post Accident Samples Source Term Calculation I
- 5.6 Contingency Calculations Source Term Generation l i l
j 5.7 Decay Scheme Calculation ' 5.8 Noble Gas to lodine Ratio Calculations t (v ) 5.9 Effluent Release Flow Rates 5.10 Two-Phase Steam Flow Determination j 5.11 Qource Term Filtration i l 5.12 Dispersion Model l 5.13 Liquid Release Calculation 5.14 Off-Site Air Sample Analysis 5.15 Protective Action Recommendation Logic l 5.16 Dose Projection Model Overview l Each part of this section explains what each process does and how it does it. 11
Number N TMt NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.1 Source Term Calculations - The source term portion of the TMI dose assessment modelis used to generate the quantity and radionuclide make up of the radioactive material released (or available for release) to the environment. Once the source term is measured or estimated, meteorological and dosimetry models are applied to the assessment. 5.2 Release Pathways and Characteristics 5.2.1 The following are the Release Pathways: e OTSG Tube Rupture Release Includes: via the condenser off-gas or directly to atmosphere. ' e Reactor Building Release e Station Vent Release includes: Auxiliary Building, Fuel Handling Building (FHB) and ESF FHB. p 5.2.2 The following are the Release Characteristics e OTSG Tube Rupture via condenser off-gas e OTSG Tube Rupture directly to atmosphere via the Main Steam Reliefs or Atmospheric Dump Valves e LOCA in the Reactor Building e Fuel Handling Accident in the Reactor Building e Fuel Handling Accident in the Fuel Handling Building, including ESF Fuel Handling Building Releases e LOCA in the Auxiliary Building e Waste Gas Tank Release , 5.2.3 The following methods are used for source term generation: e Use Post Accident Sample Result e Use RMS i e Use Contingency Calculation NOTE The least accurate to the most accurate source term generation methodology is contingency calculations, RMS calculations, and post accident sample result calculations. 12
l Number I / TMI ( mmR Radiological Controls Procedure 6610-PLN-4200.02 Tile Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.2,4 Airbome releases from the following pathways are evaluated (See Figure 5.9-4): A. The OTSG Tube Rupture: , O RM-A-5 Condenser Off-gas
- O RM-A-5 High-Condenser Off-gas S RM-G-25 Condenser Off-gas i e RM-G-26 Main Steam Reliefs and Atmospheric Dump Valves 9 RM-G-27 Main Steam Reliefs and Atmospheric Dump Valves :
9 RM-A-5 MAP-5 Samples - e Main Steam Release directly to the atmosphere ! O Contingency Calculations without RMS or Post Accident Samples I l B. The Reactor Building: 0 RM-A-9 Reactor Building Purge 8 RM-A-9 High-Reactor Building Purge O RM-G-24 High High-Reactor Building Purga O RM-A-2 Reactor Building Atmosphere e CATPASS Samples n e RM-A-9 MAP-5 Samples
/ \ 9 Contingency Calculations without RMS or Post Accident Samples \) C. The Station Vent:
0 RM-A-4 Fuel Handling Building Exhaust 9 RM-A-6 Auxiliary Building Exhaust l 0 RM-A-8 Station Vent (Auxiliary and Fuel Handling Buildings) ) e RM-A-8 High-Auxiliary Building Exhaust i S RM-A ESF Fuel Handling Building Exhaust l 9 RM-A-8 MAP 5 Samples l i 4 V i l l 13 I l l l
I Number [ U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.0? Tnie Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 I 5.3 Calculation of Fuel Damaae Class and Isotopic Percentaaes - This calculation will determine the mix or percentages of the following fifteen radionuclides. 5.3.1 Ten Noble Gases Five Radiciodines i
- 1. Kr-85m 1. 1-131
- 2. Kr-85 2. 1-132
- 3. Kr-87 3. 1-133
- 4. Kr-88 4. 1-134
- 5. Xe-131m 5. 1-135
- 6. Xe 133m l
- 7. Xe-133 l 8. Xe-135m O)
- 9. Xe-135
%)
- 10. Xe-138 l
l 5.3.2 The Fuel Damage Class Determination The determination of the Fuel Damage Class is performed using various core temperature l regions from Operations Procedure 1210-8, see Figure 5.3-1. The core temperatures used j l in the model are the average of the five highest incore thermocouples. The curves relating to saturation, and cladding failures are approximated by straight line equations. Fuel I damage classes 1 - 10 are based on the different pressure and temperature regions of Figure 5.3-1. 5.3.2.1 Core Temperature Regions - Figure 5.3-1 1 The region to the left of curve C represents normal RCS activity, Fael Damage i Class-1. The region between curves C and D represents RCS plus a percentage ) of gap activity, Fuel Damage Class 2 - 4. The region between Curve D and Curve i E represents RCS plus all gap activity plus a percentage of noble and volatile fission product release from fuel grain boundaries, (CS, I, Rb), Fuel Damage l l i l Class 5 - 7. The region between Curve E and 2550*F incore temperature j represents RCS activity p;us 100% of the gap activity and 100% of the in vessel 1 melt release assuming NUREG-1228 release fractions, Fuel Damage Class j 8-10. p ).
> l V ;
i 14 l l 1 i
Number /'N b U NUCL?AR Tui Radiological Controls Pmcedure 6610-PLN-4200.02 Tale Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 5.3.2.2 The curves represented in Figure 5.3-1 have the following t:quations:
. Curve C: 1400F Tclad Cuwe Temperature = 406.1 + (0.34027
- PRESS)-(0.0000538
- PR6SS2 )
. Curve D: 1800F Tclad Curve !
Temperature = 690.92 + (0.37178
- PRESS)-(0.0000554
- PRESS 2)
. CuNe E: 2200F Tdad Cume Temperature = 1135.3 + (0.40018
- PRESS)-(0.0000567
- PRESS 2) I Where: Temperature = Incore Thermocouple Temperature (F');
PRESS = RCS Indicated Pressure (Psig) 5.3.2.3 The matrix below shows the theory of fuel damage based on TDR-431. Fuel Damage Class Degree of Minor intermediate Major Degradation < 10% 10 - 50 % >50% No Fuel Damage (RCS) -No Damage : Class 1 or Class 1 A Cladding Failure (GAP) 2 3 4 Fuel Overheat (Fuel Matrix) 5 6 7 Fuel Melt (Fuel Matrix) 8 9 10 v 15
l (Qpy & l' (
\ WuCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 0 l 5.3.3 a C_al.culation of Radionuclide Mix Percentaaes Based on Fuel Damaae Classification. Once the determination of Fuel Damage Class 1 - 10 has been determined from the Core Temperature Regions the various radionuclide mix percentages csn be calculated based on the distribution of the RCS Activity, GAP Activity, and/or Fuel Mat.ix Activity. Various combinations of activities for each Fuel Damage Class are modeled as follows: l Fuel Damage RCS Activity GAP Activity Fuel 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 l 6 1 1 0.5 7 1 1 1 r] 8 1 1 1 g l 9 1 1 1 10 1 1 1 5.3.3.1 Therefore, as an example: Fuel Damage Class-6 would consist of 100% RCS Activity, plus 100% of the GAP Activity, pjus 50% of the Fuel Matrix Activity. l 5.3.3.2 TMI-1 Normal RCS Activity - TMI-1 normal RCS Activity is monitored regularly during operation and is readily available to emergency dose assessment personnel. 5.3.3.2.1 Plant Shutdown Radioiodine Spiking represents the " spiking" of the radioiodines and noble gases. Plant Shutdown Radioiodine Spiking occurs when the reactor is shutdown, and is caused when the fuel gap activity is washed out of fuel rods with cladding defects. l
- Default spiking factors for current fuel conditions are monitored by l Radiological Engineering and Nuclear Engineering, and made available l to emergency dose assessment personnel.
! The TSC should be requested to provide the actual" spiking" factors for the particular plant shutdown situation. The default factors are to be used if information is not available from the TSC, l' This " spiking" of radiciodine and noble gas activities represent an increase in RCS radioactivity due to a plant evolution and do not represent an indication of fuel damage; i.e., Fuel Damage Classes 2-10. [N g Spiking only occurs when fuel defects are present. 16 I
r Number U TMl
\ NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revision No.
T.*Al Emergency Dose Calculation Manual (EDCM) 9 5.3.3.3 Gap Activity and Fuel Matrix Activity - Gap activity and Fuel Matrix activity used in the model are determined from TDR-431. TMI Unjt 1 Fuel GAP Matrix Act. Fuel isotope Act. Curies GAP % Curies Matrix % Kr-85m 4.84E+04 0.45 2.13E+07 2.36 Kr-85 7.48E+04 0.70 8.59E+04 0.01 Kr-87 2.63E+04 0.25 3.90E+07 4.33 Kr-88 6.67E+04 0.63 5.91E+07 6.56 Xe-131m 7.96E+04 0.75 5.40E+05 0.06 Xe-133m 9.30E+04 0.87 3.09E+06 0.34 Xe-133 8.34E+06 78.31 1.28E+08 14.20 Xe-135m 2.72E+04 0.26 3.37E+07 3.74 Xe-135 3.45E+04 0.32 1.59E+07 1.76 Xe-138 0.00E+00 0.00 0.00E+00 0.00 l- 1.' ' 1.29E+06 12.11 6.37E+07 7.07 L' J l-132 1.85E+05 1.74 9.70E+07 10.76 l-133 2.79E+05 2.62 1.43E+08 15.86 l-134 1.74E+04 0.16 1.67E+08 18.53 1-135 8.83E+04 0.83 1.30E+C8 - 14.42 Sum 1.13E+07 100.0 Sum 9 02E+08 100.0 17
I y
\ muc e Radiological Controls Procedure 6610-PLN-4200.02 Tale Revson No.
l TMl Emergency Dose Calculation Manual (EDCM) 9 5.3.3.4 The following tables represent the Fuel Damage Class 2 - 10 mixes, percentages, curies, and concentrations used in the RAC model. I l Damage Damage Class 2 Class 3 Ci/cc" Percent Curies Ci/cc" Percent Curies i 1-131 6.03E+02 12.09 1.29E+05 3.01E+03 12.11 6.45E+05 , i 1-132 8.66E+01 1.74 1.85E+04 4.32E+02 1.74 9.25E+04 l-133 1.31E+02 2.62 2.79E+04 6.52E+02 2.62 1.40E+05 l-134 8.43E+00 0.17 1.80E *03 4.10E+01 0.16 8.76E+03 l-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 l KR-85M 2.28E+01 0.46 4.87E+03 1.13E+02 0.45 2.42E+04 [
\
KR-85 3.50E+01 0.70 7.48E+03 1.75E+02 0.70 3.74E+04 4 l b 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 1.56E+02 0.63 3.34E+04 XE 131M 3.72E+01 0.75 7.96E+03 1.86E+02 0.75 3.98E+04 XE-133M 4.36E+01 0.87 9.32E+03 2.17E+02 0.67 4.65E+04 XE-133 3.90E+03 78.29 8.35E+05 1.95E+04 78.30 4.17E+06 XE 135M 1.28E+01 0.26 2.74E+03 6.37E+01 0.26 1.36E+04 XE-135 1.72E+01 0.34 3.68E+03 8.17E+01 0.33 1.75E+04 XE-138 1.70E-01 0.00 3.64E+01 1.70E-01 0.00 3.64E+01 l Subtotal 4.12E+03 82.55 8.81 E+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 lodine Ratio l
- " THESE Ci/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 66,595 GALLONS I
a 18 L L ..
Number
/ TMI WUCLEAR Radiological Controls Procedure i
6610-PLN-4200.02 Title Revmon No. TMI Emergency Dose Calculation Manual (EDCM) 9 x . Damage Damage Class 4 Class 5 Ci/cc** Percent Curies Ci/cc" Percent Curies l-131 6.03E+03 12.11 1.29E+06 3.58E+04 7.60 7.66E+06 l-132 8.65E+02 1.74 1.85E+05 4.62E+04 9.81 9.89E+06 l-133 1.30E+03 2.62 2.7CE+05 6.81 E+04 14.46 1.46E+07 l134 8.16E+01 0.16 1.75E+04 7.81E+04 16.59 1.67E+07 l-135 4.13E+02 0.83 8.83E+04 6.12E+04 12.99 1.31E+07 - Subtotal 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 47 1.23E+02 0.25 2.63E+04 1.83E+04 3.90 3.93E+06 [ ECM 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 1.34 E+05 6 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 2.11 E+07 4 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 4.11 E+ 04 82.54 8.79E+06 1.82E+05 38.56 3.89E+07 Totai 4.98E+04 100.00 1.07E+07 4.71 E+05 100.00 1.01 E+C8 Noblo Gas 4.73 4.73 4.73 0.63 0.63 0.63 To lodine Ratio i "THESE Ci/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 66,595 GALLONS N
)
v 19
r Numbir i ' O U TMI NUCLEAR Radiological Controls Procedure
\]. 6610-PLN-4200.02 Title Revision No.
_TMl Emergency Dose Calculation Manual (EDCM) 9 Damage Damage Class 6 Class 7 Ci/cc" Percent Curies Ci/cc" Percent Cunes 1-131 1.55E+05 7.18 3.31 E+07 3.04E+05 7.13 6.50E+07 l-132 2.28E*05 10.55 4.87E+07 4.54E+05 10.66 9.72E+07 l-133 3.35E+05 15.56 7.18E+07 6.70E+05 15.71 1.43E+08 l-134 3.90E+05 18.10 8.35E+07 7.80E+05 18.31 1.67E+08 l-135 3.04E+05 14.11 6.51E+07 6.08E+05 14.26 1.30E+08 Subtotal 1.41 E+ 06 65.50 3.02E +08 2.82E+06 66.07 6.03E+08 KF;-85M 5.00E+04 2.32 1.07E+07 9.98E+04 2.34 2.13E+07 KFt 85 5.50E+02 0.03 1.18E+05 7.51E+02 0.02 1.61 E+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+0? 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 1.5BE+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 3.64E+01 1.70E-01 0.00 3.64E+01 Subtotal 7.44E+05 34.50 1.59E+08 1.45E+06 33.93 3.10E+08 Total 2.16E+06 100.00 4.61 E+08 4.26E+06 100.00 9.12E+08 Noble Gas 0.53 0.53 0.53 0.51 0.51 0.51 To louine Ratio "THESE Ci/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 66,595 GALLONS l l I l 20 ! I L- l
r Number TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 Damage Damage Class 8 Class 9 Ci/cc a Percent Curies Oi/cc" Percent Curies 1-131 3.04E+05 7.13 6.50E+07 3.04E+05 7.13 6.50E+07 l-132 4.54E+05 10.66 9.72E+07 4.54E+05 10.66 9.72E+07 l-133 6.70E+05 15.71 1.43E+08 6.70E+05 15.71 1.43E+08 l134 7.80E+05 18.31 1.67E+08 7.80E+05 18.31 1.67E+08 l-135 6.08E+05 14.26 1.30E+08 6.08E+05 14.26 1.30E+08 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 7.51 E+02 0.02 1.61 E+05 7.51 E+02 0.02 1.61 E+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 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 To lodine Ratio "THESE Ci/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 66,595 GALLONS O
21 l l 2 l l
F ( Number U TMI g NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revrsion No TMI Emergency Dose Calculation Manual (EDCM) 9 Damage Class 10 pCi/cc" Percent Curies 1131 3.04E+05 7.13 6.50E+07 l-132 4.54E+05 10.66 9.72E+07 l-133 6.70E+05 15.71 1.43E+08 l-134 7.80E+05 18.31 1.67E+08 l-135 6.08E+05 14.26 1.30E+08 Subtotal 2.82E+06 66.07 6.03E+08 KR-85M G.98E+04 2.34 2.13E+07 , KR 85 7.51E+02 0.02 1.61 E+05 KR-87 1.82E+05 4.28 3.90E+07 b
\
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 XE 138 1.70E-01 0.00 3.64E+01 Subtotal 1.45E+06 33.93 3.10E+08 Total 4.26E+06 100.00 9.12E+08 Noble Gas 0.51 0.51 0.51 To lodine Ratio "THESE Ci/cc VALUES ARE BASED ON NORMAL RCS VOLUME OF 66,595 GALLONS O V 22 l l-
Number h\/ U TMI g NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tile Revision No. ! TMl Emergency Dose Calculation Manual (EDCM) 9 l 5.3.3.5 Calculation of Radionuclide Mix Percentages - Once the total amount of combined activities or curies for a certain Fuel Damage Class has been determined, these ) curies are then normalized to 100%, i e., the percentage of each radionuclide in the total mix is calculated. Exceptions - The above percentages are replaced in cases where an assumed mix is more appropriate. These cases are: , O Contingency calculations for: l a. Spent Fuel Accident in the Fuel Handling Building - FSAR mix l is assumed. l
- b. Fuel Cask Accident in the Fuel Handling Building - FSAR mix ,
is assumed. l
- c. Spent Fuel Accident in the Reactor Building - FSAR mix is assumed.
- d. Waste Gas Decay Tank - FSAR mix is assumed.
O ( l ( LJ ' 23
Number , U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tee Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 e The RCS inputs for these cases are as follows: RCS ACTIVITY INPUTS (1) (3) (4) RB FHB (5) (2) FUEL FUEL FUEL WGDT HANDLING HANDLING CASK ISOTOPE RUPTURE ACCIDENT ACCIDENT ACCIDENT KR85M 5.56E-02 2.88E-06 9.54E-08 0.00E+00 KR85 3.23E-02 2.10E-02 1.19E-01 1.00E+ 00 KR87 3.03E-03 0.00E+00 0.00E+00 0.00E+00 KR88 9.79E-03 1.13E-08 0.00E+00 0.00E+00 XE131M 8.68E-03 4.12E-03 8.45E-03 1.07E-04 XE133M 1.01 E-02 1.36E-02 5.77E-03 0.00E+00 XE133 9.05E-01 9.53E-01 8.65E-01 2.14E-06 XE135M 3.43E-03 0.00E+00 0.00E+00 0.00E+00 V XE135 2.16E-02 1.03E-03 2.62E-04 0.00E+00 XE138 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1131 1.11 E-04 6.37E-03 1.72E-03 6.76E-05 l 1132 1.72E-04 0.00E+00 0.00E+00 0.00E+00 1133 1.32E-04 1.22E-03 4.03E-05 0.00E+00 1134 2.02E-05 0.00E+00 0.00E+00 0.00E+00 1135 7.07E-05 8.06E-06 7.46E-08 0.00E+00 TOTAL 1.00E+00 1.00E+00 1.00E+00 1.00E+00 (1) Normalized isotopic activities released from FSAR analysis to be input into model as RCS activity to get proper monitor response. (2) From FSAR Table 14.2 21 (3) From FSAR Table 14.2 5, Decayed to 72 hours (4) From FSAR Table 14.2-2 (5) From FSAR Table 14.2-25 O 24
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U TMI NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. l TMl Emergency Dose Calculation Manual (EDCM) 9 j l ' 5.4 Radiation Monitorina system fRMS) Source Term Calculation - This portion of the model allows the user to determine an effluent source term from readings on the TMI 1 Radiation Monitoring System. 5.4.1 Source Term Calcuiations Using Effluent Monitors l 5.4.1.1 To calculate a source term from a RMS reading the following parameters are used: O RMS READING: CPM, mR/HR, OR CPM / MIN 8 RMS CHANNEL EFFICIENCY RELATING TO THE CALIBRATION NUCLIDE: CPM / Cl/CC e THE METER RESPONSE FACTOR S THE FUEL DAMAGE CLASS MIXTURE e THE RELEASE FLOW RATE l l 5.4.1.2 The RMS reading, the particular radiation monitor's efficiency, the monitor p response factor, the nuclide fraction from the isotopic percentage section of the program relating to fuel damage class determination, radionuclide mix i( percentages, and the associated flow rate to the environment are used to calculate a source term. Source terms are identified for the noble gas source term and for the radiciodines. 5.4.1.3 The calculations are performed in the following fashion: O First the total monitor response factor is calculated by multiplying the individual nuclide percentages from the fuel damage class determination by the individual nuclide monitor response factors. 15 M= I P
- In 1
Where: M= total monitor response factor P= individual nuclide percentages from fuel damage class
, in = individual nuclide monitor response factors i
26 i
k' l Number [ U TMl WUCLEAR Radiological Controls Procadure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 l The In's for the various RMS detectors are listed as follows: Individual Mixture Response Factors (In) l Beta Scint. Detectors RM-A2Lo, RM-A4Lo, GM Tubes RM-A6Lo, RM-A5Lo, Scintillation Detectors lon Chamber RM-ASHi, RM-A8Hi RM-A8Lo, RM-A9Lo (4) Nuclide RM-G26 & RM-G27 (1) RM-G24(2) (3), RM-A9Hi, RM-G25 RM-A15Lo RM-A14Lo Kr-85m 70.7 212.2 2.35 1.92 Kr-85 1.0 1 0.011 1.98 Kr-87 356 324.39 3.59 9.12 Kr-88 1160 334.15 3.70 2.78 l Xe-131m 9.01 4.88 0.054 0.0 j i Xe-133m 18.6 35.15 0.378 0.0 1 Xe-133 0.0 90.24 1 1.0 l
>O (<80 kev)
I Xe-135m 193 195.12 2.16 0.0 Xe-135 111 221.95 2.54 2.59 Xe-138 1560 939.02 10.41 4.62 1-131 172 240 2.66 l-132 1030 747.8 8.286 l-133 274 219.51 2.432 1134 1180 542.44 6.011 l-135m 706 341.46 3.784 (1) I MeV* Dis: Nuclide I MeV* Dis cal Nuclido (calibration isotope is Kr-85, threshold set to exclude Xe-133 at 80 kev.) I (2) I% Probabilitv Nuclide I% Probability Cat. Nuclide (calibrc* ion nuclide is Kr-85) (3) I% Probabilitv Nuclide , I% Probability Cal. Nuclide (calibration isotcpe is Xe-133; values for RM-ASHi, except Xe-133, will l be multiplied by 4) (4) I Beta decay probability
- Beta end-Dt. enerQV nuClide I Beta decay probability
- Beta end-pt, energy cal. nuclide (cal. isotope is Xe-133)
- Radiciodines filtered out prior to noble gas channel 27
I l l Numbrr
}
bUNUCLEAR TMl Radiological Controls Procedure 6610 PLN-4200.02 ! l Title Revision No. I TMI Emergency Dose Calculation Manual (EDCM) 9 0 The noble gas source term in Ci/sec is now calculated using the following equation and input data: Ngst = (_1_
- ACT
- 1_
- Ng/100' * (Flow)
- 472 M _,Me Where: Ngst = Noble Gas source term in Ci/sec. ,
I M = Total monitor response factor, unitiess. J Act = cpm, mR/hr, cpm / min reading from the monitor. Me = Monitor sensitivity in cpm / Ci/cc, mR/hr/ Ci/cc, or cpm / min / Ci/cc. Flow = Flow rate in CFM 472 = cc/sec/CFM. Ng = Sum of Noble gas percentages from the selected Fuel Damage classification.
'- e The radioiodine source term in i/sec is then calculated using the noble gas source term and the noble gas to iodine ratio, as discussed in 3ection 5.8.
Rist = Ngst * @
- RDF Ng Where: Rist = Radiciodine source term in Ci/sec.
E = The radioiodine to noble gas ratio. Ng RDF = lodine reduction factor for the release pathway as discussed in Section 5.11, O The noble gas and the radiciodine source terms are then multiplied by the individualisotopic percentages of the fuel damage class mixture to determine the Ci/sec of each of the 15 nuclides,10 noble gas and 5 radioiodines. l l O LU t 28 j l l
I l Number
\ U TMI ,V "CN Radiological Controls Procedure 6610-PLN-4200.02 j Tdie Revtston No TMI Emergency Dose Calculation Manual (EDCM) 9 5.4.2 Source Term Calculations Using RM-G-22 and RM-G-23 l 5.4.2.1 During a Loss of Coolant Accident in the Reactor Building, the high range radiation monitors in the Reactor Building, RM-G-22 and RM-G-23, may be used to estimate the airbome activityin containment. Use of these monitors to estimate the activity released to the containment atmosphere is based on the following assumptions:
O The response of RM-G-22 and 23 to airbome activity in containment I during a LOCA is assumed to be conservatively approximated as a l semi-infinite cloud surrounding the detector. I O lt is assumed that the detectors are responding to noble gases only. While some response from airbome iodines would be expected, it is difficult to predict the fraction of iodines released from the RCS that would remain airborne for all LOCA situations. As a result, it is conservatively assumed that all response of the monitors is from noble gases. 9 The isotopic mix of the noble gas activity in containment is assumed to be the same as the distribution in the RCS for a Fuel Damage Class 2. 7 ( ]/ w,
< Since the isotopic distribution in the RCS does not dramatically change between Fuel Damage Classes 1-4, this assumption covers a wide range of the most probable core damage conditions. In addition, this mix produces higher Ci/cc/R/hr values than mixes above Damage Class 4, so it is conservative for those conditions.
5.4.2.2 The noble gas source term in Ci/sec is now calculated using the following equation and input data: Ngst = (ACT)(1/65.6)(Flow)(472) Where: Ngst = Noble Gas source term in Ci/sec ACT = R/hr reading from the monitor 65.6 = Monitor sensitivity in R/hr/ Ci/cc per Reference 7.43 Flow = Flow rate in CFM 472 = cc/sec/CFM r 29
r-Number U TMl suct.ran Radiological Controls Procedure 6610-PLN-4200.02 Tale ( Revrsion No l TMI Emergency Dose Calculation Manual (EDCM) 9 5.4.2.3 The radioiodine source term in Ci/sec is then calculated using the noble gas l source term and the noble gas to iodine ratio, as discussed in Section 5.8. Rist = (Ngst)(Ri/Ng)(RDF) Where: Rist = Radioiodine source term in pCi/sec. Ngst = Noble gas source term in Ci/sec Ri/Ng = The radioiodine to noble gas ratio RDF = lodine Reduction Factor for the pathway as discussed in Section 5.11. 5.4.2.4 The noble gas and the radioiodine source terms are then multiplied by the individual isotopic percentages of the Fuel damage class mixture to determine the Ci/sec of each of the 15 nuclides,10 noble gas and 5 radioiodines.
\
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Number [ U NUCLEAA TMI Radiological Controls Procedure 6610-PLN-4200.02 3e Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.5 Post Accident Samnies Source Term Calculation - Actual plant effluent sample results may be used to develop the release source temis. This is in fact the preferable method for estimating release quantities if the sample results are available since the method eliminates some of the built in conservatisms of using monitor readings, or contingency calculations. Using sample results also eliminates errors in the source term when the actual release mixture is diffennt from the assumed mix. The use of pos' accident samples in the RAC modelis dependent on the type of post accident sample. 5.5.1 One criteria is the sample station / method to be used. For the Reactor Building three options are presented: 1) CATPASS (Containment Atmosphere Post Accident Sample System),
- 2) Marinelli/prefilter (marinelli with a particulate and iodine filter upstream), or 3) MAP-5, Radiciodine Processor Station. For the condenser off-gas or the Aux /FHB release pathways, only the Marinelli/prefilter and MAP-5 samples are available.
5.5.2 For MAP-5 sample results each of the identified radioiodine species in the silver zeolif 3 or charcoal cartridge sample from the MAP-5 Processor Station are used for source term generation. The mixture can be decayed 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. Since the MAP-5 only provides information on the radiciodines, the expected ratio between the n iodines and noble gases is used to approximate the noble gas activity. The MAP-5 [ j measures radiciodines after they have been acted on by the lodine reduction factors for the (/ release pathway under consideration. These reduction factors are described in Section 5.11. As a result, the noble gas activity will be increased by a factor that is the inverse of the reduction factors described in Section 5.11 to account for reduction of the lodines that is not applicable to the noble gases. Based on the release rate in CFM the isotopic concentrations in Ci/cc are converted to a release rate in Ci/sec for the final source term. 5.5.3 For CATPASS sample results the 10 noble gas and the five radiciodine nuclides identified ] from air sampling are used for source term generation. The mix can be decayed from the { sample to dose projection time. Since the CATPASS only applies to the containment, j whether the containment is isolated or not and if the release is proposed or in progress must be considered. Calculations are made using the input activities to develop new isotopic percentages and the activities are changed from Ci/cc to pCi/sec based on the release rate
]
defined to arrive at the final source term. j i 5.5.4 The marinelli sample results are used if a marinelli/prefitter sample has been collected. 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. l l 1 [ I
\v/
31
( l Number U TMI suct m Radiological Controls Procedure 6610-PLN-4200.02 l Title Revision No TMl Emergency Dose Calculation Manual (EDCM) 9 5.6 Continoency Calculations Source Term Generation - The contingency calculations determine a source term based upon a prioritized set of plant conditions. In this way, credible conservative assumptions, as defined in the FSAR default parameters, are replaced with real-time accident conditions as ir dicated by plant instrumentation. This will make the calculated source terms more realistic. 5.6.1 in the contingency calculations the previously determined release pathway is c'::: zed to select: O Secondary Side Release l 9 Reactor Building Release e Station Ventilation Releases The " Secondary Side Release" includes accidents that result in release via: The condenser off-gas, the atmospheric dump valves, the main steam reliefs, and a main steam line rupture. The " Reactor Building Release" includes accidents that result in a release from the Reactor Building via: the purge duct, when the purge valves are open, or design basis leakage, when the purge valves are closed. ( n The " Station Ventilation Releases" include accidents that result in a release from the Auxiliary
) Building, Fuel Handling Building, or ESF Fuel Handling Building.
d 5.6.2 A " Secondary Side Release" contingency calculation is calculated by identifying four parameters: e RCS Activity [D1] pCi/cc e Primary to Secondary Leakage [D2] gpm e lodine Reduction Factor (RDF)[D3] 6 Two Phase Release [Tfcf] 5.6.2.1 The "RCS Activity"is determined utilizing: e RM-L1 High [A1] cpm, D1 = A1/22.2 Ci/cc e RM L1 Lo [A1) cpm, D1 = A1/1330 Cl/cc e Most recent RCS sample,in Ci/cc, or i L) 32
Number t '\ TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 TRie Revision No. 1 1 TMI Emergency Dose Calculation Manual (EDCM) 9 e Default to a RCS concentration dependent on the Fuel Damage Class: RCS Default Fuel Damage Class Ci/cc I 1 l 1A l 2 4.99E+03 1 3 2.49E+04 f 4 4.98E+04 5 4.71E+05 6 2.16E+06 i 7 4.26C+06 l 8 4.26E+06 9 4.26E+06 10 4.26E+06 Damago classes 1 and 1 A are variable, based on RCS activity entered from sample data and i radioiodine spiking factors used. 5.6.2.2 The
- Primary to Secondarv Leakage"is determined utilizing:
V O RCS identified leakage [D2] gpm 9 Default to 400 gpm for a double-ended tube shear [D2] 5.6.2.3 The 'RDF"[D3]is a function of the release pathway. They are discussed in Section 5.11. The resultant source terms in Ci/sec are calculated by: l Ngst = Di
- D2
- Ng/100
- 63.09 Rist = D1
- D2
- D3
- Rl/100
- 63.09
- 1/Tfcf.
l Tfcf is described in Section 5.10. l l l O 33
1 l Number
/D U TMI I id NUCLEAR Radiological Controls Procedure 6M O-PLN-4200.02 l
Title Revtsion No. l TMI Emergency Dose Calculation Manual (EDCM) 9
]
l 5.6.3 A " Reactor Building Release" contingency calculation is calculated by one of two methods. ' if the accident type is a LOCA then four parameters are identified: l 0 RCS Activity [A2) Ci'cc e RCS Leakage to RB [A3) gallons e RDF, E4 as discussed in Section 5.11. l 0 Release Flow Rate CFM; E3 = flow
- 472 to convert CFM to cc/sec 5.6.3.1 The RCS Activity is determined utilizing:
O RM-L1 High Channel [A1) cpm; [A2) = [A1)/22.2 Ci/cc 8 RM-L1 Low Channel [A1) cpm; [A2) = [A1)/1270 Ci/cc e Representati/e RCS sample results [A2)in Ci/cc 0 Default to a RCS concentration dependent on the Fuel Damage Class: RCS Default Fuel Damage Class pCi/cc 1A 2 4.99E+03 3 2.49E+04 4 4.98E+04 5 4.71 E+05 6 2.16E+06 7 4.26E+06 8 4.26E+06 9 4.26E+06 10 4.26E+06 Damage classes 1 and 1 A are variable, based on RCS activity entered from sample data and radioiodine spiking factors used. 9 Default mix according to core condition as previously identified. The noble gas and radioiodine released to the Reactor Building is calculated as follows: E1 = A2
- A3
- Ng/100
- 3785/5.6E10 pCi E2 = A2
- A3
- Ri/100
- 3785/5.6E10 Cl 34 l
r Numbu p (abu NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tatie Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 8 If the accident type is a Fuel Handling Accident in the Reactor Building, then the number of damaged fuel rods is identified by the " user" or an FSAR default condition is used. l E1 e 1.7
- Num rod /208 I
E2 = 0.05
- Num rod /208 5.6.3.2 RCS LEAKAGE - The RCS leakage to the Reactor Bui: ding is tne " total gallons of RCS leakage into the R3".
5.6.3.3 RDF - This fraction is used to reduce the radiciodines available for release in the Reacter Building, as discussed in Section 5.11, 5.6.3.4 RELEASE FLOW RATE - The release flow rate is determined via flow rate recorder FR-148 if the purge valves are open. If the purge valves are closed the release flow rate is determined via the design basis RB leakrate adjusted for actual RB intemal pressure as indicated on PT-291. 5.6.4 A " Station Ventilation Release" contingency calculation is calculated by utilizing the FSAR p fuel assembly mix or WGDT mix. O Waste Gas Release 9 Fuel Handling Accident in the Fuel Handling Building Q ESF Fuel Handling Release 5.6.4.1 The extent of the accident is modeled by: O Determining the number of damaged fuel rods for a fuel handling accident in the Fuel Handling Building, or for the ESF FHB. Ngst = 4.2E6
- Num rods /56 Ci/Second Rist = 750
- Num rods /56 pCi/Second S Using the worst case FSAR source term for WGDT's of 10,000 curies of noble gas and 5 curies of radioiodine, or using the typical WGDT FSAR mix of 1000 curies of noble gas and 5 curies of radioiodine.
i l l r~%. U 35
r l i l-l Number TMI s
"CW8 Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 l 9 Determining the curies released for a Waste Gas Accident. Typical source term based on a typicalinventory of: Ngst = 1.0E9/Dr Ci/second Rist = 1.0E5/Dr Ci/second or FSAR worst case: Ngst = 1.0E10/Dr Ci/second Rist = SE6/Dr Ci/second Where: Dr = duration of release. 1 J l l l l I l [
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36 l
Number
-[
U TMI
'v NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Revrsion No.
TMl Emergency Dose Calculation Manual (EDCM) 9 5.7 Decay Scheme Calculation - The model provides the capability 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 J time the sample is obtained to the time of dose projection. The calculation only decays forward in j time. This calculation adjusts the individual nuclide percentages according to the conventional j exponential decay equation: A=Ao exp (-At) , 5.7.1 Fifteen isotopes are decayed according to the equation N(w) = f(w)
- EXP (-decay time
- f(w))
where: l(w) = postulated isotopic percentage decay time = user input time f(w) = isotopic decay constants s 5.7.1.1 The adjusted isotopic percentages N(w) are corrected for Xenon buildup due to I iodine decay. For Xe-131m the equations are; j S1 = l(11)- N(11) N(5) = 0.88
- S1 + N(5) where:
S1 = amount of I-131 decayed 0.88
- S1 = amount of Xe-131M buildup 5.7.1.2 For Xe-133M the equations are:
S1 = l(13) - N(13) N(6) = 0.02
- S1 + N(6) where:
S1 = amount of I 133 decayed 0.02
- S1 = amount of Xe-133m buildup t
(. N(7) = 0.98
- S1 + N(7) calculates the amount of Xe-133 buildup from l-133 h
V 37 i l t
l l TMI ' NUCLEAR Radiological Cor,trols Procedure l
'N 6610 PLN-4200.02 Title Revision No.
J TMl Emergency Dose Calculation Manual (EDCM) 9 5.7.1.3 For Xe-135m and Xe-135 the equations are: l S1 = l(15) - N(15) N(8) = 0.3
- S1 + N(8)
N(9) = 0.7
- S1 + N(9) where:
l Si = amount of I-135 decayed 0.3
- S1 = amount of Xe-135m buildup 0.7
- Si = amount of Xe-135 buildup l l
5.~7.1.4 The isotopic percentages are recalculated as: l(w) = N(w)/ Sum (N)
- 100 where:
N(w) = adjusted / corrected postulated isotopic percentages Sum (N) = sum of the fifteen isotopic percentages l(w) = finalisotopic percentages based upon 100. O 38
Number U TMI NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title ~ Revrmon No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.8 Noble Gas to lodine Ratio Calculations 5.8.1 Whether performing dose projections based upon RMS readings, post accident samples or contingency calculations, it may be necessary to compute the NOBLE GAS TO IODINE RATIO. The uses of this ratio are discussed below. An airborne release from a nuclear power plant will primarily consist of noble gases and radioiodines. Except in the most severe and improbable accident scenarios, radioactive particulates are not expected to be important dose contributors. The RAC model was designed to incorporate ten noble gases and five radioiodines. The 15 isotopes are considered to be the most radiologically significant gaseous isotopes available for release from an operating nuclear power phnt. Pertinent radioactive decay parameters such as half life, average gamma energy per disintegration and average beta energy per disintegration for each isotope, and individualisotope source term information are used to determine dose rate conversion factors and dose rates that are specific to the isotopic mixture being released. These calculated quantities can be adjusted to account for radioactive decay during the accident sequence. 5.8.2 The TMI RAC model always projects both thyroid (CDE) and whole body (TEDE) 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 model starts
() with a core inventory of all fifteen isotopes and traces the progress of each one through vanous systems or processes untilit 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 neghgibly small. But, in gen 6ral, the model 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 I others have padiculate, iodine and gas channels. The MAP-5 sampling system yields only iodine information, where the CATPASS and the Marinelli gas sampling systems yield information on all fifteen isotopes. For release pathways where information on both noble gases and iodines is not available, the RAC model uses the noble gas to iodine ratio to fillin 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 model uses a default mixture of 15 nuclides and calculates the fraction of the mix that each isotope represents. The noble gas to iodine ratio is also calculated. Assume that the ratio was equal to 5/1 in this case. Also assume that an iodine sample was taken which indicated a total radioiodine release rate of 5000 Cifsec. Using the noble gas to iodine ratio in the absence of specific noble gas measurements, the model would calculais a gross noble gas release rate of 25,000 Ci/sec. It would also calculate individual noble gas release rates by using the isotopic fractions from the default mix. s m J 39
r-- j 1 Nurnber U TMI l (Q NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 1 i T9 Revmion No. l TMI Emergency Dose Calculation Manual (EDCM) 9 l 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 lodine ratio in these cases. The second best measurement would be one that yielded gross noble gas and gross iodine readings. This situation occurs in the low range radiation monitors which have l 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 lodine measurements are not available, for example, when only noble i gas or only iodine information is available. l 5.S.4 There are some refinements and subtleties that the model user should be aware of. The noble gas to iodine ratio changes with time because of radioactive decay. The RAC model j has the ability to account for radioactive decay and to compute a decay corrected noble gas l to lodine ratio. As explained elsewhere in this manual, the model also corrects the Dose Rate Conversion Factor (DRCF) for decay of the isotopes in the mix. When performing dose projections several hours or more after the reactor has tripped, these two decay corrections can significantly alter the resultant projections. The model provides the capability to account for decay between reactor trip and dose projection. p 5.8.5 For dose projections based upon RMS readings, the decay correction to the noble gas to j iodine ratio is straightforward. A default mixture of the fifteen noble gases and iodines is ()) r 1 selected based upon an assessment of core damage. If the mix,is to be decayed, all fifteen isotopes are decayed by the standard exponential decay law using the decay time between reactor trip and dose projection. The noble gases and iodines are totaled separately so that their ratio at dose projection time can be calculated. (Ingrowth of xenon isotopes from decay of iodine is accounted for.) The decay adjusted ratio is then used to fill in the missing noble gas or lodine information, 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 fuel damage classification. For a dose calculation based upon a radiciodine processor sample, if decay correction is desired, the model uses two decay time intervals: O The time between sampling and dose projection 9 The time between reactor trip and dose projection Sample results from the radiochemistry lab are reported as of the sample collection time. When significant time has elapsed between sampling and dose projection, the results should be decayed from sampling time to dose projection time. In order to compute the l 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 lodine ratio is computed for the decayed default mix, and any iodine reduction factors present in the specific release pathway. This ratio, along l with the gross radioiodine sample result, is used to compute a gross noble gas source term. Isotopic source terms are calculated from the decayed mixture noble gas fractions. Note that the final source term is a combination of noble gases from a dafault rmx and radiciodines from a sample. Each has been decayed to the dose projechon time. A word of caution should be added at this point. The iodine released in certain types of s accidents may be reduced by various chemical and physical processes such as iodine
\ plateout or formation of water soluble iodide salts. The rioble gas to iodine ratio, as calculated above, may not completely account for this iodine reduction. As a consequence,
! 40 L
e nurnber INI NUCLEAR Radiological Contro4 Procodure
\ 6610-PLN-4200.02 Tale Revision No.
- TMl Emergency Dose Calculation Manual (EDCM) 9
! 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 possibihty. A co:nparison of field team data and the source term dose projections would reveal agreement for thyroid (CDE) doses, but not for whole body (TEDE) doses. 1 i l l I O 41
Number U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.9 Effluent Release Flow Rates - Flow rates for effluent releases to the environment are divided into four categories: e Normal ventilation flow rates. 8 Reactor Building leakage flow rates. O Adjacent momentum plume rise (station vent and reactor purge concurrently releasing). O Flow rates for OTSG tube rupture release directly to atmosphere. ,
- Buoyant plume rise j
- Source term calculation using RMG-26 or RMG-27 Source term calculation using a contingency calculation These flow rates for accident source terms to the environment are calculated as follows: )
l 5.9.1 TMI-1 Normal Ventilation Flow Rates [m'i The RAC Model provides the option to use the actual ventilation flow rates as read from the Q / 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: e Reactor Building Purge
-FR909 0-20,000 CFM; Low Range I l
-FR1488 0-50,000 CFM; High Range 9 Reactor Building Purge and Make-up Exhaust
-FR148A 0-50,000 CFM e Reactor Building Hydrogen Purge System
-Fl282 5-50 CFM
-Fl283 20-200 CFM
-Fl284 100-1000 CFM
-Total 5-1250 CFM e Kidney Filter System s -AHE-101, AH-F-12 P/l filters 20,200 SCFM v) 42
l Number b) y U NUCLEAR TMI Radiological Controls Procedurc 6610-PLN-4200.02 Title Revision No. ( I TMI Emergency Dose Calculation Manual (EDCM) 9 9 Auxiliary Building Exhaust l -FR150 0-100,000 CFM l S Fuel Handling Building Exhaust
-FR149 0-50,000 CFM e Auxiliary and Fuel Handling Building Exhausts
-FR-151 0-150,000 CFM S Condenser Off-Gas Exhaust
-RMR15 Recorder FT 1113 Ch. A0-200 CFM l @ ESF Fuel Handling Building Exhaust
-No Flow Recorder at this time 0-8000 CFM l f3 Default values are used in the RAC Model 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 l
- Auxiliary Building Exhaust Fuel Handling Exhaust
- Auxiliary and Fuel Handling Building Exhausts.
40 CFM - Condenser Off-Gas 7000 CFM - ESF FHB Exhaust These default values allow the user to continue with dose projections even though a value is small or unknown. Therefore, once the dose projection is complete, the results may be raticed up or down depending on the situation. For example, if the default value of 5000 l CFM was used for a Reactor Building Purge and subsequently an engineering calculation j was performed indicating 1000 CFM flow. The dose projection could be ratioed down by l one fifth (1/S). 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.
)
G 43
r-Number ! U TMI q ) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 l v ; j Title Revrsion No. TMl Emergency Dose Calculation Manual (EDCM) 9 5.9.2 Reactor Building Leakage Flow Rate Another section of the model calculates a leakage flow rate out of the Reactor Building based on Reactor Building pressure. The Reactor Building pressure indicatoris 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(Pr/PA) where: Lr = Reactor Building Leak Rate in CFM LA = Maximum allowable integrated leakage rate at PA LA = 6.14 CFM PA = Peak Reactor Building internal pressure at design basis accident, PA = 50.6 psig Pr = Actual Reactor Building internal pressure in psig Therefore, the maximum leakage allowed at a design basis accident pressure of 50.6 psig is j n 6.14 CFM. Leak rates at 0-60 psig can be calculated from the above formu;a. The default j value in the modelis 50.6 psig. A graphic representation follows in Figure 5.0-1. 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 j can be calculated from the flow rate according to the following formula: l
,,y_ (1) w= nr2 l
where w e stack gas exit velocity V = flow ra' or volume flux ' r = radius c stack The Station Vent and the tieactor 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:
. I. wl/ (2) w= IV i Where w = exit velocity due to mixing l
l O O l 44
l Number U TMl NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Rewsion No. l TMl Emergency Dose Calculation Manual (EDCM) 9 l l if these adjacent stacks were modeled as a single stack, the radius of the stack would be given by-l I= [v (3) . xW At TMI-1, the reactor building stack and station vent are adjacent stacks; For computing plume rise, the stack gas exit velocity and stack radius were calculated according to Eq. (2) and (3) above. A comparison 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 Ruoture Release Directly to the Atmosphere (see Figure 5.9-2 and Figure 5.9 3). TMI-1 has 22 main steam relief and 2 atmospheric durap 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. j 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 l Q
; f 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 V (2) valves can be read at the center control panel under the turbine bypass dump controller for MS-V-4A/B from 0 - 1005 Each of the 22 valves at TMI 1 has a stack or vent where the steam is ejected into the atmosphere. The location of these stacks is shown in Figure 5.9-2. If a steam generator tube ruptures, each of the 22 valves and stacks acts as a throttle to limit the flow from the steam line to the atmosphere. When a valve opens, the flow through it will be approximately equalio the rated flow, and the flow can be assumed to be approximately constant until the pressure in the steam lino 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 sma!! fraction of the rated flow. For either a normally operating valve or stuck open valve, if the pressure and temperature in the steam line ere known, the conditions just beyond the stack exit can be estimated by assuming expalsion of the steam to atmospheric pressure and temperature.
5.9.4.1 Buovant Plume Rise 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. i For a PWR, plume rise can increase the effective stack height by a factor of 5 to 50. Since I 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 100 times higher than that from a plume rise of 2000 feet. O Modeling of plume rise begins with modeling the steam condition at the valve inlet. (U ) Table 5.9-3 outlines the calculational steps required to compute buoyant plume rise, beginning with the va!ve intet. The far left side of the table identifies the area for which the l calculation applies: valve inlet, top of stack, jet origin, and plume rise. For each area, i 45 L
l i l Number U TMI NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 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 i calculations are discuwed in the Environmental Controls document Potentially Buovant i 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 the plume may be unable to penetrate the layer and the effective stack height will be reduced to the height of the layer. This may cause ground concentrations to be higher and closer to the plant than predicted. In these conditions (i.e., highly stable meteorology with a buoyant plume) off site monitoring will provide an indication of the magnitude of the effect. It may also be possible to estimate the effect through visual observation of the plume. S.9.4.2 Source Term Calculation Usinn RM-G-26 or RM-G-27 isee Fioure 5.9-3) (
) RMG-26 and RMG-27 are effective in calculating a primary to secondary release
\ source term direct to the atmosphere when: 0 Atmospheric Dump Values (ADV) MS-V-4A or MS-V-4B are open from 0-100%, as indicated on Control Room Panel"CC", and releasing radioactive steam to the environment, and/or, O Emergency Feed Pump (EFP) relief valves, MS V-22A or MS-V-22B are open and releasing radioactive steam to the environment, and/or, 0 9P is in operation and releasing radioactive steam from the EFP exhaust to the environment. O Steam bypass dump to the condenser through MS-V-8A/B. When the model calculates a release from an OTSG tube rupture directly to the atmosphere and uses RM-G 26 or RM G-27 readings, the calculation Steam Flow Computation is used to determine a release flow rate, depending on which of the valves are open (Table 5.9-2 and Reference 7.44). The mass flow rate from each open valve is 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-G-26/27 readings converted to concentration using the monitor efficiencies to give Ci/second. fM f I
%.)
46
Numbu [ U V} Tme "CN TMI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 NOTE Calculation of a source term using RMS (RM-G-26/27)is dependent on the Atmospheric Dump Valves (ADV) status. If the ADVis open, the calculation is appropriate. If the ADVis closed but plant conditions (OTSG leakrate and core damage) have not changed significantly, or there is other flow past the monitors as noted above, then the use of a RM-G-26/27 peak reading will be appropriate. If the ADVis closed and plant conditions have changed significantly, and there is no other source of flow downstream of MS-V 2A/B, then the contingency calculation applies. 5.9.4.3 Source Temi Calcula 'on Usina a Continoency Calculation When the user pedorms a Contingency Calculation due to the lack of sample results or RM-G-26/27 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 genera, oressure is less than the set point pressure, then the flow rate is based on the tables supplied by the valve manufacturer. These tables have been incorporated into the RAC model. The flow from all the valves is totalled and modeled as a release to , the atmosphere. E s 47
Number i ,OT U TMI g] NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. _TMI Emergency Dose Calculation Manual (EDCM) 9 TABLE 5.9-1 l Adjacent Plume Rise at TMi Reactor Bldg. Stack and Station Vent Stack Actual Flow Characteristics RAC Model MIDAS Stability Reac Bld Stack Station Stack Flow Stack Plume Flow Stack Plume Stack Diameter Vent Diameter Rate Diameter Rise Rate Diameter Rise (cfm) (m) (cfm) (m) (cfm) (m) (ft) (cfm) (m) (ft)
~.
Unstablei; 10,000 1.1 10,000 1.7 20,000 1.847 30.3 10,000 1.1 25.4 Neutrali- 10,000 1.7 16.4 (ClassA)j
, 10,000 1.1 120,000 1.7 130,000 1.827 199.2 65,000 1.1 165.4 f'% 65,000 1.7 107.0 y ..
Statie J 10,000 1.1 10,000 1.7 20,000 1.847 24.8 10,000 1.1 22.1 10,000 1.7 16.5 (C'l ass F)q 10,000 1.1 120,000 1.7 130,000 1.827 85.8 65,000 1.1 75.9 65,000 1.7 57.0 l l l O 48 L_
n Number U
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TMI I (v- ) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Reymon No. TMI Err.ergency Dose Calculation Manual (EDCM) 9 TABLE 5.9-2 l TMI STEAM GENERATOR RELIEF VALVES l Valve Stack j Valve Valve Discharge Set Point Stack Number Function Manufacturer Area Pressure Flow Rate Diameter (MS-V) (sq.in.) (psig) (Ibs/hr) (inches) 17A-D Relief Valves, Bank 1 Dresser / 16 1050 792,617 10.02 Consolidated 18A-D Relief Valves, Bank 2 Dresser / 16 1060 800,065 10.02 Consolidated 19A-D Relief Valves, Bank 3 Dresser / 16 1080 814,960 10.02 Consolidated l 20A&D Relief Valves, Bank 1 Dresser / 16 1050 792,617 10.02 yr Consolidated 20B&C Relief Valves Bank 4 Dresser / 16 1092.5 824,269 10.02 Consolidated 21A&B Relief Valves, Dresser / 3.97 1040 194,820 10.02 i Small Safety Consolidated l 22A Safety Relief, Lonergan 6.38 200 70,211 13.13 Emergency Feed Pump , 22B Safety Valve, Emerg. Lonergan 6.38 220 76,795 13.13 l (F.W.P.T. Steam Inlet) l 4A&B Relief Valves (manual) to Fisher Variable 1010 402,792 13.13 Atmosphere A
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Number T U TMl WUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No. j l TMI Emergency Dose Calculation Manual (EDCM) 9 l TABLE 5.9-3 i Calculation Steps For Computing Plume Rise Step Quantity Computed Input Values Needed Source of input Valve i Inlet 1 Steam flow rate thru valve pressure of steam Plant Instrumentation i Top 2 Pressure of steam at top a. steam flow rate Step 1 of of stack, below chamfer b. internal radius of stack constant Stack, (if choked flow) c. enthalphy < constant > Below Chamfer 3 Specife volume of steam at pressure at top of stack Step 2 l to of stack, below chamfer Below 4 Temperature of steam at pressure at top of stack Step 2 ; Chamfer top of stack, below chamfer v Also 5 Velocity of steam at top of a. specific volume of steam Step 3 Jet stack, below chamfer b. flow rate of steam Step 1 Origin c. internal radius of stack constant 6 Density of steam at a. temperature of steam Step 4 jet origin (ambient pressure) b, pressure of ambient air < constant > Jet 7 Jet radius at origin a. density of steam Step 6 Origin (Needed for MIDAS only; b. velocity of steam Ste, y used in RAC but not c. flow rate of steam Step 1 really needed) 8 Plume rise a. jet radius at origin Step 7 i Plume b. density of steam at origin Step 6
- c. velocity of steam at origin Step 5 Rise d. density of ambient air < constant >
- p. ;
- e. wind speed at origin Plant Instrumentation f.150'-33' delta T Plant Instrumentation 50
s y I , NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tme Revtsion No. TMI Emergency Dose Calculation Manual (EDCM) 9 FIGURE 5.9-1 TMI-1 RB LEAK RATE VS RB PRESSURE Psc "O - I I I I L> MAX IRZ IFAK Rife N Ape.HCPli 50 th C M y E Plb=PEAX RB INTERNAL PREBURE $Jpsig) # IE5Di d BASS MUDM g,g P> ACTUAL RBINHRNAL N E j [ g 30 i O > g 2. . l 10 - l o e I 0 1 2 3 d 5 6 7 cnn ROCTOR BUR. DING LEAK RATE, Lt O 51
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Number I t TMI suct m Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 j l 5.10 Two-Pi ase Steam Flow Determination -A two-phase (liquid and gas) release calculation was included for an OTSG tube rupture accident in response to INPO SOER 83-2 (Recommendation
#12). 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 "Eniergency 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 levelinstrumentation) 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. P!950A and 951 A, or the console center, SPGA PT1 and 2 or SPGB PT 1 sad 2.
5.10.1 The model deteimines a two-phase correction factor [Tfcf] vehich is a function of OTSG pressure in psia. This factor is only calculated if the OTSG water level is indicating a liquid release (greater than 600 inches on the wide range level instrument reading). The correction factors are used to account for the radiciodine that would remain in the liquid portion of the resultant two-phase release to the environment. g\ 4
/ \ j Using the OTSG rressure in Psia, the mcdel selects a correction factor which is subsequently used with other iodine reduction factors applicable to the release pathway to correct the radioiodine source term. These other iodine reduction factors are described in Section 5.11. The algorithm used by the RAC modelto determine Tfcfis as follows:
OTSG pressure 21200 Tfcf = 2.43 1000 s OTSG pressure < 1209 Tfcf = 2.61 809 s OTSG pressure < 1000 Tfcf = 2.83 600 s OTSG pressure < 800 Tlef = 3.18 400 s OTSG pressure < 600 Tfcf = 3.77 200 s OTSG pressure < 400 Tfcf = 5.09 OTSG pressure < 200 Tfcf = 7.27
~
NOTE Increasing OTSG A/B water level will possibly help cut down the release of radioiodine due to the partitioning effect of the iodine in water. l Increasing OTSG level should be discussed with the Emergency Director as a means of reducing off-site doses. N
< )
v l l 55 i u
Number
/ U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Remon No.
TMI Emergency Dose Calculation Manual (EDCM) 9 5.11 lodine Reduction Factors Radioiodine normally exists in chemical forms which are highly reactive. They readily adsorb onto surfaces and can be scrubbed chemically from the atmosphere. The radioiodine removal methods available are specific to each release pathway and are described in Reference 7.21. The TMI RAC , Model provides the capability to include or disregard radiciodine reduction during the movement of these isotopes from the core to the atmosphere. The reduction factors (RDF) used by the TMI RAC ! Model are those described in Reference 7.21. 5.11.1 Releases frcm Condenser Offgas During an OTSG Tube Rupture For a release through the condenser off-gas the, radioiodine RDF is 0.0075. The radioiodine RDF is a product of: The fraction of radiciodine 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. This is discussed in Reference 7.17. 5.11.2 Releases from Reactor Building Design Basis Leakage Radioiodines released into the containment building are subject to airborne lodine reduction from either natural processes (e.g., gravitational settling and plateout) or from Reactor p\ Building sprays. The effectiveness of these reduction processes are dependent on the
\. holdup time of the activity in the building prior to the beginning of the release.
For Reactor Building design basis leakage where Reactor Building spray is not on, iodine reduction factors are as follows (Reference 7.21):
,H 2 24 hours RDF = 0.01 2 hours s H < 24 hours RDF = 0.04 H < 2 hours RDF = 0.4 For Reactor Building design basis leakage where Reactor Building spray is on, iodine reduction factors are as follows (Reference 7.21)-
l H 2 24 hours RDF = 0.002 2 hours s H < 24 hours RDF = 0.02 H < 2 hours RDF = 0.03 3 l 5.11.3 Direct to Atmos.ohere Releases During an OTSG Tube Rupture , lodine reduction factor for direct to atmosphere steam releases during leakage of reactor I coolant into the secondary side is 0.5. The RDF is the result of partioning in the steam generator per Reference 7.21. l
]
v , t 56 i
I 1 i Number N TMI
\ NUCLEAN Radiological Coritrols Procedure 6610-PLN-4200.02 Tdle Revision No.
l l l TMi Emergency Dose Calculation Manual (EDCM) 9 l As described in Section 5.10, if OTSG level is > 600 inches, a two phase correction factor also applies. In these conditions the RDF becomes: RDF=(0.5)(1/Tfcf) Where 0.5 = Steam generator partition factor for B&W type OTSG (Reference 7.21) l Tfcf = Two phase steam flow correction factor described in Section 5.10. 5.! 1.4 Releases from Auxiliary and Fuel Handling Buildings via Station Verst i Releases from Auxiliary and Fuel Handling Buildings via Station Vent are considered as l bypass accidents per Reference 7.21. The reduction factors for this pathway are a function I of primary system retention and filtration by the building ventitation system. The RDF for I this type of release is as follows: RDF=(0.4)(.01)=0.004 Where y 0.4 = Prime y system %ention and plateout te %. io+rer* 7.21) 0.01 = Normal reduction for effective ventilation filters (keteh < 7.216 5.11.5 Releases from Reactor B# ding via Reactor Building Purge l Releases from the peactor Buildings may occur through the purge ventilation system if the main or hydrogen purge valves are not shut. Radiciodines released into the containment building are subject to airborne iodine reduction from either natural processes (e g., gravitational settling and plateout) or from Reactor Building sprays. They are also subject to reduction by filtration through the building ventilation system if the release is through the purge exhaust valves. The RDF for this type of release, with no Reactor Puilding Spray, is as follows: RDF=(0.4)(0.01)=0.004 Where 0.4 = reduction by natural processes for minimal hold up time of 0.5 hours (Reference 7.21) 0.01 = Normal reduction for effective ventilation filters (Reference 7.21) b V 57
Nurnber b NUCLEAR TMI . Radiological Controls Procedure 6610-PLN-4200.02 ) Tdle Rewsion No. TMl Emergency Dose Calculation Manual (EDCM) 9 The RDF for this type of release, with Reactor Building Spray on, is as follows: l RDF=(0.01)(.03)=0.0003 1 Where 0.01 = Normal reduction for effective ventilation filters (Reference 7.21) 0.03 = Reduction by Reactor Building Spray for minimal hold up time of 0.5 hours ! (Reference 7.21) i The reduction factor for filtration of 0.01 applies to normal filter conditions, if the filters are known to be degraded (for example, due to moisture) but still partially functioning, this is accounted for in the RAC Model. The filters are either functioning at the full capacity,99%, or from 0 - 99% due to degradation. If the leakage is not through the purge system fillcrs, the reduction factor component for filtration is 1. 5.11.6 Releases from Fuel Handling ESF Vent System' j During a fuel handling accident in the fuel handling building, noble gases and iodines may be released into the fuel pool and subsequently into the atmosphere of the fuel handling j building. The ESF ventilation system would then exhaust the atmosphere in the building, () through filter banks, to the environment. The RDF for this type of release is as follows: RDF=(0.01)(0.01)=0.0001 Where 0.01 = Normal reduction for effective ventilation filters (Reference 7.21) 0.01 = Retention of iodines in the fuel poni water (References 7.21 and Reg. Guide 1.25) i I
'd 58
Numbir f U TMI 1 WCW Radiological Controls Procedure 6610-PLN-4200,02
% / l Title Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 1 5.12 Dispersion Model 5.12.1 Semi-Infinite Dose Model ) The TMI-1 RAC model computes both wilote body dose (TEDE) and thyroid dose (CDE) using a semi-infinite model. In the semi-infinite model, the ground is considered to be an infinitely large flat plate and the receptor is assumed to be standing at the center of a I hemispherical cloud ofinfinite radius. The radioactive cloud is limited to the space above the ground plane. In computing whole body dose, the semi-infinite plume modelis based on the assumption that the dimensions of an effluent plume are large compared to the distance that gamma rays can travelin air. If the plume dimensions are larger than the gamma ray range, then the radius of the plume might just as well be infinite since radiation emitted from beyond a certain distance will not reach the receptor. This ass"mption forms the basis for the Dose Conversion Factors (DCFs) provided in Reference 7.42 which are used by the RAC Model to calculate v. hole body dose. These DCFs combine doses from external exposure, inhalation from the plume, and exposure for an individual immersed in a plume. The whole body (TEDE) DCFs used in the TMI RAC Model are as follows: ) DCF (S i i lsotope (mfem/hr)( Ci/cc) Kr 85m 9.30E+04 Kr 85 1.30E+03 Kr 87 5.10E+05 Kr 88 1.30E+06 Xe 131m 4.90E+03 Xe 133m 1.70E+04 Xe 133 2.00E+04 l Xe 135m 2.50E+05 Xe 135 1.40E+05 Xe 138 7.10E+05 l131 5.30E+07 1132 4.90E+06 1133 1.50E+07 1134 3.10E+06 1135 8.10E+06 3 V 59
1 i Number t h U TMI ( NUCLEAR Radiological Controls Procedure 6610-PLN-42's0,02 ' Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 The TMI 1 RAC model calculates the thyroid dose rate due to inha'ation of l-131,1-132, l-133,1-134, and 1-135 in the hemispherical cloud. The thyroid dcse rate is proportional to X/Q. The thyroid dose conversica factors are calculated usiry the child breathing rato of i 0.42 m /hr from Table E-5 of Reference 7.35 and the child inhalation dose factors from ! Table E-9 of Reference 7.35. The dose is computed by multiplying the dose rate by the expected duration of release. The thyroid DCFs use'lin the TMI RAC Model are as follows: DCF lsotope (mr3m/hr)(pCi/cc) I131 1.84E+09 l l132 2.21 E+07 l133 4.38E+08 l l134 5.76E+06 l l135 0.00E+07 , 1 The radioiodines are ri ecayed during plume travel time. The decay constants for 1-131 through I-135 are fram the Radiolooical Health Han@qq. h. They are as follows: I Decay Constant g Isotope (min) Kr 85m (V) Kr 85 2.58E-03 1.23E-07 Kr 87 9.12E-03 Kr 88 4.04E-03 Xe 131m 4.09E 05 Xe 133m 2.20E-04 Xe 133 9.18E-05 Xe 135m 4.43E-02 Xe 135 1.27E-03 Xe 138 4.89E-02 1131 5.99E-05 l132 5.06E-03 1133 5.53E-04 1134 1.32E-02 1135 1.75E-03 5.12.2 Calculations Used in the Model i 5.12.2.1 X/Q Calculation: The basis forthe X/Q ca!culation is the Gaussian diffusion equation and a 10 x 7 l array of sigma y's and sigma z's. The array of values cenesponci to sigma y's and sigma z's for 7 stability classes and at 10 fixed downwind distances For l distances other than the fixed downwind distances, the sigma y's and sigma Zh are linearly interpo!ated before X/Q is computed for that distance. The ten f;xed distances are: 200, 500,1000, 2000, 3000, 6000,10000, 30000, 50000, and 80000 meters. d I 60
Number ( U TMl Q) NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Trtle Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 t 5.12.2.2 Compute Buildina Effect Returns one of seven pre-computed virtual source distances, depending on , stability class. The virtual source distances for each of the seven stability l 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 j building. Building wake effects are simulated by adding the virtual source ! distance, for a particular stability class, to the actual downwind distance for the purpose of computing X/Q, For example, suppose we wanted to know X/Q , without building wake effects at 800 meters downwind with stability class D. X/Q would then be computed at 800 meters downwind. With building wake effects, i X/Q would be computed at 1108 meters downwind (800 + 308) Thus building 1 wake effect is simulated by computing X/Q at a distance greater than the actual l downwind distance and is called only for ground level portion of release. 5.12.3 Other Calculations I l 5.12.3.1 Compute TMI-1 Emergency Action Level. Declares the emergency action level i from highest dose whether whole body (TEDE) or thyroid (CDE). l p 5.12.3.2 Compute Site Boundarv
\
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 distance is in meters. l l I l l I O 1 61 I
, N TMI
() NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.12.3.3 Compute Terrain Factor Computes terrain height in meters for a given downwind distance. At downwind ; distances other than the given distances, 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. 5.12.3.4 Compute Stability Class As measured by the TMl 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 temperature difference per 100 ft is shown in the last column of the table. Stability class is determined by the measured temperature lapse rate per Reg. Guide 1.21. 5.12.3.5 Adiust Wind Spee_d Adjusts wind speed from the anemometer height to the release height. The wind speed is adjusted according to the following equation: fm
\j\ u=u(h/ho)"
o 1 where the subscript "0" denotes the anemometer height and "u" and "h" are the wind speed and height above ground, respectively. The exponent p is a function i of stability: 0.25,0.33 and 0.50 for unstable, neutral and stable cases, respectively, if the adjusted wind speed is less than 0.5 mph, the adjusted wind speed is set equal to 0.5 mph. 5.12.3.6 Compute Exit Velocity Computes exit velocity of the released materialin feet per second by dividing l cubic feet per minute by the stack cross-sectional area. 5.12.3.7 Compute Plume Rise Computes the plume rise in meters for the elevated portion of a spilt wake release. Two formulas are used to calculate the plume rise; for enstable and neutral i conditions jet plume rise, momentum dominated, is calculated from Briggs' l Plume Rise, Eq. 4.33; for stable stability, it is calculated using Eq. 4.28 from Briggs' Plume Rise. oO 62
I Number h) U TMI l () T,tle NUCLEAR Radiological Controls Procedure 6610-PLN-4200 02 Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 5.12.3.8 Compute Entrainment Factor Computes entrainment factor for wake split flows. A mixed mode release is I assumed when: (1) the release point is at the level of or above adjacent solid structures but lower than elevated release points, (2) the ratio of plume exit velocity to horizontal wind speed is between one and five. Specifically, the entrainment factor, Ei, is computed according to the following formulas: Ei = 1.0 for w c/u le 1 Ei = 2.58 - 1.58(wdu) for 1 gt wdu le 1.5 Ei = 0.30 - 0.06(wdu) for 1.5 gt wdu le 5.0 Ei = 0 for wdu gt 5.0 where wois the stack gas exit velocity and u is the wind speed at stack height in miles per hour. Note that the entrainment factor does not address the case of two adjacent plumes mixing with each other, as would be the case in TMI-1, where it is possible for a clean plume and a contaminated plume to be emitted from adjacent ( / but separate stacks. These plumes are examples of co-located adjacent jets; little is known about the modeling of co-located adjacent jets such as the ones at TMI-1. (O) m 63
l N wnbst U TMI , NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdte Revson No. TMI Emergency Dose Calculation Manual (EDCM) 9 TABLE 5.121 Classification of Atmospheric St..bility Stability Pasquill Delta T Delta T Classification Cateoories (150'- 33') ('F/100j M G Extremely Unstable A < -1.22 < -1.04 Moderately Unstable B 2 -1.22to< -1.09 2 -1.04to< -0.93 Slightly Unstable C 2 -1.09to< -0.96 2 -0.93to< -0.82 j Neutral D 2 -0.96 toc -0.32 2 -0.82to -0.27 ) Slightly Stable E 2 -0.3210< +0.96 2 -0.27to< +0.82 Moderately Stable F 2 +0.96to< +2.56 2 +0.8210< +2.19 I Extremely Stable G > +2.56 > +2.19 O 64
l Number O TMI l
/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 5.13 Liauid Release Calculation - In this section of the model calculations are performed for liquid source term determination, (see Figure 5.13-1 and Tables 5.13-1,5.13-2) concentrations in the river, travel time to downstieam users, and ingestion dose commitment calculations. The methods used to perform the calculations are as follows: e The concentrations of the liquid effluents are deterrr. '.ad by samphng and analysis of the effluent stream. If sample results are not ir 'ncOi.9ely available, previous Chemistry data may be used to provide a best esumaw uf the effluent concentrations. Only the four usual iodine isotopes (1-131,1-132,1133,1-135), H-3, Cs-134, Cs-137, Co-58 and Co-60 are considered in the calculation. 9 The dilution in the river is calculated by first obtaining the river flow rate and inputting the value in the model. The flow rate may be obtained by the following methods: Calling the York Haven Dam at 266-3654 or, l Oalling the River Forecast Center at 1-814-234-9861 The river flow rate is then used along with the discharge flow rate to calculate the concentration in the river. Complete mixing is assumed by the time it reaches the I first drinking water station. The concentration in the riveris then divided by the l (j effluent concentrations in 10 CFR 20 Appendix B Table 2, Column 2, to determine the fraction of these concentrations in the river to downstream users. O The river concentration is used along with the total discharge time, to calculate the dose commitment to an individual from drinking the 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 (Reg. Guide 1.109), and the infant usage factor (330 liters /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 the predominant isotopes. O A fiume arrival time is estimated for each known downstream user. The river l volume flow is used in an algorithm based on a model derived from river dye dilution and flow studies conducted from the TMI discharge. The algorithm uses the following equation: l Time (hrs) = Distance to Downstream User (Miles) [ River flow (cfs) X .0283 (m'/ft ))
- X 0.019667
, 1 l 9 If the concentration of any nuclides in the river exceeds the concentrations l
- specified in 10 CFR 20 Appendix B, Table 2, Column 2, downstream users must j
be informed and recommended to curtail usage. Refer to Procedure 1203-44 for downstream users and telephone numbers. lNm) 65 I
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Number b\ j t TMI g NUCLEAR Radiological Controls Procedure 6610-PLN 4200,02 l Tdle Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 TABLE 5.13-1 TMI-2 Sump Capacity Sump Topl Capacity Gallons / Inch Turbine Building Sump 1346 gals 22.43 j Circulating Water Pump House Sump 572 gals 10.59
]
Control Building Area Sump 718 gals 9.96 Tendon Access Gallery Sump 538 gals 9.96 Control to Service Building Sump 1346 gals 22.43 Emerger.cy Diesel Generator Sump A/B wet 837 gals 9.96 A/B dry 1200 gals 14.29
, Chlorinator House Sump --- ---
v Water Treatment Sump 1615 gals 22.43 Air Intake Tunnel Normal Sump 700 gals --- Emergency Sump 100,000 gals 766.00 Condensate Polisher Sump 2617 gals 62.31 l Sludge Collection Sump 1006 gals 26.33 Heater Drain Sump --- --- Solid Waste Staging Facility Sump 1476 gals 24. Auxiliary Building 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 l
A {G l 67
Number i U NUCLEAR TMI Radio;ogical Controls Procedure 6610 PLN-4200.02 Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 TABLE 5.13-2 TMI 1 Sump / Tank Capacities Sump Capacity (Gallons) Turbine Building Sump (TBS) 10,000 Auxiliary Building Sump (ABS) 10,000 t Reactor Building Sump (RBS) 10,000 Intermediate Building Sump West 1,000 Tendon Access Gallery Sump 1,000 I intermediate Building Sump East 1,000 < Auxiliary Boiler Sump 2,000 t Powdex Sump 40,000 t (_ Industrial Waste Treatment System Sump (IWTS) 300,000 l Industrial Waste Treatment System Sump (lWFS) 80,000 l l Tanks l I TMI-2 Condensate B Tank 250,000 TMI-1 OTSG A or B (secondary) 25,000 l TMI-1 WECST A or B 8,000 ! Neutralizer Tank 100,000 BWST 350,000 Condensate Tank A/B 265,000 D b l 88 L.
Numb 1 b i U NUCLEAR TMI Rndiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 5.14 Off-site Air Sample Analysis 5.14.1 Introduction The results provided from field teams can be used 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 iodine cartridge are called into the RAC or EACC from the field teams. The RAC or EACC staff then can estimate the off-site 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 Ebertine E140N with a HP-210/260 type probe to be used for counting in the field. The results of several tests on combinations of different probes and ratemeters showed a consistent 0.0039 (0.39%) counting efficiency. (Reference 7.7,7.10,7.11). Since 1-131 has a fairly strong beta (0.6 MeV max.), the usual particula'e filter counting efficiency n of 0.1 (10%)is used. The collection efficiency for both filters for these calculations are .I ()\ 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 off-site dose commitment. Since the plant RAC model 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 1131. 5.14.4 Example: Given an off-site 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 = 1 hour DRCF = 4.0E8 mrem /hr Ci/cc b 69
l Numbir i O U TMI
) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
l TMl Emergency Dose Calculation Manual (EDCM) 9 l The RAC model follows the logic below to calculate an off-site thyroid dose commitment for this sample,
- a. net particulate countrate = 200 - 100 = 100 cpm
- b. net particulate activity = 100/.1 = 1000 dpm i I
- c. net cartridge count rate = 200 - 100 = 100 cpm
- d. net cartridge activity = 100/0.0039 = 25600 dpm
- e. total activity in sample = 1000 + 25600 = 26600 dpm I
- f. total microcuries = 26600/2.22E6 = 0.012 Ci
- g. sample volume = 50
- 1000
- 10 = SES cc
- h. air concentration = 0.012/SE6 = 2.4E-8 Ci/cc
- i. dose commitment = 2.4E-8 Ci/cc*1 hour *4E8 mrem /hr = 9.6 mrem Ci/cc U 5.15 Protective Action Recommendation Loaic - The Logic Diagram is designed to enable the user to l develop protective actions based primarily upon declaration of a General Emergency and also taking l into account plant conddions, release duration and dose assessments. The logic is diagramed in l Procedure EPIP-TMI-27, and EPIP-TMI .02.
l i O 70
Number U ()i Tme NUCLEAM TMI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 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 has 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, and evaluation of the EDCM and to assist with the user interface of the RAC program software. Emergency Preparedness is responsible for ensuring that the EDCM and the RAC computer models are current and compatible. 6.5 The Technical Support Center {TSC) has the responsibility to provide the RAC with appropriate fuel damage data and other information pertinent to performing dose calculations. 6.6 The Emergency Preparedness Department and Radiological Engineering have the responsibility to ensure any changes to the RAC program are performed and documented IAW Procedure 1000-ADM-1230.10. 'V ) 6.7 The Radiological Controls and Emergency Planning Departments have the responsibility to make any changes to the RAC Code, and to compile and distribute new versions of the program.
7.0 REFERENCES
7.1 American National Standard (ANS), ANSI /ANS-18.1 1984, Radioactive Source Term for Normal l Operations of Light Water Reactors l 7.2 APS Source Term Report - Report to the American Physical Society of the Study Group on Radionuclide Releases From Severe Accidents at Nuclear Power Plants, February 1985 7.3 Dose Assessment Manual for Emergency Preparedness Coordinators, February 1986, INPO 86-008 7.4 Efficiency Check using an Air 1131 Source Cartridge and a Ba-133 Source Cartridge, Memorandum 9502-88-0139, September 28,1988 7.5 EPA 520/175-001 - Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 7.6 EPIP-TMI .07 - Off site /On-site Dose Projections 7.7 Evaluation of a Front Loaded lodine Cartridge using Various Survey Equipment, Memorandum 9100-88-0194, May 12,1988 7.8 Field Measurements of Airbome Releases of Radioactive Material, Memorandum 9502-88-0098, May 25,1988 I 7.9 FSAR, TMI-1 Chapter 11, Radioactive Waste and Radiation Protection &l 7.10 FSAR, TMI-1 Chapter 14 - Safety Analysis 71
p Number ! [ U TMI g NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 7.11 TMI 2 PDMS Safety Analysis Report l 7.12 TMI Emergency Plan,1000-PLN-1300.01 7.13 ICRP Report of the Task Group on Reference Man,1981 7.14 INPO SOER 83-2 " Steam Generator Tube Ruptures" 7.15 Introduction to Hea:th Physics, Herman Cember,2nd Edition,1985 7.16 NRC-BNL Source Term Report i 1 7.17 NUREG-0017 PB-251718 - PWR - GALE Code; Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWR, Revision 1,1985 7.18 NUREG-0133 - Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, October 1978. 7.19 NUREG-0654 - Revision 1 - Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants I g < 7.20 NUREG-0737 - Clarification of TMI Action Plant Requirements, U.S. Nuclear Regulatory I d Commission, November 1980, Generic Letter 82-33, Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability, U.S. Nuclear Regulatory Commission, Washington, D.C., December 1982 7.21 NUREG-1228 - Source Term Estimation during incident Response to Severe Nuclear Power P ant Accidents, October 1988 7.22 NUREG/CR-3011 - Dose Projection Considerations for Emergency Conditions at Nuclear Power Plants 7.23 N1830 - Post Accident Reactor Coolant Sampling 7.24 N1831 - Post Accident Atmospheric Sampling 7.25 N1832 - Post Accident Sample Analysis j l 7.26 Deleted 7.27 1210 Abnormal Transients Rules, Guides and Graphs 7.28 1210 RCS Super Heated 7.29 1202 Excessive Radiation Levels 7.30 Operational Qunlity Assurance Plan,1000-PLN-7200.01 7.31 Proprietary Midas User Documentation, Pickard, Lowe, and Garrick i V ! 7.32 Radioactive Decay Data Tables, David C. Kocher, ORNL, DOE / TIC-11026,1981 72
r l l Number O OU TMI l ( NUCLEAR Radiological Controls Procedure o610-PLN-4200.02 TM Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 I l 7.33 Radiological Health Handbook, Revised Edition Jan.1970, US Dept. HEW q j 7.34 Reg. Guide 1.21 Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases j of Radioactive Materials in Liquid and Gaseous Effluents from Light Water Cooled Nuclear Power Plants, Rev.1, June 1974. 7.35 Re0. Guide 1.109 - Calculation of Annual Doses to Man Frnm Routine Releases of Reactor ! E Nents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix 1, October 1977, Rev.1 7.36 SER-419626-003. Rev. 7, Instrument Calibration Facility, Feb. 12,1988 7.37 TDR-390 - TMI-1; Primary-to-Secondary OTSG Leakage and its On-site /Off-site Radiological Impact, April 1983 7.38 TDR-405 - TMI-1; Evaluation of Plant Radiation Release and its 10CFR50, Appendix I Conformance , for Different Operating Conditions i 7.39 TDR-431 - Method for Estimating Extent of Core Damage Under Severe Accident Conditions 7.40 WASH 1400 - 1975 Nuclear Safety Study WASH-1400 (also known as Rasmussen Report) k 7.41 Sutron Report No. SCR-358-82-063, Study of Travel Time and Mixing Characteristics for the j Susquehanna River Below Three Mile Island ! I 7.42 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 7.43 RAF 6612 97-019, " Source Term Algorith for RMG-22/23 7.44 RAF 3640-98-19, " Documentation of Calculation of Direct-to-Atmosphere Mass Flow Rates" 8.0 EXHIBITS Exhibit 1 Emergency Dose Assessment User's Manual 1 f o ( l v 73
i NumbIr i [ T U TMI -
! NUCLEAR l Radiological Controls Procedure 6610-PLN4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 INDEX Section Eagg a 1 Overview of the Emergency information Network 75 l l Emergency information Network Block Diagram 77 l Emergency information System Main Menu 80 ED/ESD Screen 81 l i RAC Screen 82 l Parameter Edits Screen 83 Sample Types Screen 85 Effluent Sample Screen 87
]
Midas Plume Plots 88
,D '
k Field Team Air Sample Calculations 90 Offsite Field Monitoring Team Data 91 Reuter Stokes Data and MIDAS Calculation Comparison 92 individual Pathway Data Screens 93 Effluent RMS Screen 97 Area Gamma and Liquid RMS Screen 98 Emergency Action Level Screen 99 Quick Code 100 Manual Codes 101 If the LAN is Down 102 Instructions for Rebooting the COLA 103 Basic RAC Checkoff 104 Indications of Mechanical Fuel Damage 105 g EDCM Calculation Guides 106 I i D'Pimary to-Secondary Leak Rate Based on COG Monitors 228 74
Number U TMI NUCLEAR Radiological Controis Procedure
\ 6610-PLN-4200.02 Tfue Revemon No.
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 1 of 164 s Overview of the Emergency Information Network The Emergency information Network (EIN) at TMl is a LAN based system that allows emergency organization personnel to view information on plant conditions and offsite dose projections during a plait emergency. Because it is a LAN based system, the same information can be viewed from any of the majo* emergency facilities. The ability of the differert facilities to share ommon data during an emergency enhances communications within the emergency team, and n'inimizes the potential for conflicting hformation. The heart of the EIN is the Continuous Online Assessment System or COLA. The COLA produces a ne.v dose projection every 4 to 6 minutes and displays it on the EIN. This is accomplished by a central " Host Computer" retrieving Met Data, RMS readings, RCS Activity, lodine Spiking Factors, and plant information such as pressures, temperatures, flowrates etc from the plant's computer system. This information is used to produce a separate dose calculation for each cf TMI's release pathways that are then added together for a complete dose projection from TMI. The dose projection is then sent to the Local Area Network (LAN) and is available for review by Emergency Management Personnel. Since the host computer automatically runs dose projectior s every 4 to 6 minutes, the RAC cannot produce a dose projection any time he/s?.e wishes using the COLA. In addition to the COLA, the EIN also provides the capability to run manual versions of the code. The Manual Codes are located on the hard drives of the dose projection computers in the emergancy response facilities and are f independent of the LAN system. As a result, they can be used if the LAN system is not operational for some reason. ( Because the manual codes are independent of the LAN system, the results are NOT displayed on the network. Therefore, information that is important to other personnel such as the Group Leader R&EC must be faxed to them. The two versions of the manual code available on the EIN menu are: 9 Manual Code - used to perform dose projections out to 30 miles. IREO personnel performing manual dose projections would use this code. 9 Quick Code - a simplified manual code that would be used by the GRCS during the first hour if the COLA is not functional. Only an RMS reading and met data are required for entry. The Manual Codes can be used for some of the following reasons: O To perform a dose projection if the COLA is out of service e To perform dose projections between COLA dose projections 9 To perform a "What If" dose projection. 9 To perform a dose projection if the data being retrieved by the COLA is considered inaccurate (i.e. a bad RMG-25 reading). O To perform a dose projection if the release is not properly monitored by an RMS monitor. e To compare or validate the COLA dose projection for any other reason. A much more detailed description of Manual Code calculations is discussed later in this manual The third dose projection component of the EIN is the MIDAS Plume Plot System. The COLA calculated source term leaving the plant is automatically imported into the M!DAS program, which produces plume plots. MIDAS calculates dose projections differently than the RAC codes. The RAC code is a straight-line gaussian model. That 75
Number [)/ U NUCLEAR TMl Rcdiological Controls Procedure 6610-PLN-4200.02 Trle Revson No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 2 of 164 means it takes the source term leaving the plant at that moment and calculates the dose based on the most current met conditions. It ignores releases of radioactivity that occurred prior to the most current dose projection. It projects the doses that will be seen offsite if things do not change. The MIDAS code is a particle in cell model. That means it models the offsite doses by tracking the activity released every 15 minutes and coupling that activity with the met data for that 15 minutes. As a result, it more closely approximates dose rates that should currently be seen offsite. As a result, the MIDAS Plume Plots are used for comparison with field team and Reuter Stokes data. The EAC compares field team data it receives with the MIDAS Plume Plots. In addition, the EIN provides an option which displays a comparison of MIDAS projected doses with actual Reuter Stoke 3 monitor readings. This comparison is used to check that the source term produced by the COLA is reasonable based on what we are seeing in the field. The MIDAS Program can be run off the "C" drive by EACC personnel with manualinput of source terms when the COLA is out of service. Finally, the EIN also displays effluent RMS and plant area monitor readings. It also provides a listing of the parameter values that were used in calculating the dose projection for each pathway. A schematic illustration of the EIN appears on the next page. t \ U O V 78
I l Number
/D ' & NUCLEAR TMl Radiological Contrels Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 3 of 164 Emergency infonnation Network Block Diagram SAMPLE EDITS: 1 Met PPM RCS Activity PARAMETER EDITS: ) LAN Iodine \NG Spiking Factors Files Data Data MSR Release ' Inplant Sample Results From Release Duration 10 Release Pathways Fuel Da,nage Class l Met Data j 1r
' l y Host Computer
+
PRODUCED A EVERY ( 4-6 Mrs. COLA Dose Projection
\.
9r 1r 1r 1r 1r Output ED/ESD Screen: RAC Screen: Individual Pathway Data includes: eft 1uent Area 3 ,, '*, Decision Screen Downwind Doses Source Term + Inputs + Dose Data RMS RMS 1r Projected , m MIDAS l>lume Plots Reuter Stokrs Updates every 15 minutes Values 1r 1r
'I Gamma Plot Thyroid Plot Comparison of actual / projected 1r 5r Reuter Stokes values Gamma - mR/hr nyroid Rate - mR/hr Open Window - mR/hr Cartridge - cpm Frisker cpm Filter -cpm Compare Field Team Readings to MIDAS Plume Plots to verify
[m Dose Projection 4 " k~ 77
f Number I ' U TMl l NUCLEAR Radioiogical Controls Procedure
/ 6610-PLN-4200.02 i
~
Tme Revisson No. l l TMI Eniergency Dose Calculation Manual (EDCM) 9 l 1 EXHIBIT 1 Page 4 of 164 ) Emergency Information Network Block Diagram Description The Host Computer, located on the 1" f'oor of the OSF, produces the COLA dose projection screens by retrieving a lot of different information that resides on the LAN. This information ir.Judes-4 9 Plant Performance Monitor (PPM) Data -includes all of the data residing on the PPM such as RMS data, Condenser flow rates, Condenser vacuum, RCS temperature / pressure, OTSG pressures, OTSG levels etc. i 9 Met Data - Retrieves wind speed, wind direction, delta temperature directly from the Met Tower. Updated every 15 minutes. Sample Edit Screen 9 RCS Activny - Retrieves the default RCS Activity entered into the Sample Edit screen by Rad Engineering or the edited RCS Activity entered by RAC/ Rad. Eng. Support personnel into the Sample Edit screen. g e lodine /NG Spiking Factors - Retrieves the default Spiking Factors entered into the Sample Edit
- g 4 screen by Rad Engineering or the edited Spiking Factors entered by RAC/ Rad. Eng. Support l
L ,/ personnelinto the Sample Edit screen. The Nuclear Engineering Group determines the default spiking factors. O In-plant Sample Results - The sample results from TMI 1's 10 release pathways may be entered into this screen by RAC/ Rad. Eng. Support personnel. Examples are MAP - 5/ sample panel results for RMA-5,8,9, effluent pathways and CATPASS samples for the Design Basis Leakage pathway. Parameter Edit Screen 9 MSR Release - The default MSR position is closed but should a release occur through this pathway an open MSR position can be indicated by using this screen. O Release Duration - 8 hours is the default release duration but can be edited by RAC/ Rad. Eng. Support personnel. 9 Fuel Damage Class - The COLA automatically determines the Damage Class from the RCS
- Temperature (average of the 5 highest incores). If plant conditions warrant it, rRAC/ Rad. Eng.
l Support personnel can edit the Damage Class. O Met Data -If the COLA is retrieving inaccurate Met Data from the Met Tower, this screen can be used to input the correct information. The COL / . after retrieving all of this information, calculates a new dose projection every 4 to 6 minutes and displays the associated screens on the LAN. The only information that is not updated within the 4 to 6 minute run is the Met Date, which is updated every 15 minutes. All of the information on the Emergency Information Network is retrieved by the Host Computer and displayed on the LAN. The COLA dose projection includes the following screens: (V/9 O ED/ESD Screen RAC Screen e Individual pathway Data 78
Number b U TMI ( NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 5 of 164 0 Parameter Edit Screen
-e Sample Edit Screen The Host Computer then uses the source term from the dose projection to produce a new MIDAS Plume Plot every 15 minutes. The MIDAS program produces a Beta / Gamma plot and a Thyroid Plot. These plots are produced in three different map sizes, island view,2.5 mile view, and 6 mile view. The primary purpose of MIDAS is to verify the dose projection.
The MIDAS program then produces the projected Reuter Stokes values again every 15 minutes. These projected Reuter Stokes values can be compared to the actual Reuter Stokes values as another confirmation of the dose projection. This is possible since MIDAS uses the same source term as the COLA dose projection and then produces the Reuter Stokes projected values. O
%-) !
l k-79 ~.
I Nurnber N TMl I, NUCLEAR 1 Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 l l EXHIBIT 1 Page 6 of 164 l Emergency Information System Main Menu
- (Typical) i EMERGENCY INFORMATION NETWORK 1 A. ED/ESD SCREEN 18. RAC SCREEN 1C. PARAMETER EDITS 1D. SAMPLE EDITS 2A. MIDAS PLUME PLOTS 28. FIELD TEAM DATA 2C. AIR SAMPLE CALC 2D MIDAS Plume Plot Edit (EACC Only)
- 3. REUTER STOKES FIELD DATA AND MIDAS CALCULATION COMPARISON
- 4. INDIVIDUAL PATHWAY DATA
- SOURCE TERM
- INPUT DATA
- DOSE DATA "N
- 5. EFFLUENT RMS
- 6. AREA GAMMA AND LIQUID RMS
- 7. DELETED
- 8. QUICK CODE
- 9. MANUAL CODE
- 10. EXIT TO DOS ENTER THE APPLICATION NUMBER:
This is tha main menu for the Emergency Information Network. Each selection will be discussed in detail in the subsequent pages. ( V 80 l \ l
r 1
\
Number
/ U TMI \ , NUCLEAR Radiological Controls Procedure 6610-PLN 4200.02 TXIe Reymon No. I TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 7 of 164 EDIESD Screen (1 A)
(Typical) ACTIONS TO BE TAKEN BASED UPON THIS DOSE CALCULATION: Declare an ALERT per A1.1 (5.40E+01) mrem /Hr CDE) Do NOTissue A Protective Action Recommendation Presently at Most Limiting Dose (Rate) Dose Type Location Percent of Next Limit 1.71E+01 mrem TEDE 1 Mile 17% 4.32E+02 mrem CDE 1 Mile 36 % RELEASE PATHWAY CONTRIBUTION TO MOST LIMITING DOSE using CDE Dose CONDENSER OFFGAS 0.0% B OTSG ADV 100.0 % l RX BLDG DESIGN BASIS LEAKAGE 0.0% B OTSG MSR 0.0% m EMERGENCY FEED PUMP EXHAUST 0.0% STATION VENTILATION 0.0% f) A OTSG ADV 0.0% RX BLDG PURGE EXHAUST 0.0% V A OTSG MSR 0.0% ESF EXHAUST 0.0% WIND SPEED (MPH): 5.0 WIND DIRECTION (DEG FROM): 90 DELTA TEMP (DEG): -1.00 STABILITY CLASS (A TO G): C RELEASE DURATION (HRS): 8 NRC CLASS (1-10): 1 CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT)
- 1. The ED/ESD Screen provides the Radiological Emergency Action Level (EAL) for the ever vraight out of the Emergency Classification and Basis procedure, it also ptovides the most limiting dose s ' ;ose rate that required entry ireto the EAL.
- 2. Prompts a Protective Action Recommendation (PAR)if a General Emergency is declared. If a Site Area Emergency is declared, the message will prompt RAC personnel to review the PAR Logic Diagram.
- 3. Most Limiting Dose (Rate) provides the highest integrated Total Whole Body Dose / Dose Rate (TEDE) or Child Thyroid Dose / Dose Rate (CDE) based on the latest dose projection. The Most Limiting Dose (Rate)is the summation of all 10 of TMI's release pathways and shows the downwind location of the highest dose or dose rate. Dose rates are shown for dose projections requiring less than an Alert classification. Integrated dose is displayed for dose projections requiring an Alert classification or above (i.e., approaching a Site Area or General Emergency).
- 4. Presently at Percent of Next Limit provides the user with an indication of how close the dose projection is to approaching the next emergency classification level that would require escalation of the event. The most limiting type of dose is highlighted.
l f g 5. Displays TMI's 10 release pathways in % of dose contribution. The pathway with the highest % doses will be highlighted on the screen. The screen also displays which type of dose (TEDE or CDE)is being used to (V ) calculate the percentages. 81
Number ( U TMI
) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 8 of 164
- 6. Provides Met Date: Wind Speed, Wind Direction (degrees from), Delta T, and Stability Class (A to G). This information is updated every 15 minutes from the Met Tower. The Met Data is also the average of all of the readings taken over 15 minutes from the Met Tower. If the Met Tower stops transmitting to the LAN, inaccurate Met Data will display on this screen and will be used by the COLA. Editing of the Met Data is then required and will be discussed in the " Parameter Edits" screen section.
- 7. Provides Release Duration default (8 hours). This parameter can be edited and will also be discussed in a
the " Parameter Edits screen section. Protective Action Guidelines (PAGs) are based on the integrated dose at the site boundary therefore a conservative release duration would cause a conservative dose projection.
- 8. Fuel Damage Class is calculated automatically by the COLA using PPM data that uses the average of the 5 highest incore detectors. If the calculated Fuel Damage Class is not accurate based on plant conditions
( i.e. mechanical fuel damage would not affect the incore detectors) editing can be performed . Thb will also be discussed in the " Parameter Edits" section. RAC Screen (1B) l (Typical) l g _ t, j TOTAL DOSE FROM ALL PATHWAYS IN MREM v TIME: 07:23:18 DATE: 01-13-1998 CALCULATION #: 1 DISTANCE COMMITTED DOSE EQ (MREM) TOTAL EFFECTIVE DOSE EQ (MREM) Site Boun. 1.SE-01 1.0E-02 Ex. Area 1.0E-01 6.5E 03 1 MILE 3.5E-02 2.6E-03 2 MILE 1.4E-02 1.2E-03 5 MILE 3.4E-03 2.6E-04 10 MILE 1.4E-03 1.0E-04 HIGHLIGHTED DOSES ARE MAXIMUM VALUES CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT H (HARDOOPY) HOME (EXIT) The RAC Screen lists the " Total Dose from All 10 Pathways. This screen displays the Committed Dose Equivalent CDE (Child Thyroid) and Total Effective Dose Equivalent TEDE in ranges from the Site Boundary out to 10 miles. If RAC personnel suspect that significant doses are present outside of 10 miles, then the Manual RAC Code must be used since it can calculate doses beyond 10 miles. The downwind location with the most limiting dose is highlighted on the screen. O f l 82
Mumbir e [\ U TMI l
) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 1itle Reveston No.
TMl Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 9 of 164 l Parameter Edits Screen (1C) SELECT PARAMETER EDITS Minimum: 0 Default: 0 Maximum: 1 PARAMETER EDITS PARAMETER NAME PARAMETER VALUE ACTIVE EDIT MSR RELEASE O (1 IS OPEN) 0 (1 IS ACTIVE) RELEASE DURATION 8 (0-99 HOURS) 0 (1 IS ACTIVE) DAMAGE CLASS 1 (CLASS 1 10) 0 (1 IS ACTIVE) WIND SPEED 4 (2-99 MPH) 0 (1 IS ACTIVE) WIND DIRECTION 210 (0-360 DEG) 0 (1 IS ACTIVE) n DELTA TEMP -1.1 (-10 TO +10 DEG) 0 (1 IS ACTIVE) F1: Next F2: Prev Alt F10: Abort F10: Done
- 1. MSR Release -This option allows RAC personnel to indicate to the COLA whether a Main Steam Relief Valve is open. A "1" entered in the parameter value indicates that a valve is open and a *0" indicates a closed position. The COLA will then calculate the flow rate based on the main steam pressure obtained from the PPM. This option is only accurate if ONE MSRV is open. In the unlikely event that more than one valve is stuck open the Manual RAC Code must then be used. The Manual Code allows inputs for the position of all of the MSRVs and ADVs. Also the CCLA always uses RMG-26 or RMG -27. If this RMS monitor is isolated, which occurs when MSV 2A or 2B are closed, then the Manual Code (Contingency Calculation) must be used. An operator must be sent up to the Intermediate Building roof to visually check the valve to verify whether a MSRV is really open. This requirement is incorporated into plant procedures.
- 2. Release Duration -This edit option is probably the one most often changed by the RAC. Eight hours is the default release duration used by the COLA. Anytime you want to enter another release duration other than 8 hours simply select a number from 1-99 and enter it in the Parameter Value. Since this option has a direct effect on the integrated dose released from TMI, the number used must always come only from the Emergency Director. If you suspect the ED is giving you a conservative valt for rel. ase duration question him on it.
- 3. Fuel Damage Class -This very important parameter is calculated by the COLA, udng the average temperature of the 5 highest incore detectors. Howeverif you don 1 think this ce!. .ation is accurate, the parameter can be edited by selecting 1 through 10 and entering it as the Parameter Value. One example of the need to edit would occur if the core suffered mechanical damage unrelated to temperature and pressure.
A Damage Class of 2,3, or 4 would then be entered as the Parameter Value. [O
\j Keep in mind that once core damage occurs it can never be repaired. The COLA calculated damage class may decrease based on the cooling of the core but the previously higher Damage C! ass must continue to be used by using the editing option.
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I l Numbrr ! U TMl NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 , l Title Revision No. 1 1 TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 10 of 164 l The purpose of the Fuel Damage Class is to determine what percentage each of the 10 noble gas and 5 iodine isotopes contribute to the total RCS activity. The tatsl RCS activity is not important since the activity ; leaving the plant is being dvectly measured by an RMS monitor for all COLA calculations. If the COLA uses l damage class 1, then the current RCS activity that was entered on the LAN or the edited RCS Activity will be used. Il Damage Class 2 though 10 is either calculated by COLA or manually edited, then the RCS activities listed in the EDCM for each Damage Class will be used in the calculation of the source term. Also the noble i gas and iodine Spiking factors are only used when the Damage Class is 1. Damage Class 1 with a spike is 1 considered to be Damage Class 1 A. J Also the Damage Class can determine how much offsite CDE Child Thyroid Dose will occur as compared to the offsite TEDE Whole Body Dose, i.e.,if the Damage Class is 1 though 4 the NG to iodine ratio is approximately 5 to 1. When the Damage Class increases to 5 though 10 the NG to iodine ratio changes drastically to 0.63. Therefore a increase to Damage Class 5 or above will cause a significant increase in the Child Thyroid Dose as compared to the TEDE Whole Body Dose
- 4. Wind Speed - Enter 1 - 93 as the Parameter Value Wind Direction - Enter 0 - 360 as the parameter value Delta Temp - Enter + 10 as the parameter value Bear in mind that the COLA will retrieve Met Data from the Met Tower whether the data is accurate or not.
g } This could occur if the Met Tower goes out of service (O's would be displayed by the COLA). This also would render the weather charts in the Control Room inaccurate. If this occur contact the EACC for the correct Met Data to input in the Parameter Edits. Another problem might be that the met tower is operational, but has stepped commanicating with the host computer. In this case the host computer will continue to use the last met data it received from the tower. The user should be alert to the fact that met data that does not change at all over the course of 30 to 45 minutes may indicate that the communication has broken down. Control room instrumentation should be cheked to venfy that the met data is accurate. Active or inactive Parameter Edits - Placing a "1"in the Parameter Edit box will cause the entered Parameter Value to be used by the COLA or to be active. Therefore a "0" would cause the value to be inactive. Remember that once a value is made active,it stays that way until the inactive option is chosen. l After all of the editing of the Parameter Edits screen is completed and the selections made active, hit the F10 key to close out the screen
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Number t TMt l NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Ttle Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 11 of 164 Sample Types (1D) SELECT TYPE OF SAMPLE DATA Minimum: 1 Default: None Maximum: 12 ACTIVE SAMPLE TYPE NO1 CONDENSER OFFGAS PATHWAY NO 2. REACTOR BUILDING DESIGN BASIS LEAKAGE PATHWAY NO 3. EMERGENCY FEEDPUMP EXHAUST PATHWAY NO4. ALPHA OTSG ADV PATHWAY NO 5. BRAVO OTSG ADV PATHWAY NO6. ALPHA OTSG MSR PATHWAY ! ( NO7. BRAVO OTSG MSR PATHWAY NO 8. STATION VENTILATION PATHWAY NO 9. REACTOR BUILDING PURGE PATHWAY l l NO 10. ESF VENTILATION PATHWAY l NO 11. REACTOR COOLANT BASELINE ACTIVITY NO 12. REACTOR COOLANT SPIKING FACTORS ENTER CHOICE (1 THRU 12) F1: Next F2: Prev Alt 3:10: Abort F10: Done The first 10 selections are TMI's 10 release pathways. Isotopic results from in-plant samples such as CATPASS, or RMA-5, RMA-8, RMA-9 sample panels can also be entered into these screens. MAP-5 samples cannot be input into the COLA, as there are no noble gas results. MAP-5 sample calculations must be performed using the manual codes. These samples provide a more refined dose projection than ones calculated from RMS data since they provide a GELI analysis of each isotope. The concentration obtained from these samples is combined with the release rate to calculate the source term leaving the plant. It is recommended that in-plant samples be obtained as practicable and entered into the sample edit screen.
- 1. Condenser Offgas Pathway - accepts RMA-5 sample panel marinelli beaker / iodine pre-filter sample results.
- 2. Reactor Building Design Basis Leakage Pathway- Accepts CATPASS sample results.
l [p\ t Q# 3. Emergency Feed Pump Exhaust Pathway- Sample Edit Screen inactive at this time. I
- 4. Alpha OTSG ADV Pathway - Sample Edit Screen inactive at this time.
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)
GU NN Tui j Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 12 of 164
- 5. Bravo OTSG ADV Pathway - Sample Edit Screen inactive at this time.
- 6. Alpha OTSG MSR Pathway - Sample Edit Screen inactive at this time.
- 7. Bravo OTSG MSR Pathway - Sample Edit Screen inactive at this time. !
- 8. Station Ventilation Pathway - accepts RMA-8 sample panel marinelli beaker / iodine pre-filter sample results.
- 9. Reactor Building Purge Pathway accepts RMA-9 sample panel marinelli beaker / iodine pre-filter sample results.
- 10. ESF Ventilation Pathway - Accepts iodine and noble gas sample results.
The 10 pathway sample edit screens, once sample results are entered, must be made active in order to be used by the COLA.
- 11. Reactor Coolant Baseline Activity - This is where RCS sample results are entered for use by the COLA.
Rad Engineering updates the default RCS activities on this screen as RCS changes during the core life. tgh RCS sample results are only used in the COLA dose projection if the Fuel Damage Class is 1. If the core
\Q damage is Damage Class 2 through 10, the COLA will use the RCS activities listed in the EDCM. Since the RCS activities listed in the EDCM are conservative it would be justified to request a Post Accident RCS Sample and input the results in the Sample Edit screen. You would then have to edit the Damage Class to 1 in order for the results to be used by the COLA. One major difference between this screen and the 10 pathway screens are the results are always active.
- 12. Reactor Coolant Spiking Factors - Normally the spiking factors, which are determined by the Nuclear Engineering Group, are entered onto the LAN by Rad Engineering. If these values are considered inaccurate, then the spiking factors may be edited using this screen. Remember that spiking factors are only used when the Damage Class is 1 AND the plant is at 0% power. One major difference between this screen and the 10 pathway screens is the results are always active.
To use this screen, select the release pathway and hit the F10 key to reach the actual isotopic input screen. Also you can select either the RCS activity or spiking Factors screen and then hit the F10 key for these inputs. Remember that the RCS activity and spiking factors screens are always active and used by the COLA, therefore the results in these two screens must always be accurate and up to date, if Post Accident RCS results are being used, ensure the spiking factors on this screen are set to 1, or the code will erroneously spike the sample results.
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V 86
Number I I U TMI g NUCLCAN Radiological Controls Procedure 6610-PLN-4200.02
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Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 13 of 164 Effluent Sample Screen Entry MUST be Y or N RADIONUCLIDE CONCENTRATIONS IN UCl/CC NOBLE GAS ISOTOPES RADIOIODINE ISOTOPES KR_85M 0.00E+00 l_131 0.00E+00 KR_85 0.00E+00 l_132 0.00E+00 KR_87 0.00E+00 l_133 0.00E+00 KR_88 0.00E+00 l_134 0.00E+00 XE_131M 0.00E+00 l_135 0.00E+ 00 XE_133M 0.00E+00
<G XE_133 0.00E+00 XE_135M 0.00E+00 XE_135 0.00E+00 XE_138 0.00E+00 Make Active (Y/N) N Enter values for only those nuclides designed to be collected by the analysis.
If these values are <MDA then enter the MDA value, not a zero. Enter a zero value for nuclides not designed to be collected by the analysis F1: Next F2: Prev Alt F10: Abort F10: Done The bottom statement on this screen may need some clarification. i.e. MAP-5 sample results only include the 5 iodine isotopes but not the 10 noble gas isotopes. As a result. the COLA will not calculate a nobte oas source term from a MAP-5 sample. The COLA can only use pre-filter marinelli sample results or CATPASS results, wht.re both noble gases and iodines are analyzed, to calculate the TEDE dose properly. If the MAP-5 sample is input inte the COLA, the COLA will calculate the correct thyroid dose. To enter sample resuits first select the release pathway and hit the F10 key. Enter the isotopic sample results for each isotope. When all 15 isotopic results are entered hit the F10 key and select "Y" to make the screen active. Hit the F10 key again to close out the screen. 87 I
6610-PLN-4200.02 Revision 9 EXHIBIT 1 Page 14 of 164 MIDAS Plume Plots (2A)
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MIDAS Screen -The primary purpose of the MIDAS Screen is compare field team readings and to indicate the plume location in order to send out field monitoring teams. If the field monitoring team results match up with this screen, then the accuracy of the dose projection is confirmed. The Group Leader R&EC is in charge of the dose , assessment process and will determine which dose projection / field team ratio is acceptable to confirm the dose I projection and possible PAR. Both the dose projection model and the MIDAS Program use the same source temi and should be in reasonable ' agreement if the COLA calculations are correct. The COLA dose projection source term is automatically used by MIDAS, howeverif the Manual Code is used by RAC personnel, the source term results must be provided to the EACC in order to be added to MIDAS. The MIDAS Program provides a snapshot that takes into account residual airborne radiation for a 2 hour block of time. It uses 8,15 minute sets of Met Data and is updated with a new set of data every 15 minutes. A new MIDAS Plume Plot snapshot is also updated every 15 minutes with the oldest Met Data set being eliminated and the new set of data added. As you can see MIDAS uses a different dispersion model than the RAC dose projection program, which projects the highest dose for the current time. This is why the COLA or Manual dose projection and not MIDAS is used for comparison to the PAGs and possible PAR. A PAR only looks at what the predicted integrated dose is for the future time of the release duration, not what has been released in the past. MIDAS includes past radioactivity in the atmosphere in its plume plots. 88
6610-PLN-4200.02 Revision 9 EXHIBIT 1 Page 14 of 164 MIDAS Plume Plots (2A) iTE-
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14 7.eE es,2.4E e 80 ' 3_.._. g . - , J;; as a,eE-e,e.4E-e,s 3.4E-e n.4E-ee 4.eE-Es Pot pggy' . - ggtycy . Ce0 t0eSH 10 FILE INTEREST N TURES eREA - MIDAS Screen -The primary purpose of the MIDAS Screen is compare field team readings and to indicate the plume location in order to send out field monitoring teams. If the field monitoring team results match up with this screen, then the accuracy of the dose projection is confirmed. The Group Leader R&EC is in charge of the dose ; assessment process and will determine which dose projection / field team ratio is acceptable to confirm the dose ' projection and possible PAR. 1 Both the dose projection model and the MIDAS Program use the same source term and should be in reasonable agreement if the COLA calculations are correct. The COLA dose projection source term is automatically used by l MIDAS, however if the Manual Code is used by RAC personnel, the source term results must be provided to the , EACC in order to be added to MIDAS. ! The MIDAS Program provides a snapshot that takes into account residual airborne radiation for a 2 hour block of time. It uses 8,15 minute sets of Met Data and is updated with a new set of data evary 15 minutes. A new MIDAS Plume Plot snapshot is also updated every 15 minutes wsh the oldest Met Data set being eliminated and the new set of data added. As you can see MIDAS uses a different o. ,persion model than the RAC dose projection program, which projects the highest dose for the current time. This is why the COLA or Manual dose piojection and not MIDAS is used for comparison to the PAGs and possible PAR. A PAR only looks at what the predicted integrated dose is for the futue time of the release duration, not what has been released in the past. MIDAS includes past radioactivity in the atmosphere in its plume plots. 88
Number ( ) N TML NUCLEAN \j Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 i EXHIBIT 1 Page 15 of 164 ) Secause the dose projection model and MIDAS use the same source term but different dispersion models (15 minutes vs. 2 hours), their results will most likely not match up. Since both programs use the same source term, if the field monitoring team results match up with MIDAS the dose projection is confirmed. If the results are unacceptable an investigation among dose assessment personnel is prudent. Here are some potential problems:
- 1. Did the Field Team really find plume centerline
- 2. Is there an unusual release pathway that is not indicated on the COLA
- 3. Is the OOLA using inaccurate information such as Met Data, RMS, flow rates
- 4. Are RMG 26 or RMG 27 results being incorrectly used by the COLA.
- 5. Is the Manual Code in use with more than one release pathway occurring from TMI.
Another difference between the dose projection and MIDAS is that MIDAS uses precipitation (wet deposition)in its calculation unlike the RAC Code models, Midas normally uses the same source term and met data as the COLA dose projection program. MIDAS Is \ automatically retrieves these parameters when the COLA is valid. Hovaver if the Manual Dose Projection modelis k used the source term and met data are not automatically added to MIDAS ay the Host Computer. EACC personnel must then use the " Stand Alone MIDAS" program to add met data and source term data. RAC personnel must then provide the source term results from the manual dose projec* ion to the EACC. EACC personnel can get the met data inputs from the LAN that need to be added to MIDAS. If LAN is down, a Meteorologist or the National Weather Service can be contacted. Also if the manual dose projection program is being used and the release is a Buoyant Plume Rise the release height found on the downwind dose results must be faxed to the EACC. This is only for releases via a MSRV or ADV. MIDAS produces the following plots:
- 1. Gamma Plot - Readings = Gamma - mR/hr Open Window - mR/hr Frisker - cpm
- 2. Child Thyroid Plot - Readings = Thyroid Rate mR/hr Cartridge cpm Filter - cpm Note that the cartridge and filter count rates are based on a standard 300 liter air sample, f")
N v 69
l l Numbsr I h U TMI NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 l \, 1 Tme Revmon No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 16 of 164 Air Sample Calculation (2C) Offsite lodine Air Sample Analysis inputs Location
...-................) ENE11 Date ----..---------..---) 2/3/98 Time --------------------> 1100 inputs Background (cpm) 50
.-_-.--.--..-..)
Flow Rate (LPM) ~ ~ ~ ~ - - - - - - - - - - - >29 Sample Run (min) ----------------> 11 Gross Cartridge (cpm) ------------> 300 Gross Particulate (cpm)
..~..-....) 50 Outputs ENTER TO QUIT Concentration (uCi/cc)
Thyroid Dose Rate (mrem /hr) The Air Sample Calculation Screen is used to ccnvert iodine and particulate results in net cpm to thyroid dose rate. The results displayed in the fields in the above screen are transmitted from the onsite/offsite field monitoring teams. The thyroid dose rate is multiplied by the release duretion in hours to provide the integrated thyrcid dose. The thyroid dose is then compared to the MIDAS Plume Plots to confirm the dose projection. To use this screen: , es
- 1. Select the 2C option from the EIN
- 2. Enter the results in each field and hit the " Enter" key,
- 3. Once the screen inputs are complete hit the " enter" key again
- 4. This will put you back on the EIN, print a copy of the results and clear the Air Sample Calculation Screen for future use.
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Number U TMI IQ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 17 of 164 Offsite Field Monitoring Team Data (28) Date: Measurements Dose Distance Monitoring Whole Body Thyroid Projection Time Field Team Miles Location mR/hr mrem /hr Agreement? Offsite Field Monitoring Team Data - This screen is used to add the results from the Field monitoring Team to the Emergency Information Network. EACC personnel for the offsite team results usually add this data to this screen. RAC personnel can also use the screen when the RAC computer is available. The information on this screen must be manually added. It is not automatically retrieved. To add information from the Emergency Information Network Main Menu perform the following:
- 1. Exit to DOS
(,s) 2. At the Y: prompt type in fielddat. bat and press enter.
/ 3. After entering the DOS screen, use the arrow keys to maneuver around to add the information required for this screen: Time, Field Team Designation, Downwind I)istances in miles ,REMP Map Monitoring Location, Whole Body Dose mR/hr, Thyroid Dose mR/hr, and Dorie Projection Agreement After entering the appropriate information select File, select Saie, and select Exit to leave the screen. %.)\ !
1 91 1 i
7 ' l Number TMI
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revson No TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 18 of 164 -
Reuter Stokes Data and MIDAS Calculation Comparison (3) COMPARISON OF RS MONITOR ACTUAL VICE MIDAS CALCULATED VALUES ACTUAL. MIDAL MONITOR . 01/08 08:09 01/08 10:56 CONFIRMATION? MIDDLETOWN 0.0083 0.0000 NOT AFFECTED NORTH GATE TMI 0.0058 0.0000 NOT AFFECTED ; MIDDLETOWN SUB 0.0062 0.0000 NOT AFFECTED ALWINE FARM 0.0087 0.0000 NOT AFFECTED VISITORS CTR 0.0077 0.0000 NOT AFFECTED 500 KVA SUB 0.0077 0.0000 NOT AFFECTED O g/ BECKER FARM 0.0077 0.0000 NOT AFFECTED v FLAMOUTH SUD 0.0069 0.0000 NOT AFFECTED CLY SUBSTATION 0.0061 0.0000 NOT AFFECTED TMIWAREHOUSE 0.0068 0.0000 NOT AFFECTED TMIMDCT 0.0079 0.0000 NOT AFFECTED GOLDSBORO 0.0064 0.0000 NOT AFFECTED TMI INTAKES 0.0070 0.0000 NOT AFFECTED FAIRVIEW TWP 0.0071 0.0000 NOT AFFECTED HPG AIRPORT 0.0080 0.0000 NOT AFFECTED CRAWFORD STATION 0.0091 0.0000 NOT AFFECTED CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT) Reuter Stokes Field Data and MIDAS Calculation Comparison - The Reuter Stokes screen provides actual readings from 16 lon chamber monitors located in each of the 16 sectors surrounding TMI. These readings are updated every 4 hours until any monitor exceeds 0.08 mR/hr (80 R/hr) which increases the updates to every 15 minutes. The primary purpose of this screen is to quickly obtain onsite/offsite doses in order to verify the dose projection.
; The system is set up to compare the actual monitor readings to the MIDAS projected monitor readings when any
( monitor exceeds 0.020 mR/hr (20 R/hr). The third column on this screen, "CONFlRMATION"is not operational and should be ignored. 92
l Number l TMI
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 l Tele Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 ) l EXHIBIT 1 Page 19 of 164 Individual Pathway Data - Source Term / input Data / Dose Data (4) RELEASE PATHWAY INFORMATION CONDENSER OFFGAS D1 DOSE 11 INPUTS S1 SOURCE TERM l RX BUILDING DESIGN BASIS LEAKAGE D2 DOSE 122 INPUTS S2 SOURCE TERM 1 EMERGENCY FEED PUMP EXHAUST D3 DOSE 13 INPUTS S3 SOURCE TERM A OTSG ADV DIRECT D4 DOSE 14 INPUTS S4 SOURCE TERM A OTSG MSR DIRECT D5 DOSE 15 INPUTS SS SOURCE TERM B OTSG ADV DIRECT D6 DOSE 16 INPUTS S6 SOURCE TERM I B OTSG MSR DIRECT D7 DOSE 17 INPUTS S7 SOURCE TERM STATION VENTILATION D8 DOSE 18 INPUTS S8 SOURCE TERM l RX BUILDING PURGE EXHAUST D9 DOSE 19 INPUTS S9 SOURCE TERM [D ESF EXHAUST D10 DOSE 110 INPUTS S10 SOURCE TERM L) RETURN TO MAIN MENU P1 ENTER THE APPLICATION NUMBER: To retrieve individual pathway information, select the pathway, such as the Reactor Building Design Basis Leakage. Select D2 for dose,12 for inputs, and S2 for source term. P1 retums to the Emergency information Menu. Listed are the 10 release pathways from TMI-1. - i i o 93 i { l
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v/ Radiological Controls Procedure 6610-PLN-4200.02 l Tdle Revrsion No. TMI Emergency Dose Calculation Manual (EDCM) 9 i EXHIBIT 1 Page 20 of 164
- Individual Release Pathway Dose Screen (D{1-10})
l (Typical) CONDENSER OFFGAS DOSE IN MREM (PATH #1) !
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TIME: 05:33:46 DATE: 01-151998 CALCULATION #: 1 l DISTANCE COMMITTED DOSE EQ (MREM) TOTAL EFFECTIVE DOSE EQ (MREM) Site Boun. 6.2E-06 1.7E-06 ; Ex. Area 4.0E-06 1.1 E-06 1 MILE 1.3E-06 3.4E-07 2 MILE 5.0E-07 1.3E-07 5 MILE 1.2E-07 2.8E-08 Q 10 MILE 5.2E-08 1.1 E-08 HIGHLIGHTED DOSES ARE MAXIMUM VALUES CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXil) i Individual Release Pathway Dose Screen Lists the downwind doses from the Site Boundary out to 10 miles for the j individual pathway selected. Whole Body (TEDE) and Child Thyroid (CDE) integrated doses are listed with the highest downwind dose location highlighted. 1 1 l J 94
Number ( U TMI g NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revson No. _TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 21 of 164
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Pathway Specific PPM Data (l{1-10}) RX BLDG DESIGN BASIS LEAKAGE (PATH #2) INPUT DATA TIME: 05:33:46 DATE: 01-15-1998 CALCULATION #: 1 PATHWAY SPECIFIC PPM DATA RB SPRAY L2825 (0 ON,1 OFF) 1.0E+00 RB SPRAY L2827 (0 ON,1 OFF) 1.0E+00 RB PRESS (PSIG) 1.1 E-01 RMA2 GAS (CPM) 6.1 E+02 RMA2 IOD (CPM) 1.3E+02 i9 RMG22 (R/HR) 9.8E-01 RMG23 (R/HR) 9.7E-01 CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT) Pathway Sper.ific PPM Data - The screen displayed for the Designed Basis Leakage shows the inputs or PPM i Data that was used by the COLA to calculate this individual pathway source term. Remember the COLA calculates each of the 10 pathways individually and adds them up for the most limiting dose. In this example RMG 22/23, along with the RB Pressure, are used as PPM inputs to calculate the source term. It also displays the closed position of the RB Spray system that can be used by operations, when conditions warrant it, to reduce the airbome atmospheric concentration of iodines in the Reactor Building after a LOCA. I l l I I ( v/ 4 95 i i
1 1 Humber U lQ l TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Reviston No. TMl Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 22 of 164 l Isotopic Release Rates (S(1-10}) RELEASE PATH #2 RX BLDG DESIGN BASIS LEAKAGE TIME: 05:33:46DATE: 01-15-1998 CALCULATION #: 1 ISOTOPIC RELEASE CONCENTRATIONS IN UCl/SEC l NOBLE GAS NUCLIDES RADIOlODINE NUCLIDES KR85M 9.1E-02 1131 1.4 E-02 KR85 5.8 E-08 1132 1.2E-01 KR87 2.9E 01 1133 2.01 E-01 KR88 3.2E-01 1134 6.5E-01 K XE131M 5.8E-08 1135 3.8E-01 O XE133M 5.8E-08 XE133 1.2E-01 TOTAL RI = 1.4E=01 XE135M 2.1E-01 XE135 9.9E-01 XE138 5.8E-08 TOTAL NG = 2.0E+00 CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT) Isotopic Release Rates - This screen shows the individual source term for the Design Basis Leakage release pathway. The results are listed for the 10 noble gas and 5 lodine isotopes used in the dose projection process. Also listed are the totals for the noble gas and iodine isotopes. This infonnation is frequently requested by the State BRP so they can verify our dose projections using their own dose projection model. l l l m
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I i I Number N TMI g NUCLEAR Radiological Controls Procedure 6610 PLN 4200.02 j l Tme Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 23 of 164 l Effluent RMS (5) EFFLUENT RMS READINGS TIME: 07:55:39 DATE: 01-16-1998 CALCULATION #: 1 RADIATION MONITORING SYSTEM RELEASE PATHWAY MONITOR NAME VALUE STATUS UNITS CONDENSER OFFGAS RM-A5(G) LOW 3.4E+01 NORMAL CPM RM-A15(G) LOW 3.7E+01 NORMAL CPM RM-A5(G) HIGH 1.9E+01 NORMAL CPM RM-G25 .2 E-01 NORMAL MR/HR STA VENTILATION RM-A8(G) LOW 4.6E+01 NORMAL CPM l RM-A8(l) LOW 1.6E+01 NORMAL CPM \ MIN RM-AB(G) HIGH 1.8E+ 01 NORMAL CPM RB PURGE RM-A9(G) LOW 3.9E+01 NORMAL CPM RM-A9f t) LOW 4.6E+ 01 NORMAL CPM \ MIN RM-A9(G) HIGH 2.4E+01 NORMAL CPM RM-G24 .2E+00 NORMAL MR/HR DIRECT OTSG RM-G26 2.8E+01 NORMAL CPM RM-G27 3.6E+01 NORMAL CPM CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT) The Effluent RMS Screen is updated from the Plant Performance Monitor (PPM) every 4 - 6 minutes. The screen displays the associated RMS effluent monitors by release pathway. This screen can be very useful to help validate the release pathway durin0 evaluation of the COLA dose projection. It also can be helpful to trend RMS data since these screens are saved once any Radiological EAL is resched. This screen provides a
- quick look" to see if any other release pathway RMS are increasing which could possibly affect the dose projection.
l 9 Normal RMS reading = gray background 9 Alert RMS reading = green background e High alarm reading = red background 0 Off scale reading = b!ack background [
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\g NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 ' EXHIBIT 1 Page 24 of 164 Area Gamma and Liquid RMS Screen (6) AREA GAMMA AND LIQUID RMS TIME: 10:55:50 DATE: 01 16-1998 CALCULATION #: 1 AREA MONITORS =MR/HR : RMG22/23=R/HR : LIQUID RMS=(CPM) RX BUILDING RM-G22 9.8 E-01 NORMAL RMG-23 9.7E 01 NORMAL RM-G6 2.8E+02 NORMAL RM-G7 5.2E+01 NORMAL RM-G5 5.5E-01 NORMAL AUX BUILDING RMG 10 6.0E-02 NORMAL RMG 112.6E-01 NORMAL RMG-12 8.0E-02 NORMAL RMG-13 9.0E-02 NORMAL RMG 14 8.0E-02 NORMAL RMG-15 9.0E-02 NORMAL FH BUILDING RMG-9 1.3E+00 NORMAL CONTROL TOWER RMG-1 8.0E-02 NORMAL RMG-2 5.0E-02 NORMAL RMG-3 1.6E-01 NORMAL RMG-4 6.0E-02 NORMAL g i SELECT LIQUID RML1L 4.2E+03 NORMAL RML1H 3.6E+01 NORMAL CURSOR LEFT (BACKWARD) CURSOR RIGHT (FORWARD) ALT-H (HARDCOPY) HOME (EXIT) The Area Gamma and Liquid RMS Screen displays the associated area gamma monitors by building along with selected liquid monitors. This screen is also very useful for evaluating in plant dose rates prior to sending personnel reentry teams into the plant. Area gamma monitor readings are useful to verify improving or degrading plant conditions. l l l bT U l 98 l
I l Number NI NUCLEAR Radiological Controls Procedure l 6610 PLN-4200.02 ! Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) g EXHIBIT 1 Page 25 of 164 Screen (7) This screen is reserved for future use. l i i l 99 I i
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/ U TMl l Q MUCLEAR Radiological Controls Procedure 6610 PLN-4700.02 l Title Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 26 of 164 Quick Code (8) l Minimum: 1 Default: None Maximum: 9 Input Data Release via Condenser Offgas
- 1. RMAS LOW OR RMA15 2. RMAS HIGH 3. RMG25 Release via ADV or MSR
- 4. RMG26 OR RMG27 Release via Reactor Building Purge
- 5. RMA9 LOW 6. RMA9 HIGH 7. RMG24 Release via Station Ventilation l 8. RMAS LOW 9. RMA8 HIGH l
Dx t 1 Enter Radiation Monitor Selection (1 thru 9) V Enter the Monitor Reading (1 thru 10 million) Enter the Wind Speed in MPH (1 thru 100) l Enter the Wind Direction in Degrees From (1 thru 360) Enter the Delta Temperature in Degrees F ( 9 to 10) l F1: Next F2: Prev Alt F10: Abort F10: Done l The Quick Code was designed with the on shift GRCS in mind. In the first hour of an event if the COLA goes out of , service the GRCS has two options to produce a dose projection, the Manual RAC Code or the Quick Code. It takes time to gather all of the inputs for the Manual RAC Code and the luxury of time may not be available. As you can i see above, the Quick Code inputs are all on one screen and consist of the RMS monitor and reading associated with the release pathway and met data. Default flow rates for each pathway are used in the Quick Code .As with the Manual RAC Code the Quick Code can only handle one release pathway. Also this code produces very conservative results during core damags situations as it uses a 1 1 iodine to noble gas ratio. l l l G 100 I l
T I Number I . TMl NUCLEAR Radiological Controls Procedure 6610-PLN-4200,02 { l l Title Revrsion No. l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 27 of 164 Manual RAC Code (9) l The Manual RAC Code is an option on the Emergency Information Network that will produce a dose projection based on user input, rather than data from the plant computer. The Manual RAC Code can be used for some of the following reasons: O To produce a dose projection if data being retrieved by the COLA is considered inaccurate (i.e. a bad RMG-25 reading or met data). e To perform a "What If' dose projection e Having a release through the MSRV's with MSV 2A/2B in the closed position since the COLA automatically uses RMG-26/27. O Multiple MSRV's are stuck open at the same time. 8 RB Design Basis Leakage release with RMG 22/23 out of service. 9 To compare or validate the dose projection for any other reason.
\ 8 To estimate offsite doses beyond 10 miles. l l
Results from the Manual Code are not displayed on tha LAN therefore they must be faxed to the EACC. Also if the Manual RAC Code is used, the source term results must be faxed to the EACC so they can be entered into the MIDAS program. If the COLA is inaccurate then the MIDAS results pro <iuced from the COLA source term is inaccurate. The Manual Code source term results must be entered into MIDAS in this situation. In addition,if the l COLA is inaccurate then the Group Leader RE&C, the EAC, and the state BRP representative must be notified that ' the COLA displays are not accurate. One disadvantage of the Manual Code is that it can handle only one release pathway at a time. If two release pathways are significant, each one must be run separate and the results added together. l f 101
I Number i N TMI ( WUCLEAR Radiological Controls Procedure 6610-PLN 4200.02 Tale Revision No. TMI Eme.gency Dose Calculation Mai;ual (EDCM) 9
- EXHIBIT 1 Page 2G of 164 If the LAN GOES OUT OF SERVICE the RAC computer will automatically boot up to the following menu:
EMERGENCY INFORMATION NETWORK LAN BACKUP CALCULATIONS IF THIS MENU APPEARS THE LAN CONNECTION WAS NOT MADE CONTACT l.S. AT X8393 WHEN TIME PERMITS USE ONE OF THE FOLLOWING MANUAL CODES TO PERFORM DOSE PROJECTIONS UNTil THE LAN PROBLEM iS CORRECTED
- 1. QU!CK CALC CODE (RECOMMENDED FOR ON-SHIFT GRCS/RAC)
- 2. MANUAL RAC
- 3. AIR SAMPLE CALC
- 4. EXIT TO DOS ENTER THE APPLICATION NUMBER:
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l Number (mj %u muct m m Radiological Controls Procedure 6610 PLN-4200.02 Tdle Revision No TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 29 of 164 Instructions for Rebooting the COLA , There are times when transients in the LAN system occur that cause the host computer performing the dase projections to lock up. As a result, the host computer quits performing dose projections and the last dose projection on the EIN will be more than 10 - 15 minutes old. When the COLA is locked up, it needs to be rebooted. A remote COLA reboot device has been instatied to do this. The following steps can be used to reboot the COLA frcm any site phone. Reboot can also be accomplished from an off-site phone simply by dialing "948" before the extension. If the automated dose projection system host computer (COLA host) is " locked up", reboot it as follows: Dial the COLA computer's reboot device at extension 8297. The call will connect and an electronic voice will state "please enter your access code,". Wait until the message is complete then enter "1979". The reboot electronic voice will then state " power is on" or " power is off" (normally "poner is on")" To reboot the COLA: Press "2" (the electronic voice will state " power is off"). Press "1" (the electronic voice will state " power is on"). Hang up the phone. The host computer will begin rebooting immediately. 1 Go to ine Y:\EPRAC directory and type reset, then press enter. i The process may take up to 15 minutes before new dose projections appear. 1 1 ! k I P s i 1 103
l Number i [ ') U TMI i C/ Tale NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Revision No. 1 TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 30 of 164 Basic RAC Checkoff l 1. Determine Release Pathways with Emergency Director.
- 2. Print Decision Support Screen (1A), RAC Screen (18), and input Screen on Predominate Pathway. (4 then 11-1 0)
- 3. Print Source Term Screen of Predominate Pathway (4 then S1 10) and review noble gas to iodine ratio.
(Normally a SO:1 or greaterratio.) l 4.. Validate inputs into the RAC Code Check that RMS is functional and operational. S. Review Field Team Data. (Reuter Stoks sonitors are like field teams.) Remember, Reuter Stokes uses 15 minute averages and compares against MIDAS. Field team data should be checked against MIDAS. MIDAS does not take into consideration environmentaliodine plateout or precipitation. MIDAS can estimate precipitation depletion of the plume upon request. The RAC Code is considered validated if field team data is within a tenth but not greater than MID/.S. 0 6. Use the TSC to verify inputs. (i.e., condition of charcoal filtration, damage class determination, leak rates in plant systems, etc.)
- 7. If the COLA i;, not providing accurate dose projections, for whatever reason, ensure that the Group Leader RE&C, the EAC, and the state BRP representative understand that the COLA screens are not displaying accurate dose projections.
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TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 31 of 164 Indications of Mechanical Fuel Damage For Thermal Hydrautic damage use RAC Code. For Mechanical Damage and as a gross backup to the RAC code here are 2 tables for consideration: LOCA with Leakrate and isotopic Concentrations in Equilibrium Fuel Damage Class RCS Activity (uCi/ml) RMG 22/23 (R/hr) 2-4 4500- 45,000 6000-45,000 5-7 5.6ES - 5.2E6 >45,000 LOCA Has NOT Occurred Fuel Damage Class RCS Activity (uCi/ml) RMG 22/23 (R/hr) 2-4 4500- 45,000 1.5-20 , ) 5-7 5.6ES - 5.2E6 >20 N) Reference RAF 6612-89-002 RM-G-6/7 Indications of Mechanical Fuel Damage During Non-LOCA Situations If 5% of the core has experienced cladding damage and released its gap activity, RM-G-6 would be expected to read between 30,000 to 40,000 mR/hr. Due to its normallocation, RM-G-7 would be expected to read about 10% of what RM-G-6 reads. Reference RAF 3640-98-034, b i
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Numbet / TMI () NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 32 of 164 EDCM Calculation Guides Condenser Offgas Releases Calculation Type When to Use Page COLA Using RMS COG RMS and flowrate instrumentation available on PPM 110 COLA Using Samples Representative RM-A-5 pre-filter /marinelli results available with flow rate available on PPM 113 Manual Code Using RMS COLA RMS calculation not available due to PPM or LAN problems 115 Manual Code Using Samples Representative COG pre-filter /marinelli results available with flow rate unavailable on PPM. Use this calc for MAP 5 samples. 118 Manual Code Contingency Use when RMS is not available. S'iould be used for Calculation performing "What If" calculations 120 4 Quick Code Used by on-shift RAC in first hour if COLA is not avavable 123 %) Reactor Building Design Basis Leakage Calculation Type When to Use Page COLA Using RMS RMG-22/23 and RB pressure available on PPM 125 COLA Using Samples Representative CATPASS results available with RB pressure available on PPM 128 Manual Code Using RMS COLA RMS calculation not available due to PPM or LAN problems or if using RMA-2 monitor results 130 Manual Code Using Samples Representative CATPASS results available with RB pressure unavailable on PPM 133 Manual Code Contingency Use when RMS is not available. Should be used for ! Calculation performing "What If" calculations 135 Quick Code Not available for this pathway 138 I i i v i 106 l
Number U TMl ( ,/ NUCLEAA Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 l i* EXHIBIT 1 Page 33 of 164 Emergency Feed Pump Exhaust (Also used for Main Steam Line Breaks) I Calculation Type When to Use Page i COLA Using RMS RMG-26/27 and emergency feedwater flow rate available on PPM 139 COLA Using Samples Not available for this pathway 142 Manual Code Using RMS RMG-26/27 or emergency feedwater flow not available due to PPM or LAN prob l ems 143 Manual Code Using Samples Not available for this pathway 146 Manual Code Contingency Use when RMS or feedwater flow is not available. Use to Calculation perform "What If" calculations. This calculation should be used for a main steam line break. 147 Quick Code Not available for this pathway 150 Atmospheric Dump Valve Release (MS-V-4A or MS-V-48) Calculation Type When to Use Page
, ) COLA Using RMS RMG-26/27 and main steam pressure available on PPM 151 COLA Using Samples Not available for this pathway 154 Manual Code Using RMS RMG-26/27 or main steam pressure not available due to PPM or LAN problems 155 Manual Code Using Samples Not available for this pathway 158 Manual Code Contingency Use when RMS or main steam pressure is not available.
Calculation Use to perform "What If" cales 159 Quick Code Used by on-shift RAC in first hour if COLA is not available 162 Main St- m Relief Valves ('A' or'B' side) Calculation Type When to Use Page COLA Using RMS RMG-26/27 and main steam pressure available on PPM and MS-V-2 is open 164 COLA Using Samples Not available for this pathway 167 Manual Code Using RMS RMG-26/27 or main steam pressure not available due to PPM or LAN problems and MS-V-2 is open 168 Manual Code Using Samples Not available for this pathway 171 Manual Code Contingency Use when RMS or main steam pressure is not available, Calculation or if MS-V-2 is closed. Should be used for performing j "What If" calculations 172 Quick Code Used by on-shift RAC in finst hour if COLA is not available 175 (O) v 107
i Number U ( ,) TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revm;on No. l i TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 34 of 164 Station Ventilation Calculation Type When to Use Page COLA Using RMS Station Vent RMS and ventilation flow rate are available on PPM 177 COLA Using Samples Representative RM-A-8 pre-filter /marinelli results available with flow rate available on PPM 180 Manual Code Using RMS Station Vent RMS or ventilation flow rate not available due to PPM or LAN problems 182 Manual Code Using Samples Representative RM-A-8 pre-filtertnarinelli results available with flow rate not available on Pf'M. Use this for MAP-5 samples 185 Manual Code Contingency Use when RMS or Station Ven' flow rate is not available. Calculation Should be used for performing V/ hat If" calculations 187 Quick Code Used by on-shift RAC in first hourif COLA is not available 188 Reactor Building Purge Exhaust N
\ / Calculation Type When to Use Page COLA Using RMS RB Purge RMS and ventilation flow rate are available on PPM 190 COLA Using Samples Representative RM-A-9 pre-filter /marinelli results available with flow rate available on PPM 193 Manual Code Using RMS RB Purge RMS or ventilation flow rate not available due to PPM or LAN problems 195 Manual Code Using Samples Representative RM-A-9 pre-filter /marinelli results available with flow rate not available on PPM. Use this for MAP-5 samples 198 Manual Code Contingency Use when RMS or RB Purge flow rate is not available.
Calculation Should be used for performing *What If" calculations 201 Quick Code Used by on-shift RAC in first hour if COLA is not available 204 l J 108 l
r Numbu i bU"C" TMI (e Radiological Controls Procedure 6610 PLN-4200.02 Tme Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 35 of 164 ESF Fuel Handling Building Calculation Type When to Use Page COLA Using RMS RM-A-14 and FHB vent flow rate are available on PPM 206 COLA Using Samples Representative RM A-14 pre-filter /marinelli results available with FHB flow rate on PPM 208 Manual Code Using RMS RM-A-14 or FHB flow rate not available due to PPM or LAN j problems 210 { Manual Code Using Samples Representative RM-A 14 pre-filter /marinelli results available with FHB flow rate not available on PPM. Uss this for iodine j cartridge samples 212 ) Manual Code Contingency Use when RM-A-14 is not available. J Calculation 214 ) Quick Code Not available for this pathway 216 Other Manual Calculations Available lAl Calculation Type When to Use Page D Manual Code RB Fuel Handling Fuel handling accident in the Reactor Building with RMS Accident Using RMS available 217 Manual Code RB Fuel Handling Fuel handiing accident in the Reactor Building with sample Accident Using Samples results available 220 Manual Code RB Fuel Handling Fuel handling accident in the Reactor Building with RMS or Accident Using Contingency sample results not available Calculation 221 Manual Code Waste Gas Decay Waste Gas Decay Tank rupture with RMS available Tank Rupture Using RMS 223 Manual Code Waste Gas Decay Waste Gas Decay Tank rupture with sample results available Tank Rupture Using Samples 225 Manual Code Waste Gas Decay Waste Gas Decay Tank rupture with RMS or sample results Tank Rupture Using Contingency not available Calculation 226 l l 5 v 109
NumbIr U TMI ( ) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tiue Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 36 of 164 Condenser Offgas COLA Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is 0.0075 (1/133) based on partitioning of iodine in the condenser
. The RMS monitor (RM-A-5, RM-A-15, RM-A-5 Hi, or RM-G-25) determines the activity leaving via this pathway. These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate.
. The logic the COLA uses in picking which monitor to perform the calculation is as follows:
O O If RM-G-25 and RM-A-5 Hi both read $; 100, the COLA is using the highest reading on RM-A-5 or V) ( 9 RM-A-15. If RM-A-5 Hi reads > 100 cpm, but RM-G-25 is < 100 mR/hr, the COLA is using RM-A-5 Hi. O If RM-G-25 reads > 100 mR/hr, the COLA is using RM-G-25
. Once the COLA determines the isotopic concentrations leaving via COG, it uses the COG flow rate from the plant computer to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. THIS PATHWAY iS ALWAYS CONSIDERED A GR( 'D LEVEL RELEASE User inputs Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours.
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
. Fuel Damage Class using the Parameter Edit screen if the damage class is believed to be other than that indicated by the COLA using current RCS temperature and pressure.
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i I j Nu nber U NUCLEAN TMI i Radiological Controls Procedure ' 6610-PLN 4200.02 Title Revision No. ! TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 37 of 164 i l Condenser Offg3s COLA Calculation I Using RMS , i Problems te Watch For
. Condenser offgas monitor increases can result from increased RCS activity orincreased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs:
O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core, i 9 C%cck for signs of increased primary-to-secondary leakage Check with STA or TSC to determine if i loak rate changes have been observed. I
. Remember that if fuel damage has occurred, it car'not get better.
) e if damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel L./ Damage Class.
9 If this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class. I O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. if post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel damage class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default isotopic activities for that damage class instead of the post accident samples that were input. If the spiking factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results.
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E 520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field l teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
l . If C.OLA dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field I [mk reading should be performed prior to issuing a PAR beyond 10 miles. U e Periodically verify the release duration and pathway with the ED. 111
Numt,ar U TMI
)
NUCLEAR Radiological Controls Procedure G610-PLN-4200.02
- Title Revision No TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 38 of 164 e Consider perfomiing manual "What It* dose projections for losing condenser vacuum or increases in damage class.
. Consider getting an effluent sample from this pathway (MAP 5 or pre-filter marinelli) when conditions appear to have stabilized.
. Always communicate the uncertainty of the dose projection. is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
O l l 112
r Number i U TMI
/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No.
I TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 39 of 164 Condenser Offgas COLA Calculation Using Samples How it works:
. Effluents leaving via COG are sampled using a pre-filter marinelli sample
. The user enters the positive noble gas and iodine isotope concentrations from the analysis into the Condenser Offgas Sample Edit Screen.
. If a MAP-5 sample is used, the manual code must be used to perform the calculation, since the COLA will not produce a noble gas source term.
. Once the COLA determines the isotopic concentrations leaving via COG. it uses the COG flow rate from the plant t,omputer to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the l q concentrations and doses at distances from the plant.
]
. THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inputs Needed to Perform the Calculation ,
l
. Sample results need to be entered on the Condenser Offgas Sample Edit Screen j 1
. Release duration using the Parameter Edit screen if other than 8 hours I l
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad 1
i l (m
\
113
c Number l f\ U TMI NUCLEAR
/
Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 40 of 164 Condenser Offgas COLA Calculation Using Samples Problems to Watch For
. The user must ensure that the sample results are represaative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should make the Condenser Offgas Sample Edit Screen inactive and resume COLA RMS calculations. ;
i
. MAP 5 samples cannot be used solely with the COLA code since noble gases are not measured. The COLA code will not produce a noble gas source term from a MAP-5 sample. MAP-5 samples must be calculated using the manual codes to get the correct TEDE dose. The MAP-5 sample should still be input into the COLA code since the thyroid doses produced by the COLA will still be accurate.
. Monitor met data, if met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
p) ( V
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If COLA dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for losing condenser vacuum or increase in damage !
class using a manual code contingency calculation.
. Always communicate the uncertainty of the dose projection. is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses
- actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
v) 114 L
l Number TMl y,/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revivon No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 41 of 164 Condenser Offgas Manual Code Calculation Using KMS How it works-1
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of ;
activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution. i
. The iodine reduction factor for this pathway is 0.0075 (1/133) based on partitioning of iodine in the ;
condenser. It is accounted for in the manual code. j
. The RMS monitor (RM-A-5, RM-A-15, RM-A-5 Hi, or RM-G-25) determines the activity leaving via this l pathway. These monitors will respond to both increases in RCS activity and primary-to secondary leakrate. I The user selects which monitor will be used for the calculation. Generally, the highest range monitor reading
> 100 stiould be used. l Once the code determines the isotopic concentrations leaving via COG, it uses the COG flow rate input by IA) e the user to determine the uCi/sec leaving the plant (source temi).
(./
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always considered a ground level release User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. The monitor to be used and the monitor reading
. The time in minutes since reactor shutdown
. Condenser offgas flowrate
. Met Data (wind speed, wind direction, and delta t)
. Release duration O
b 115 i.
l l Number TMI
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revmon No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 42 of 164 Condense' r Offgas Manual Code Calculation Using RMS Problems to Watch For
. Condenser offgas monitorincreases can result from increased RCS activity orincreased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs:
O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G 7 may indicate increased RCS activity from fuel damage. Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. O Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if leak rate changes have been observed.
. Remember that if fuel damago has occurred, it cannot get better.
( O If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class. 8 If this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there ;
has been no power transient when the code asks. If the code is using a damage class greater than 1, the ' code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up f with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get '
lodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /' r based on a 10 minute air sample. Consider longer sampling times if ,
additiorial sensitivity is desired. !
= If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading j should be performed priorto issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
l p). (
%/
Consider performing manual "Whr if" dose projections for losing condenser vacuum or increase in damage class using a manual code contingF.cy cc culation, l j l . Consider getting an effluent sample from this pathway (MAP-5 or pre-filter marinelli) when conditions appear to have stabilized. j 116
Number e N "CEN TMI 's Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 43 of 164
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be. \
l a I ( 117
r I 1 ( I f4urnber I i U TMI ( ,) t NuctFAR Radiological Controls Procedure 6610-PLN-4200.02 T0e Revision No.
]
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 44 of 164 Condenser Offgas Manual Code Calculation l Using Samples ' l How it works: l
. Effluents leaving via COG are sampled using a MAP-5 sample or a pre-filter marinelli sample l
. If a pre-filter marinelli sample is used, the user enters the positive noble gas and iodine isotope i concentrations from the analysis when prompted by the code. '
. If a MAP-5 sample is used, the user enters the positive iodine isotope concentrations from the analysis when prompted by the code. Since MAP-5 samples do not provide noble gas results, manual code will scale in the noble gases. It assumes that the noble gas to iodine ratio leaving COG is the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway.
. The iodine reduction factor for this pathway used to scale in noble gases from MAP-5 samples is 0.0075 (1/133) based on partitioning of iodine in the condenser
- d. Once the code determines the isotopic concentrations leaving via COG, it uses the COG flow rate input by the user to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The code will prompt the user to input values for RCS temperature and pressure or Fuel Damage Class values. These must be input but will have no bearing on the calculation since effluent concentrations leaving the plant are being determined by isotopic sample analysis.
. Sample results and sample time from condenser offgas
. Time since reactor shutdown
. Condenser offgas flowrate
. Met Data (wind speed, wind direction, and delta t)
A l ). Release duration LJ 118
l Number i U TMI NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Tale Revision No. l TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 45 of 164 Condenser Offgas Manual Code Calculation Using Samples Problems to Watch For
. The MAP-5 sample panel may present a dose rate hazard to personnelif the release rate is high or if it has j been running for a long time with the same cartridges, i
. The user must ensure that the sample results are representative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should begin using COLA or manual code RMS calculations.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
(b;
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual What If" dose projection for losing condenser vacuum orincreases in damage class.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what i believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
i r b V
- 119 L
I i Numbir l OT \ U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 l Tdle Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 46 of 164 Condenser Offgas Manual Code Calculation Contingency Calculation How it works:
. This calculation is very useful for performing "What If" calculations, where the precise response of an RMS j monitor to a change in plant conditions cannot be predicted. j
. This calcuktion uses no RMS data. The isotopic distribution and total activity being released via this i pathway is equal to the product of the primary-to-secondary leak rate and the activity concentration in the l RCS, after adjusting for iodine losses (reduction factors) applicable to this pathway. It is independent of condenser offgas flow rate.
. The iodine reduction factor for this pathway is 0.0075 (1/133) based on partitioning cf iodine in the condenser. It is accounted for in the manual code.
. Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has f- m the ability to use RM-L-1 high or low readings to estimate total RCS activity. The user may also specify a total RCS activity.
(v) . Havirg developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant. THIS PATHWAY iS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. RM-L-1 high or low readings if letdown is not isolated.
. An estimate of total RCS activity,if known.
. The primary-to-secondary leak rate, in gpm, provided by the STA or TSC.
. The time in minutes since reactor shutdown
. Met Data (wind speed, wind direction, and delta t)
. Release duration n
i \ G 120
g l l y WUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Remion No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 47 of 164 Condenser Offgas Manual Code Calculation Contingency Calculation Problems to Watch For
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are ideal for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an l RM-L-1 reading is input. Prior to inputting an R'LL-1 reading, verify that RCS letdown is not secured. If it has been secured (and it typically is during this type of accident), RM-L-1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS activity.
If RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc)if this [m (j . T is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed. If the dose rate on the sample is a factor of 10 higher than a normal RCS sample, it can be inferred that the RCS activity is 10 times higher than it was p.-ior to the incident. Be aware, entering a total RCS activity will replace damage class default RCS activities and spiking factors.
. Remember that if fuel damage has occurred, it cannot get better.
8 If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the co<1e will reduce the Fuel Damage Class. 8 If this occurs, the user must specify the damage classinstead of entering pressure and temperature. 8 After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage cbs assumed.
. Consider getting an effluent sample from this pathway (MAP 5 or pre-filter /marinelli) to increase the j accuracy of the dose projection. ;
I
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class grecter than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input. l
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
;. Depending on the physical form of iodine in the field, the *LLD" for thyroid dose rates measured by field
/ teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
121 I f I L !
Number i N TMI (g NUCLEAA Radiological Controls Procedure 6610-PLN-4200.02 Title Revtsion No. TMI Emergency Dose Calculation Manual (EDCM) 9 CXHIBIT 1 Page 48 of 164
.e if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
- Periodically verify the release duration and pathway with the ED.
. When performing "What if' dose projections, clearly labelinem as "What 11" calculations. . Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know ,
doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses < actually could be? Depending on the inputs, this type of calculation could go either way. The key will be the l accuracy of the RCS total activity and primary-to-secondary leak rate estimates. ! O ( 122
Number [ U TMI Q WCN Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 49 of 164 Condenser Offgas Quick Code Calculation How it works:
. This code provides a quick, but very conservative, calculation. As such it should only be used by the on-shift RAC, during the first hour of an emergency,if the COLA Code is not available.
. This code requires an RMS reading for its calculation. If an RMS reading is not available, a manual code contingency calculation must be performed.
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The RCS mixture is assurr a to be a 50 to 1 noble gas to iodine ratio and a radioiodine spike of 50. The resulting 1 to 1 noble gas to iodine ratio is very conservative for situations involving core damage.
. The iodine reduction factor for this pathway is 0.0075 (1/133) based on partitioning ofiodine in the cordenser. It is accounted for in the code.
/
( . The RMS monitor (RM-A-5, RM-A-15, RM-A-5 Hi, or RM-G-25) determines the activity leaving via this ( pathway. These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate. The user selects which monitor will be used for the calculation. Generally, the highest range monitor reading
> 100 should be used.
. The code assumes a COG flow of 40 cfm. Dose results may be ratioed by the actual cfm (i e. if the actual flow is only 20 cfm, the dose projection is reduced by %).
. The code assumes an 8 hour release duration.
. Once the code determines the isotopic concentrations leaving via COG, it uses the flow rate to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. THIS PATHWAY lS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
. Met Data (wind speed, wind direction, and delta t)
V.\ Since the COLA is not available, all readings must come from control room instrumentation. 123
E NumbIr i TMl V NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02
)
Tale Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 _ EXHIBIT 1 Page 50 of 164 Condenser Offgas Quick Code Calculation Problems to Watch For
. The IREO RAC should not use this code. The full manual code should be used when the IREO organization is activated.
. This calculation produces a bounding estimate of offsite dose, particularly for core damage situations due to the 1 to 1 noble gas to iodine ratio assumed in the RCS. When communicating the results of this code to the ED, the conservative nature of the results should be emphasized.
. As soon as time is available, a manual code RMS calculation should be performed so a dose estimate using more representative plant parameters can be obtained.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
(O i (I. . Depending on the physical form of iodine in the field, the "LLD" for thy oid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Label all Quickcode calculations as "Quickcode" I
t s i v' 124 l l l
Number > U NUCLEAR-TMl Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 51 of 164 Reactor Building Design Basis Leakage COLA Calculation Using RMS tigw it works:
. The isotopic distribution of activity leaving the plant via reactor building leakage is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is 0.4 with RB spray off, or 0.03 with RB spray or', based on minimal hold-up time in the building.
The RMS monitor (RM-G-22 or RM-G-23) determines the airborne activity in the reactor bui, ding. The COLA uses the higher of the readings on RM-G-22 or RM-G-23. If RM-G-22 or 23 read less than 1.5 R/hr, the COLA assumes they are seeing only background and assumes they read 0 for dose calculations. It does not use RMA-2 in the calculation. A l . Once the COLA determines the isotopic concentrations in the reactor building, it uses the leakage flow rate
\v from the building, based on building pressure, to determine the uCi/sec leaving the plant (source term).
Reduction of RB pressure to atmospheric pressure stops releases from this pathway.
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inouts Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours !
1
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
. Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that )
indicated by current RCS temperature and pressure. l i I i k i n V 125 l l
~
\
Number U TMI I
) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 1 Title Revision No 1
TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 52 of 164 i Reactor Building design Basis Leakage COLA Calculation Using RMS Problems to Watch For
. This calculation should only be used when the purge valves are closed. If they are open, some fomi of reactor building purge calculation should be performed.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel Damage Class. O If this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class A does not increase beyond the damage class assumed.
! \ l
. If post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel Damage Class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default isotopic activities for that damage class instead of the post accident samples that were input. If the spiking factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results.
. Monitor met data. If met data shows absolutely no changes for a 30 -45 minuto period,inere may be a problem with the data. Verify met data with EACC and/or Control Room Indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for losing containment integrity or increase in damage class.
t
- p. Consider getting an effluant sample from this pathway (pre-filter marinelli or CATPASS as applicable) when conditions appear to have stabilized.
\-s 126
Number J O Dau "C#" Radiological Co trols Procedure 6610 PLN-4200.02
- Revision No.
TMI Emergency Dose Calculation Manual (EDCM) g EXHIBIT 1 Page 53 of 164 e Always communicate the uncertainty of the dose projection, is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what i believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be. f 4 l l 1 s I l l 0 ( 1 m 127
Number [ j U TMI ( ,/ NN Radiological Controls Procedure 6610-PLN-4200.02 i Tdtle Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIB!T 1 Page 54 of 164 Reactor Building Design Basis Leakage , COLA Calculation ! Using Samples l How it works:
)
l
= A sample of the reactor building atmosphere can be obtained using the CATPASS system. j
. The user enters the positive noble gas and iodine isotope concentrations from the analysis into the Reactor !
Guilding Design Basis Leakage Sample Edit Screen. l l
. Once the isotopic concentrations in the reactor building are input, the COLA uses the leakage flo'n rate from j the building, based on building pressure, to determine the uCi/sec leaving the plant (source term). !
Reduction of RB pressure to atmospheric pressure stops releases from this pathway, ; i
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at dir snces from the plant. l
- p. THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE l l I !
LJ ' User inputs Needed to Perform the Calculation
. CATPASS sample results reed to be entered on the Reactor Building Design Basis Leakage Sample Edit Screen
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad m
128
Number U TMI J NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revmon No. l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 55 of 164 Reactor Building Design Basis Leakage COLA Calculation Using Samples Problems to Watch For
. At very high levels of airbome activity in the reactor building, a CATPASS sample can involve personnel exposure to high dose rates when obtaining the sample.
. . The CATPASS system takes 45 minutes to warm up so direct the OSC to start the warm up as soon as you think you may want this type of sample.
. The sample will typically be collected from the cooler discharge rather than the isokinetic nozzle to reduce moisture in the sample.
. The user must ensure that the sample results are representative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should make the Reactor Building Design Basis Leakage Sample Edit Screen inactive and resume COLA RMS calculations.
(h Je
; Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a l \. ) problem with the data. Verify met data with EACC and/or Control Room indications.
e Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired, e if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
1
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual "What If" dose projections for losing containment integnty using a manual code sample calculation, or increases in damage class using a manual code contingency calculation, l
e Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that i rcow ! doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be. l T
'V
( 129 I i
F Number I U TMl g NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title . Revtston No
- TMI Emergency Dose Calculation Manual (EDCM) 9 l
EXHIBIT 1 Page 56 of 164 Reactor Building Design Basis Leakage Manual Code Calculation l Using RMS l How it works: l
. The isoto)ic distribution of activity leaving the plant via reactor building leakage is a.ssumed to be tne same as the iso'opic distribution of artivity in the RCS after adjusting for iodi'le losses (reduction factors) ]
i applicable to this pathway. The total activity in the RCS is not important to this calculation,just the RCS isotopic distrioution, i
. The iodine roduction factor ior this pathway is 0.4 with RB spray off, or 0.03 with RB spray on, based on minimal holeup time in tN building.
. The RMS monitor (RM-A-2) determines the airborne activity in the reactor building. The manual code does not use RM-G-22 or RM-G-23. If RM-A-2 is isolated &4 osiin RB). this calculation cannot be used. Use COLA RMS calculation or manual contingency calculation, Once the code determines the isotopic concentrations in the reactor building, it uses the leakage flow rate i g g) . from the building, based on a user input building pressure, to determine the uClisec leaving the plant
(./ (source term). Reduction of RB pressure to atmospheric pressure stops releases from this pathway.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the i concentrations and doses at distances from the plant,
. THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
- RCS temperature and pressure or the user specified Fuel Damage Class
. Status : ' .:.e purge valves (closed if this calculation is being used)
. The RM-A-2 readings:
O Currea' RM-A-2 noble gas reading 9 Cunen. RM-A-2 lodin9 reading 9 Previaus RM-A-2 iodine reading e Time since previous RM-A-2 iodine reading
. Ths dime in min *9s since rescior shutdown
- p. Reactor building pressure
! I V
130
r Number t N TMI
/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9
~
EXHlBIT 1 Page 57 of 164 l Reactor Building Design Basis Leakage Manual Code Calculation Using RMS
. Met Data (wind speed, wind direction, and delta t)
. Release durat;on Problems to Watch For
. This calculation should only be used when the purge valves are closed. If they are open, some form of reactor building purge exhaust calculation should be performed.
. This calculation option is only good when reactor building pressure is below 4 psi. If RM-A-2 is isolated &4 osi in RBL this calculation cannot be used. Use COLA RMS calculation or manual contingency calculation if building pressure is above 4 psi.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class. 9 ff this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class l does not increase beyond the damage class assumed. j
. If post accident RCS results are to be input mto code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose iate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure fic .eams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of lodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections s Sw PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
n . Consider performing manual "What If" dose projections for losing containment integrity or increases in damage class using a manual code contingency calculation. (v) ) 131
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I l Number OUNUCLEAR TMI V Radiological Controls Procedure 6610-PLN-4200.02 TWJe Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHlBIT 1 Page 58 of 164 l l Reactor Building Design Basis Leakage l Manual Code Calculation j Using RMS
. Consider getting an effluent sample from this pathway ( pre-filter marinelli or CATPASS as applicable) when conditions appear to have stabilized.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
f ( I l i 132 L
Number U TMl s NUCLEAR Radiological Controls Procedure 6610-PLN4200.02 TCe Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 59 of 164 Reactor Building Design Basis Leakage Manual Code Calculation Uslog Samples How it works:
. A sample of the reactor building atmosphere can he obtained using the CATPASS system. ;
. The user eriters the positive noble gas and iodine isotope concentretions from the analysis when prompted .
by the code.
. Once the isotopic concentrations in the reactor building are input, the code uses the leakage flow rate from the building, based on a user input building pressure, to determine the uCi/sec leaking from the reactor building (source term). Reduction of RB pressure to atmospheric pressure stops releases from this pathway.
. Flaving developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
l' j . THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE
%./
User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA) j Met Data is available on Area 19, Group 2 of the PPC (See STA) i I
. The code will prompt the user to input values for RCS temperature and pressure or Fuel Damage Class l values. These must be input but will have no bearing on the calculation since effluent concentrations l leaving the plant are being determined by isotopic sample analysis. l
. CATPASS sample results and time of sample results, i
I
. Purge valve position (normally closed but open if doing a "what if containment is lost" calculation) '
. Reactor building pressure
. Met Data (wind speed, wind direction, and delta t)
]
. Release duration i
V ; I 133
Number i U TMI Q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tr' i Revision No. . TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 60 of 164 Reactor Building Design Basis Leakage Manual Code Calculation Using Samples Problems to Watch For
. This is a good calculation to perform a "What If" calculations for loss of containment integrity when a CATPASS sample has been obtained.
. The CATPASS system takes 45 minutes to warm up so direct the OSC to start the warm up as soon as you I think you may want this type of sample.
. The sample will typically be collected from the cooler discharge rather than the isokinetic nozzle to reduce moisture in the sample.
. The user must ensure that the sample results are representative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should begin using COLA or manual code RMS calculations.
b. g ( Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling timesif additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, vaSdition of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual "What If" dose projections for losing containment integrity or increase in damage class using a manual code contingency calculation.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
I 1 134 l L
r-Number
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U TMI NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 61 of 164 Re. actor Building Design Basis Leakage Manual Code Calculation Contingency Calculation How it works:
. Thi calculation is very useful for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathwai;s equal to the product of the total RCS leakage into the reactor building and the activity concentration in the RCS, after adjusting for iodine losses (reduction factors) applicable to this pathway.
.. The iodine reduction factor for this pathway is 0.4 with RB spray off, or 0.03 with RB spray on, based on j minimal hold-up time in the building..
. Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has the ability to use RM-L-1 high or low readings to estimate total RCS activity. The user may also specify a total RCS activity.
- d. Once the code determines the isotopic concentrations in the reactor building, it uses building leak rate, based on a user input building pressure, to determine the uCi/sec leaking from the building (source term).
Reduction of RB pressure to atmospheric pressure stops releases from this pathway.
. Having developed the source term, the code uses the meteorologkal dispersion model to calculate the concentrations and doses at distances from the plant.
. THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fue1 Damage Class l . RM-L-1 high or low readings if letdown is not isolated.
I . An estimate of total RCS activity,if known.
. The total RCS leakage into the RB, if known. The STA or TSC should provide RCS leak rates.
. The time in minutes since reactor shutdown
. The status of reactor building spray (on or off) 135
I l l Number U [Q NUCLEAR TMI Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. i TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 62 of 164 Reactor Building Design Basis Leakage Manual Code Calculation Contingency Calculation
. The status of the purge valves (closed if performing this calculation)
. Met Data (wind speed, wind direction, and delta t)
. Release duration Problems to Watch For l
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are idsal for perfomiing "What if" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an
' []) (V RM-L 1 reading is input. Prior to inputting an RM-L-1 reading, verify that RCS letdown is not secured. if it has been secured (and it typically le during this type of accident), RM-L 1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to i travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS activity.
. if RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc)if this is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed. If the dose rate on the sample is a factor of 10 higher than a normal RCS semple, it can be inferred that the RCS activity is 10 times higher than it was prior to the incident. Be aware.
enterino a total RCS activity will replace damaae class default RCS activities and spikino factors.
. The code will prompt the user for the total RCS leakage into the reactor building. If the leakage is less than 63,000 gallons (one RCS volume), the actualleakage should be input. If the leakage is more than 63,000 gallons, a maximum of 63,000 gallons should be used. Inputting more than 63,000 gallons will overestimate the source term because it will put more activity in the building than is available for release from the RCS.
. If the total RCS leakage is not known, the code will assume the entire RCS volume (63,000 gallons) has l been released to the building using a damage class 4. If total RCS leakage is not known and the damage class is other than 4, the user should respond that the leakage is known and specify the leakage as 63,000 gallons.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input t into the code, the code will reduce the Fuel Damage Class. 9 If this occurs, the user must specify the damage class instead of entering pressure and temperature.
?
p\ C/ O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed. 136
r l Number I8 l [j
\ NUCLE'AN TMI Radiological Controls Procedure 6610-PLN-4200.02 Title Reymion No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 63 of 164 Reactor Building Design Basis Leakage Manual Code Calculation Contingency Calculation
. Consider getting a CATPASS sample to increase the accuracy of the dose projection. The CATPASS system takes 45 minutes to warm up so direct the OSC to start the warm up as soon as you think you may want this type of sample.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is inc!uded in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. When performing "What If" dose projections, clearly label them as "What If" calculations.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? Depending on the inputs, this type of calculation could go either way.
l m
~J 137
Number U TMI q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 64 of 164 Reactor Building Design Basis Leakage Quick Code Calculation How it works: This calculation is not performed by the Quick Code. l 1 i l
)
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138
l Number ( N TMI NUCLEAR Radiological Controls Procedure f 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 65 of 164 Emergency Feed Pump Exhaust COLA Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is 0.5 based on partitioning ofiodine in the steam generators.
Additionaliodine reduction can occur if the OTSG level is raised above 600 inches.
. The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate.
. The COLA uses the higher reading of RM-G-26 or RM-G-27 p.. Once the COLA determines the isotopic concentrations in the main steam line, it uses the mass flow rate through the EFP turbine (based on emergency feedwater being supplied) to determine the uCl/sec leaving b) the plant (source term).
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User Inouts Needed to Perform the Calculatio_n
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
. Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that indicated by current RCS temperature and pressure.
I v 139 l
I I Number U I y; NUCLEAR TMI Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 66 of 164 Emergency Feed Pump Exhaust , COLA Calculation l Using RMS l l Problems to Watch For I i e Main steam line and condenser offgas monitor increases can result from increased RCS activity or j increased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs: i e Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the ! ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. 9 Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if leak rate chances have been observed. 1
. If the steam driven EFP (EFP-1) is on, request that the ED switch over to the electrically driven pumps to ]
terminate this release pathway, if this is not possible, request that the ED use the auxiliary boilers via l (p~J AS-V-4 to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used. v e Raising the OTSG level to 600 inches will provide additionaliodine reduction. Request that the ED consider increasing OTSG level for this purpose e if the steam driven EFP is on and being fed by steam from the OTSG's, the steam must travel past RM-G-26 or RM-G 27. There is no concern about the monitors being isolated by a closed MS-V-2.
- RM-G-26 and RM-G-27 are accident range monitors and are relatively insensitive. As a result, they will tend to overestimate offsite doses at low count rates. Normal background from the check sources on these monitors is typically 30 to 50 cpm.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel Damage Class. 9 )f this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class. O After doing this, the ser must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. Effluent samples are not available for this pathway. If fuel damage is suspected, consider asking for an RCS post accident sample to refine the dose projection.
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O l 140
F 1 1 l Number ' U Nuct m TMl Radiological Controls Procedure 6610 PLN-4200.02 Tme Remion No. I TMi Emergency Dose Calculation Manual (EDCM) 9
- EXHIBIT 1 Page 67 of 164 i Emergency Feed Pump Exhaust COLA Calculation Using RMS
. If post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel Damage Class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default isotopic activities for that damage class instead of the post accident samples that were input. If the spiking ,
factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results. '
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the 'LLD" for thyroid dose rates measured by field l teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if (Q additional sensitivity is desired. l
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
l l
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for an increase in damage class.
. Consider performing a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dose projection if RM-G-26 and RM-G-27 are reading close to background. l
. If a Site Protection Officeris stationed at the reactor building equipment hatch, consider moving the individual from that area.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that i know doses can1 be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? If RM-G-28 and RM-G-27 readings are less than 100 cpm, this will probably be a bounding type calculation. If they are above 100 cpm, this type of calculation should be a pretty good approximation of what the doses could be.
m) (O 141
s O y TMI N u ct t!A R Radiological Controls Procedure 6610-PLN-4200.02
~1 stie Revtsion No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 68 of 164 Emergency Feed Pump Exhaust COLA Calculation Using Samples How it works: This option cannot be used on the COLA at this time O I v 142
Number U TMl
\j NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 69 of 164 Emergency Feed Pump Exhaust Manual Code Calculation Using RMS How it works: l
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total ,
activity in the RCS is not important to this calculation, just the RCS isotopic distribution. l
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additional iodine reduction can occur if the OTSG level is raised above 600 inches
. The RMS monitor (RM-G-26 or RM-G27) determin s the activity in the main steam. These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate.
l
. The choice of which monitor should be based on which generator is supplying the pump.RM-G-26 for the 'A' side or RM-G-27 for the 'B' side.
fs
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Once the code determines the isotopic concentrations in the main steam line, it uses the mass flow rate (U) . through the EFP turbine. The user must specify this mass flow rate (obtain from TSC). Use default of 30,000 lb/hr until the TSC can provide a better value.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant. ,
THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. The monitor to be used and the monitor reading.
. The time in minutes since reactor shutdown
. The mass flow rate of steam exiting the EFP turbine must be specified by the user in the " Sum of other Direct to Atmosphere Releases' space on the screen used to specify which MSR's or ADVs are open. The mass flow rate can be obtained from the TSC. Use default of 30,000 lb/hr until the TSC can provide a better value.
, . Met Data (wind speed, wind direction, and delta t) be Release duration 143
1 l l Number U TMI k WCN , g Radiological Controls Procedure 6610-PLN-4200.02 l Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 70 of 164 l Emergency Feed Pump Exhaust l Manual Code calculation i Using RMS ' Problems to Watch For
. Main steam line and Condenser offgas monitor increases cari result from increased RCS activity or increased primary to-secondary leakage, or a combination of the two. If a sudden increase occurs:
9 Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G 7 may indicate increased RCS activity fcom fuel damage. Also check the ! ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. 9 Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if leak rate changes have been observed.
. If the steam driven EFP (EFP-1) is on, request that the ED switch over to the electrically driven pumps to
[T terminate this release pathway. If this is not possible, request that the ED use the auxiliary boilers via j j AS-V-4 to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used. V e Raising the OTSG level to 600 inches wi:1 provide additionaliodine reduction. Request that tne ED consider increasing OTSG level for this purpose.
. If the steam driven EFP is on and being fed by steam from the OTSG's, the steam must travel past RM-G-26 or RM-G-27. There is no concern abot.t the monitors being isolated by a closed MS-V-2.
. RM-G-26 and RM-G-27 are accident range monitors and are relatively insensitive As a result, they will tend to overestimate offsite doses at low count rates. Normal background from the check sources on these monitors is typically 30 to 50 cpm.
. If the mass flow rate through the pump canr'ot immediately be determined, use the manual code contingency calculation for this pethway.
. This pathway is always considered a ground level release. However,if this calculation is run with an ADV or MSR also open, tne mde may inappropriately classify it as an elevated release. If the EFP is running with an ADV or MF both cases must be run separately and the doses added together.
. Remembc mm if fuel damage has occurred,it cannot get better.
O if damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, tne code will reduce the Fuel Damage Class. 9 If this occurs, the user must specify the damage class instead of entering pressure and temperature, p)- 9 After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed. 144
I Number t
, TMl G WUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Tee Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 j
- EXHIBIT 1 Page 71 of 164 l
l Emergency Feed Pump Exhaust 1 Manual Code Calculation Using RMS
. Effluent samples are not available for this pathway. If fuel damage is suspected, consider asking for an RCS post accident sample to refine the dose projection.
l
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there I has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up j with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get I lodine samples even if E-520 dose rates are low.
I
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field I teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
5
% . If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What if' dose projections for losing condenser vacuum orincreases in damage class using a manual code contingency calculation,
. Consider perfonning a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dor,e projection if RM-G-26 and RM-G-27 are reading close to background.
. If a Site Protection Officer is stationed at the reactor building equipment hatch, consider moving the individual from that area.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what i believe offsite doses
! actually could be? If RM-G-26 and RM-G-27 ret. dings are less than 100 cpm, this will probably be a very bounding type calculation. If they are above 100 cpm, this type of calculation should be a pretty good l approximation of what the doses could be. O D 145
r -' Number U TMI j NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Tale . Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 72 of 164 Emergency Feed Pump Exhaust Manual Code Calculation Using Samples l l How it works- l This option cannot be used on the manual code at this time i i l i l l l [ t 146
Number l TMI
\ NUCLFAR Radiological Contr61s Procedure 6610-PLN-4200.02 Title Revison No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 73 of 164 Emergency Feed Pump Exhaust / Main Steam Line Break Manual Code Calculation Contingency Calculation How it works:
. This calculation is very useful for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted, it is also good to use if the mass flow throu0h the EFP cannot immediately be obtained and the 30,000 lb/hr default is not believed to be representative, or if RM-G-26 and RM-G-27 are at background levels.
. This is the calculation that should also be used in the case of a main steam line break.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is equal to the product of the primary-to-secondary leak rate and the activity concentration in the RCS. The source term rate is independent of mass flow through the steam driven EFP or the main steam line break.
fN . The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the stcam generators. ( Additionaliodine reduction can occur if the OTSG levelis raised above 600 inches.
. Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has the ability to use RM-L-1 high or low readings to estimate total RCS activity. The user may also specify a total RCS activity.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
THIS PATHWAY IS ALWAYS CONSIDERED A GROUND LEVEL RELEASE User inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. RM-L-1 high os low readings if letdown is not isolated.
. An estimate of total RCS activity,if known.
. The primary-to-secondary leak rate, in gpm, provided by the STA or TSC.
l
. The time in minutes since reactor shutdown ge Met Data (wind speed, wind direction, and delta t) \ ]
U= Release duration l l 147
p i t Numbef U TMI l NuCLFAA Radiological Controls Procedure 6610-PLN 4200.02 T ti. Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 i EXLIBIT 1 Page 74 of 164 Emergency Feed Pump Exhaust / Main Steam Line Break Manual Code Calculation Contingency Calculation Problems to Watch For
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are ideal for performing "What if* calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. If the steam driven EFP (EFP-1) is on, request that the ED switch over to the electrically driven pumps to terminate this release pathway. If this is not possible, request that the ED use the auxiliary boilers to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used.
l . Raising the OTSG level to 600 inches will provide additionaliodine reduction. Request that the ED consider increasing OTSG level for this purpose. ! . If the mass flow rate of steam through the pump turbine cannot immediately be determined, this is a good l p) ( v calculation to use. If all ADVs and MSR's are kept closed, all primary activity entering the secondary side will be assumed to be released via this pathway.
. If RM-G-26 and RM-G-27 are at background levels, this is a good calculation to perform to verify dose i projections made using RMS data.
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an RM-L-1 reading is input. Prior to inputting an RM-L 1 reading, verify that RCS letdown is not secured. If it has been secured (and it typically is during this type of accident), RM-L-1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to l
travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS activity.
. If RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc) if this is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed, if the dose rate on the sample is a factor of 10 higher than a normal RCS sample, it can be inferred that the RCS activity is 10 times higher than it was prior to the incident. Be aware, entering a total RCS activity will replace damage class default RCS activities and spiking factors.
l
. Remember that if fuel damage has occurred, it cannot get better. I e if damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class.
9 If this occurs, the user must specify the damage class instead of entering pressure and temperature. 8 After dcing this, the user must watch RCS temperature and pressure to ensure that damage class e does not increase beyond the damage class assumed. 148 ! l i
t l l l Number
/ U TMI Q NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Reymon No.
TMI Emergency Dose Calculation Manual (EDCM) 9 1 EXHIBIT 1 Page 75 of 164 Emergency Feed Pump Exhaust / Main Steam Line Break i Manual Code Calculation l Contingency Calculation
. Effluent samples are not available for this pathway. Consider requesting an RCS sample (normal or post accident) to increase the accuracy of the dose piojection.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class in.-?ead of the post accident samples that were input.
. This pathway is always considered a ground level release. However,if this calculation is run with an ADV or ;
MSR also open, the code may inappropriately classify it as an elevated release. If the EFP is running with I an ADV or MSR open, both cases must be run separately and the doses added together.
. Remember that the CDE contnbutes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get e iodine samples even if E-520 dose rates are low.
U. Depending on the physical form of iodine in the field, the "LLD* for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading l should be performed prior to issuing a PAR beyond 10 miles. l
. Periodically verify the release duration and pathway with the ED. l
. When performing 'What If" dose projections, clearly label them as "W1at if' calculations.
. if a Site Protection Officeris stationed at the reactor building equipment hatch, consider moving the l individual from that area.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know l doses cant be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? Depending on the inputs, this type of calculation could go either way.
l G 149 l
l l Number N NUCLEAR TMI ' Radiological Controls Procedure 6610-PLN-4200.02 i l Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 76 of 164 Emergency Feed Pump Exhaust Quick Code Calculation < How it works: This calculation is not performed by the Quick Code. O i a 150
l I Number U TMl NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revnion No. l TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 77 of 164 Atmospheric Dump Valve Release (A or B Side) COLA Calculation Using RMS How it works: I
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution. . The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additionaliodine reduction can occur if the OTSG levelis raised above 600 inches.
. The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam. These monitors wih respond to both increases in RCS activity and primary-to-secondary leakrate. . The COLA uses the reading on RM-G-26 if MS-V-4A is open and the reading on RM-G-27 if MS-V-4B is open.
r\ ( (/
). Once the COLA determines the isotopic concentrations in the main steam line, it uses the mass flow rate out the ADV. The mass flow rate is a function of the main steam pressure and the percent demand on the MS-V-4 valve. Both of these parameters are picked up off the plant computer. . Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
AS A RESULT OF BUOYANT PLUME RISE. THIS PATHWAY IS ALMOST ALWAYS AN ELEVATED RELEASE. User inputs Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours ; . Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad . Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that indicated by current RCS temperature and pressure.
O O 151
r Number TMI (/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200,02 l Title Revision No. ] l \ l TMl Emergency Dose Calculation Manual (EDCM) 9 _ EXHIBIT 1 Page 78 of 164 l Atmospheric Dump Valve Release (A or B Side) COLA Calculation Using RMS Problems to Watch For )
. Main steam line and condenser offgas monitor increases can result from increased RCS activity or increased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs: )
l 0 Check for signs of increased RCS activity. Coincident increases on RM-L 1, RM-G-22, RM G-23, I RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. 8 Check for signs of increased primary-to secondary leakage. Check with STA or TSC to determine if leak rate changes have been observed. 1 . Be aware that the COLA measures the degree of the ADV opening by the percent demand (how much it ; j - should be open for plant conditions). If an ADV were to stick open, but did not need to be open, the percent ' l'm V) demand will be 0 and the COLA will not recognize the ADV is open. As a result, it will compute a source term of 0. If th;s occurs, a manual code RMS calculation should be performed specifying the valve as 100% l open. An altemative would be to edit the COLA Parameter Edit Screen to say an MSR is stuck open. The latter will provide a more conservative dose estimate (factor of 2), but will keep the EIN and MIDAS plume plots reasonably accurate. i e if the steam driven EFP (EFP-1)is on, request that the ED switch over to the electrically driven pumps to l terminate this release pathway, if this is not possible, request that the ED use the auxiliary boilers via AS-V-4 to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used. l
. Raising the OTSG level to 600 inches will provide additional iodine reduction. Request that the ED consider l increasing OTSG level for this purpose )
l
. If the release is through an ADV, the steam must travel past RM-G-26 or RM-G-27. There is no concern i about the monitors being isolated by a closed MS-V-2. l 1
. RM-G-20 and RM-G-27 are accident range monitors and 2,re relatively insensitive. As a result, they will tend
- to overestimate offsite doses at low count rates. Normal background from the check sources on these l monitors is typically 30 to 50 cpm.
. Remember that if fuel damage has occurred, it cannot get better.
1 0 If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel l Damage Class. c ; N l 152 i
Number U TMI ( ,/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Titie Remson No. __TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 79 of 164 Atmospheric Dump Valve Release (A or B Side) COLA Calculation Using RMS 8 If this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. Effluent samples are not available for this pathway. If fuel damage is suspected, consider asking for a RCS post accident sample to refine the dose projection.
. If post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel Damage Class is set to 1 on the Pammeter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default isotopic activities for that damage class instead of the post accident samples that were input. If the spiking factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample resulta A
Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a (v) . problem with the data. Verify met data with EACC and/or Control Room indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE close. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivityis desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
.- Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for an increase in damage class using the manual code contingency calculation
. Consider performing a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dose projection if RM-G-26 and RM-G-27 are reading close to background.
. Always communicate the uncer'ainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what i believe offsite doses actually could be? If RM-G-26 and RM-G-27 readings are less than 100 cpm, this will probably be a bounding type calculation. If they are above 100 cpm, this type of calculation should be a pretty good
- g. m approximation of what the doses could be.
t I v 153
l Number
\
U TMI NUCLEAR 1 Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 l l EXHIBIT 1 Page 80 of 164 ! I Atmospheric Dump Valve Release (A or B Side) ! COLA Calculation l Using Samples How it works: This option cannot be used on the COLA at this time n U I I if f 1 i 154
I Nmber f TMI
'd WCN Radiological Controls Procedure 6610-PLN-4200.02 Tdl* Revtsion No.
TlW'l Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 81 of 164 Atmospheric Dump Valve Release (A or B Side) Manual Code Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additionaliodine reduction can occur if the OTSG levelis raised above 600 inches. These reduction factors are accounted for in the manual code. I
+
The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam. These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate. { l
. The choice of which monitor should be based on which ADV is open. RM-G-26 for MS-V-4A or RM-G-27 for l N MS-V-4B. If both valves are open, use the higher of the two monitors.
. Once the code determines the isotopic concentrations in the main steam line, it uses the mass flow rate out the ADV. The mass flow rate is a function of the main steam pressure and the percent open of the MS-V 4 valve. The user specifies both of these parameters.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. As a result of buoyant plume rise, this pathway is almost always an elevated release.
User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. The monitor to be used and the monitor reading.
. The *.ime in minutes since reactor shutdown
. The main steam pressure
. The percent the ADV is open.
I
. Whether generator level is above 600 inches or not '
p j i e Met Octa (wind speed, wind direction, and delta t)
. Release duration 155 l i
( l Number t U TMI l
'\_) NUCLEAR Radiological Controls Procedure 6610-PLN-4200,02 Title Rewsion No. j TMl Emergency Dose Calculation Manual (EDCM) 9 l
EXHIBIT 1 Page 82 of 164 l Atmospheric Dump Valve Release (A or B Side) . Manual Code Calculation l Using RMS Problems to Watch For l . Main steam line and condenser offgas monitor increases can result from increased RCS activity or I increased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs: i l 0 Check for signs of increased RCS activity, Coincidem increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. l 8 Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if I leak rate changes have been observed.
. If the steam driven EFP (EFP-1) is on, request that the ED switch over to the electrically driven pumps to terminate this release pathway. If this is not post request that the ED use the auxiliary boilers to drive the pump. Ask to switch to the unaffected OTSt al the boilers can be used.
[]
. Raising the OTSG level to 600 inches will provide additior al iodine reduction. Request that the ED consider increasing OTSG level for this purpose e if the release is through an ADV, the steam must travel past RM-G-26 or RM-G-27. There is no concern about the monitors being isolated by a closed MS-V-2.
. RM-G 26 and RM-G-27 are accident range monitors and are reatively insensitive. As a result, they will tend to overestimate offsite doses at low count re'es. Normal backgro ind from the check sources on these monitors is typically 30 to 50 cpm.
. Remember that if fuel damage has occurred, it cannot get better.
O if damage has occurred but RCS temperature and pressure improvc. and the new values are input j into the code, the code will reduce the Fuel Damage Class. l l 9 ff this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. Effluent samples are not available for this pathway. If fuel damage is suspected, consider asking for an RCS post accident sample to refine th0 dose projection.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient wnen the code asks. If the code is using a damage class greater than 1, the (N code will use the default isotopic activities for that damage class instead of the post accident samples that i
; were input.
! \ 156 1 l l
1 Number U TMI () NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 I Title Reymon No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 83 of 164 Atmospheric Dump Valve Release (A or B Side) Manual Code Calculation Using RMS
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get l lodine samples even if E-520 dose rates are low.
. Depending on the physical forrn of iodine in the field, the 'LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manuil"What If" dose projections for an increase in damage class using the manual
{] code contingency calculation.
. Consider performing a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dose projection if RM-G-26 and RM-G-27 are reading close to background.
. This pathway is always considered an elevated release. However, if this calculation is run with the EFP running, the code will erroneously classify the EFP release as an elevated release. With the EFP running both cases must be run separately and the doses added together.
. Always communicate the uncertainty of the dose projection, is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? If RM-G-26 and RM-G-27 readings are Icss than 100 cpm, this will probably be a very bounding type calculation. If they are above 100 cpm, this type of calculation should be a pretty good approximation of what the doses could be.
O (s l- 157 l
F-J i-J l Number U TMt
~
NUCLEAR Radiological Controls Procec:.re 6610-PLN-4200.02 Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) l 9 EXHIBIT 1 Page 64 of 164 l Atmospheric Dump Valve Release (A or B Side) i Manual Code Calculation l l Using Samples I i l How it wort g l This option cannot be used on the manual code at this time l l !O 1 l I t O 156 f
r l Number l n TMI (w NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revasion No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 85 of 164 Atmospheric Dump Valve Release (A or B Side) Manual Code Calculation Contingency Calculation How it works:
. This celculation is very useful for performing What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted, or if RM-G-26 and RM-G-27 are at background levels.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is equal to the product of the primary-to-secondary leak rate and the activity concentration in the RCS. The source term rate is independent of mass flow through the ADV.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additional iodine reduction can occur if the OTSG level is raised above 600 inches.
. Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has O the ability to use RM-L-1 high or low readings to estimate total RCS activity. The user may also specify a f total RCS activity.
. Having developed the source tenn, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. As a result of buoyant plume rise, this pathway is almost always an elevated release.-
User inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. RM-L-1 high or low readings if letdown is not isolated.
. An estimate of total RCS activity,if known.
. The primary-to-secondary leak rate, in gpm, provided by the STA or TSC.
. The main steam pressure
. The percent the ADV is open
. Whether generator level is above 600 inches or not l C. N The time in minutes since reactor shutdown U. Met Data (wind speed, wind direction, and delta t) l
. Release duration 159
r- i l l Number i U TMI r (/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Remion No. j TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 86 of 164 Atmospheric Dump Vaive Release (A or B Side) Manual Code Calculation Contingency Calculation Problems to Watch For I l
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are ideal for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot I be predicted.
. If RM-G-26 and RM-G-27 are at background levels, this is a good calculation to perform to verify dcse projections made using RMS data.
. If the steam driven EFP (EFP-1)is on, request that the ED switch over to the electricayy driven pumps to terminate this release pathway. If this is not possible, request that the ED use the auxiliary boilers to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used.
. Raising the OTSG level to 600 inches will provide additional iodine reduction. Request that the ED consider N increasing OTSG level for this purpose. l
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an !
RM-L-1 reading is input. Prior to inputting an RM-L-1 reading, verify that RCS letdown is not secured. Ifit l has been secured (and it typically is during this type of accident), RM-L-1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS activity.
. If RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc)if this is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed. If the dose rate on the sample is a factor of 10 higher than a normal RCS sample, it can be inferred that the RCS activity is 10 times higher than it was prior to the incident. Be aware, entering a total RCS activity will replace damage dass default RCS activ" es and spiking factors.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input l into the code, the code will reduce the Fuel Damage Class. 8 If this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class l does not increase beyond the damage class assumed.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power traqsient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that
[n) NJ were input.
. Effluent samples are not available for this pathway. Consider requesting an RCS Sample (normal or post accident)to increase the accuracy of the dose projection.
160 L
Number
/ U TMI
() NUCLEAR Radiological Controls Procedure 6610-PLN-4200 32
~
Title Revmon No. l TMI Emergency Dose Calculation Manual (EDCM) 9 ) EXHIBIT 1 Page 87 of 164 Atmospheric Dump Valve Release (A or B Side) Manual Code Calculation Contingency Calculation
. This pathway is etways considered an elevated release. However,if this calculation is run with the EFF running, the code will erroneously classify the EFP release as an elevated release. With the EFP running both cases must be run separately and the doses added together.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose sate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading thould be performed prior to issuing a PAR beyond 10 miles.
/
- j. Periodically verify the release duration and pathway with the ED.
(mJ
. When perfomiing "What If" dose projections, clearly label them as "What If" calculations.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what i believe offsite doses actually could be? Depending on the inputs, this type of calculation could go either way.
l 1 l t
- 0. i b/
161
Number N (jb NUCLEAR TMI Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. l TMl Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 88 of 164 Atmospheric Dump Valve Release (A or B Side) Quick Code Calculation liow it works:
. This code provides a quick, but very conservative, calculation. As such it should only be used by the on-shift RAC, during the first hour of an emergency, if the COLA Code is not available.
. This code requires an RMS reading forits calculation. If an RMS reading is not available, a manual code contingency calculation must be performed.
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of i
activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The RCS mixture is assumed to be a 50 to 1 noble gas to iodine ratio and a radioiodine spike of 50. The resulting 1 to 1 noble gas to iodine ratio is very conservative for situations involving core damage.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators. It is accounted vrin the code.
(h V) . ( The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam. These monitors will respond to both increases in RCS activity and primary-to secondary leakrate. The user selects which monitor will be used for the calculation.
. The code assumes the mass flow rate to be equivalent to an ADV 100% open at a main steam pressure of 1040 psi.
l . The code assumes an 8 hour release duration.
. Once the code determines the isotopic comentrations in tb9 main steam line, it uses the flow rate out of the ADV to determine the uCi/sec leaving the plant (source te: ..).
l 1
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always considered a ground level release l User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
i
. Met Data (wind speed, wind direction, and delta t)
Since the COLA is not available, all readings must come from control room instrumentation. Q. 162
Number i f\ e v ER
- TMi
\j NW Radiological Controls Procedure 6610-PLN42r .02 Tme Revision No TMl Emergency Dose Calculation Manual (EDCM- 9 EXHIBI' 1 Page 89 of 164 Atmospheric Dump Valvt I.alease (A or D Side)
Quick Code r e.lculation Problems to Watch For j
. The IREO RAC should not use this code. The full manual code should be used when the IREO organization is activated.
1
. This calculation produces a bounding estimate of offsite dose, particularly for core damage situaticas due to the 1 to 1 noble gas to iodine ratio assumed in the RCE. When communicating the results of this code to the ED, the conservative nature of the results should be emphasized.
. As soon as time is available, a manual code RMS calculation should be performed so a dose estimate using more representative plant parameters can be obtained. )
1
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up I with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-620 dose rates are low. !
O). ( V Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if I l additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles,
. Periodically verify the release duration and pathway with the ED.
. Label all Quickcode calculations as "Quickcode" IO o ;
163
l Number - U TMI ( NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 I EXHIBIT 1 Page 90 of 164 Main Steam Safety Release (A or B Side) COLA Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total .
activity in the RCS is not important to this calculation, just the RCS isotopic distribution. ) l
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additional iodine reduction can occur if the OTSG levelis raised above 600 inches.
. The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam. These (nonitors will i respond to both increases in RCS activity and primary-to-secondary leakrate.
. The COLA uses the highest reading on RM-G-26 or RM-G-27. 4 l
pg . , I There is no plant computer point that provides indication that an MSR is open. The user must tell the COLA that an MSR is open by editing the Parameter Edit Screen. Opening an MSR on the COLA does not specify l V whether it is on the A or B side. l l
. Once the COLA determines the isotopic concentrations in the main steam line, it uses the mass flow rate out the highest rated MSR. The mass flow rate is a function of the main steam pressure. This parameteris l picked up off the plant computer.
)
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. As a result of buoyant plume rise, this pathway is almost always an elevated release.
User inputs Needed to Perform the Calculation
. MSR open using the Parameter Edit screen
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
. Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that indicated by current RCS temperature and pressure.
O) t v 164
I
)
l Number U TMI l
\.j NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No, TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 91 of 164 Main Steam Safety Release (A or B Side)
COLA Calculation Using RMS Problems to Watch For
. *Aain steam line and condenser offgas monitor increases can result from increased RCS activity or increased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs:
l O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23. l RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the i ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the I core. ) l 9 Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if leak rate changes hue been observed.
. If the steam driven EFP (EFP-1) is on, request that the ED switch over to the electrically driven pumps to j p)
( v terminate this release pathway. If this is not possible, request that the ED use the auxiliary boilers to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used.
. Raising the OTSG level to 600 inches will provide additional iodine reduction. Request that the ED consider increasing OTSG level for this purpose e if the MS-V-2 valves are closed, RM-G-26 and RM-G-27 are iso %ted from steam flow and may not be j representative of current main steam conditions. Determine MS-V-2 position from the ED. With the MS-V-2 j valves closed, manual code contingency calculations should be used to calculate offsito doses unless the last known reading on these monitors is known to be representative of current main steam conditions.
i
. Since the COLA does not know whether the open MSR is on the A or B side, and because it uses the j highest of the main steam menitor readings,it can be a conservative calculation. For example,if the open ,
MSR is on the B side (RM-G-27) but it is the OTSG-A that is leaking (RM-G-26), the COLA is assuming that i an A side MSR is open, resulting in what could be a very conservative calculation. Should this type of situation occur, the manual code RMS calculation should be used to see how conservative the COLA is.
. RM-G-26 and RM-G-27 are accident range monitors and are relatively insensitive. As a result, they will tend l
to overestimate offsite doses at low count rates. Normal background from the check sources on these { monitors is typically 30 to 50 cpm. l 1 l
. Remember that if fuel damage has occurred, it cannot get better. j O If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel
' J Damage Class.
l i 8 If this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class.
.v) l i
165 l
l l Number U TMI Q NUCL EAR Radiological Controls Procedure 6610-PLN-4200.02 litle Rewsion No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 92 of 164 Main Steam Safety Release (A or B Side) COLA Calculation . Using RMS l e After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. Effluent samples are not available to this pathway. If fuel damage is suspected, consider asking for an RCS post accident sample to refine the dose projection.
. If post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Scteen, ensure the Fuel Damage Class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default isotopic activities for that damage class instead of the post accident samples that were input. If the spiking fadors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results.
. Monitor met data. If met data shows absolutely no changes for a 30 -45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
p (v). Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get lodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 rninute air sample Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field rcading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual'What If" dose projections for an increase in damage class using the manual code contingency calculation.
. Consider performing a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dose projection if RM-G-26 and RM-G-27 are reading close to background.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? If RM-G 26 and RM-G-27 readings are less than 100 cpm, this will probably be a bounding type calculation. If they are above 100 cpm, this type of calculation should be a pretty good approximation of what the doses could be.
i V 166
l number NUCLEAR Radiological Controls Procedure 6610-PLN-4200,02 Tdle Revrsion No. j TMl Emergency Dose Calculation Manual (EDCM) 9 I EXHIBIT 1 Page 93 of 164 Main Steam Safety Release (A or B Side) COLA Calculation Using Samples How it works: This option cannot be used on the COLA at this time i. s i 167
I' Numbr
-[ U NUCLEAR TMI
, Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 94 of 164 Main Steam Safety Release (A or B Side) Manual Code Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of ,
activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total l activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators. l Additionaliodine reduction can occur if the OTSG levelis raised above 600 inches. These reduction factors are accounted forin the manual code.
. The RMS monitor (RM-G-26 or RM-G27) determines the activity in the main steam. These monitors will respond to both increases in RCS activity and primary-to-secondary leakrate.
. The user inputs the reading on RM-G-26 if an 'A' side MSR is open or the reading on RM-G-27 if a 'B' side MSR open. If MSR's are open on both the A and B side, the highest of the two readings should be used.
. Once the code determines the isotopic concentrations in the main steam line, it uses the mass flow rate out
( the MSR. The mass flow rate is a function of the main steam pressure. The main steam pressure is input i by the user. l
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the j concentrations and doses at distances from the plant.
. As a result of buoyant plume rise, this pathway is almost always an elevated release.
User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. The monitor to be used and the monitor reading.
. The time in minutes since reactor shutdown
. The main steam pressure
. The specific MSR that is open.
. Whether generator level is above 600 inches or not O
! I. Met Data (wind speed, wind direction, and delta t)
V e Release duration 168 l
r l l Number U TMI k,/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200,02 Tide Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 95 of 164 Main Steam Safety Release (A or B Side) Manual Code Calculation Using RMS Problems to Watch For i e Main steam line and condenser offgas monitor increases can result from increased RCS activity or i increased primary-to-secondary leakage, or a combination of the two. If a sudden increase occurs: O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage, Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. 8 Check for signs of increased primary-to-secondary leakage. Check with STA or TSC to determine if leak rate changes have been observed.
. If the steam driven EFP (EFP-1) is on, request ti,at the ED switch over to the electrically driven pumps to terminate this release pathway. If this is not posshle, request that the ED use the auxiliary boilers to drive p) 5 the pump. Ask to switch to the unaffected OTSG until the boilers can be used. l l k/. Raising the OTSG level to 600 inches will provide additional iodine reduction. Request that the ED consider increasing OTSG level for this purpose 1
e if the MS-V-2 valves are closed, RM-G-26 and RM-G-27 are isolated from steam flow and may not be ! representative of current main steam conditions. Determine MS-V-2 position from the ED. With the MS-V-2 l valves closed, manual code contingency calculations should be used to calculate offsite doses unless the last known reading on these monitors is known to be representative of current main steam conditions.
. RM-G-26 and RM-G-27 are accident range monitors and are relatively insensitive. As a result, they will tend to overestimate offsite doses at low count rates. Normal background from the check sou:ces on these monitors is typically 30 to 50 cpm. i e Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input mto the code, the code will reduce the Fuel Damage Class. l 9 ff this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed. e if post accident RCS results are to be input into code, spedj that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input, i V )e Effluent samples are not available for this pathway. If fuel damage is suspected, consider asking for an RCS post accident sample to refine the dose projection. l i 169 l l t
1 l Number
/ U TMI
/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 i
Title Revrsion No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 96 of 164 i 1 Main Steam Safety Release (A or B Side) Manual Code Calculation Using RMS
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure fie!d tasms get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathwa, with the ED.
. Consider performing manual"What If" dose projections for an increase in damage class using the manual
/~N code contingency calculation.
D. Consider performing a manual contingency calculation based on primary-to-secondary leak rate as verification of this type of dose projection if RM-G-26 and RM-G-27 are reading close to background.
. This pathway is alwr;a considered an elevated release. However,if this calculation is run with the EFP running, the code will erroneously classify the EFP release as an elevated release. With the EFP running both cases must be run separately and the doses added together.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? If RM-G-26 and RM-G-27 reaangs are less than 100 cpm, this will probably be a very bounding type calculation, if they are above 100 cpm, this type of calculation should be a pretty good approximation of what the doses could be.
f'%
\
v
)
170
i
)
"~
r @u \ NUCLEAR 1M, Radiological Controls Procedure 6610-PLN-4200.02 l Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 j ( EXHIBIT 1 Page 97 of 164 1 Main Steam Safety Release (A or B Side) Manual Code Calculation Using Samples How it works: This option cannot be used on the manual code at this time l l t i i f f 1 2 i t 171
1 Number i U TMI Q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 , Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 98 of 164 Main Steam Safety Release (A or B Side) Manual Code Calculation Contingency Calculation I How it works: i
. This calculation is very useful for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. If RM-G-26 and RM-G-27 are at background levels, this is a good calculation to perform to verify dose projections made using RMS data.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is equal to the product of the primary-to-secondary leak rate and the activity concentration in the RCS. The source term rate is independent of mass flow through the MSR.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators.
Additionaliodine reduction can occur if the OTSG levelis raised above 600 incher.
'O t
Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has (w }. the ability to use RM-L-1 high orlow seadings to estimate total RCS activity. The user may also specify a total RCS activity.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. As a result of buoyant plume rise, this pathway is almost always an elevaled release.
User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. RM-L-1 high or low readings if letdown is not isolated.
. An estimate of total RCS activity,if known.
. The primary-to-secondary leak rate, in gpm, provided by the STA or TSC.
. The main steam pressure
. The specific MSR that is open
- p. Whether generator level is above 600 inches or not
\b
- h. The time in minutes since reactor shutdown 172
r Numb:r I N TMI Q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No. TMI Emergency Dose Calcalation Manual (EDCM) 9 EXHIBIT 1 Page 99 of 164 Main Steam Safety Release (A or B $1de) Manual Code Calculation l Contingency Calculation
. Met Data (wind speed, wind direction, and delta t)
. Release duration Problems to Watch For
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are ideal for performing "What If" calculations, whcie the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. If the steam driven EFP (EFP-1)is on, request that the ED switch over to the electrically driven pumps to terminate this release pathway, if this is not possible, request that the ED use the auxiliary boilers to drive the pump. Ask to switch to the unaffected OTSG until the boilers can be used.
O \ (d . Raising the OTSG level to 600 inches will provide additional iodine reduction. Request that the ED consider increasing OTSG level for this purpose.
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an RM-L-1 reading is input. Prior to inputting an RM-L-1 reading, venfy that RCS letdown is not secured. If it has been secured (and il typically is during this type of accident), RM-L-1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS activity.
. If RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc)if this is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed. If the dose rate on the sample is a factor of 10 higher than a normal RCS sample, it can be inferred that the RCS activity is 10 times higher than it was prior to the incident. Be aware, entering a total RCS activity will replace damage class default RCS activities and spiking factors.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure impro'. , and the new values are input into the code, the code will reduce the Fuel Damage Class. 9 If this occurs, the user must specify the damage class instead of entering pressure and temperature. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. Effluent samples are not available for this pathway. Consider requesting an RCS sample (normal or post accident) to increase the accuracy of the dose projection.
r v) 173
Number
/ TMI
(,/ ' *** Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergcncy Dos: C:ia:ation Manual (EDCM) 9 EXHlBIT 1 Page 100 of 164 Main Steam Safety Release (A or B Side) Manual Code Calculation Contingency Calculation
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. This pathway is always considered an elevated release. However, if this calculation is run with the EFP running, the code will erroneously classify the EFP release as an elevated release. With the EFP running both cases must be run separately and the doses added together, e Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field f'N teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if f additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. When performing "What If" dose projections, clearly label them as "What If" calculations.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? Depending on the inputs, this type of calculation could go either way.
174
Number U [j ( NUCLEAR TMI Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 101 of 164 Main Steam Safety Release (A or B Side) Quick Code Calculation How it works:
. This code provides a quick, but very conservative, calculation. As such it should only be used by the on-shift RAC, during tM first hour of an emergency, if the COLA Code is not available.
. This code requires an RMS reading forits calculation. If an RMS reading is not available, a manual code contingency calculation must be performed.
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The RCS mixture is assumed to be a 50 to 1 noble gas to iodine ratio and a radioiodine spike of 50. The resulting 1 to 1 noble gas to iodine ratio is very conservative for situations involving core damage.
. The iodine reduction factor for this pathway is 0.5 based on partitioning of iodine in the steam generators. it is accounted for in the code.
/%
l I. The RMS monitor (RM G-26 or RM-G27) determines the activity in the main steam. These monitors will V respond to both increases in RCS activity nnd primary-to-secondary leakrate. The user selects which monitor vllt be used for the calculation.
. The code assumes the mass flow rate out of the MSR to be equivalent to an ADV 100% open at a main steam pressure of 1040 psi.
. The code assumes an 8 hour release duration.
. Once the code determines the isotopic concentrations in the main steam line, it uses the assumed flow rate out of the MSR to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always considered an elevated release.
User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 35 & 36 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
. Met Data (wind speed, wind direction, and delta t)
Since the COLA is not available, all readings must come from control room instrumentation. []. 175
Number [ ) U TMI NUCLEAR (/ Radiological Controls Procedure 6610-PLN-4200,02 Title Revraion No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 102 of 164 Main Steam Safety Release (A or B Side) Quick Code Calculation Problems to Watch For
. The IREO RAC should not use this code. The full manual code should be used when the IREO organization is activated.
. This calculation produces a bounding estimate of offsite dose, particularly for core damage situations due to the 1 to 1 noble gas to iodine ratio assumed in the RCS. When communicating the results of this code to -
the ED, the conservative nature of the results should be emphasized. 1
. As soon as time is available, a manual code RMS calculation should be performed so a dose estimate using more representative plant parameters can be obtained.
. If the MS-V-2 valves are closed. RM-G-26 and RM-G-27 are isolated from steam flow and may not be representative of current main steam conditions. Determine MS-V-2 position from the ED. With the MS-V-2 valves closed, manual code contingency calculations should be used to calculate offsite doses unless the pg last known reading on these monitors is known to be representative of current main steam conditions.
% . Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get l iodine samples even if E-520 dose rates are low. )
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading l should be performed prior to issuing a PAR beyond 10 miles. '
. Periodically verify the release duration and pathway with the ED.
. Label all Quickcode calculations as "Quickcode".
i O) 176 I
r Nurnber
/
N NUCLEAR TMI ( Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No, TMl Emergency Dose Calculation ManualjEDCM) 9 EXHIBIT 1 Page 103 of 164 Station Vent COLA Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is a combination of primary system iodine retention (0.4) and charcoal filtration (0.01). The total iodine reduction factor is then (0.4)(0.01) = 0.004.
. The RMS monitor (RM-A-8 Gas, RM-A-8 lodine, or RM-A-8 high) determines the activity leaving via this pathway. These monitor 4 will respond to both increases in RCS activity and increased leakage into the auxiliary / fuel handling building.
. The logic the COLA uses in picking which monitor to perform the calculation is as follows:
/~~N e if RM-A-8 High reads less than 100, the COLA is using the reading on RM-A-8 Gas to compute the (V) noble gas source term.
l e if RM-A-8 High reads less than 100, the COLA is using the difference between the current reading i on RM-A-6 lodine and the previous reading on RM-A-8 iodine compute the iodine source tenn. e If RM-A-8 High reads > 100 cpm, the COLA is using RM-A-8 High to compute the noble gas and j iodine source terms. ; l
. Once the COLA determines the isotopic concentrations leaving via station vent, it uses the flow rate to I determine the uCi/sec leaving the plant (source termt ;
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the l concentrations and doses at distances from the plant. I I
l
= This pathway can be a ground level release, an elevated release, er a mixture of the two depending on the i flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present. ,
i User inputs Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours l . Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
. Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that indicated by current RCS temperature and pressure.
O O 1 177 L
Number U i [j
\ NUCLEAR TMl Radiological Controls Procedure 6610-PLN-4200.02 Title Revrsion No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 104 of 164 Station Vent COLA Calculation Using RMS Problems to Watch For j i
. Station vent monitor increases can result from increased RCS activity or RCS leakage in the Auxiliary / Fuel Handling Building, or a combination of the two. If a sudden increase occurs:
O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, l RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the i ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. l 8 Check for signs ofincreased leakage to the Auxiliary / Fuel Handling Building. Check with STA or l TSC to determine if leak rate changes have been observed. j
. If RM-A-8 High is s 100 cpm, RM-A-8 lodine is used to calculate the iodine source term. The iodine source j term is not ratioed in from the RCS noble gas to iodine ratio. The isotopic distribution of the iodines is the '
same as in the RCS.
- d. The RM-A-8 iodine calculation assumes that 4 minutes have ehpsed since the current and previous iodine channel readings. Due to differences in the timing of PPM files and COLA calculations, there can be variability in thyroid dose estimate (factor of 2) during each 15 minute period. Use the highest of the three estimates performed each 15 minutes.
. If RM-A-8 High is s 100 cpm, and RM-A-8 lodine is offscale high, the iodine source term will not be calculated correctly, in this situation, the manual RMS calculation fcr station vent should be performed.
. If RM-A-8 High is > 100 cpm, the iodine source term is scaled in from the noble gas to lodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel Damage Class. 9 If this occurs, the user must edit the damage class using the Parameter Edit Screen to lock in the higher damage class. 8 After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed. l
. If post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel Damage Class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit Screen are set to 1, if the COLA is using a damage class greater than 1, the COLA will use the default p i isotopic activities for that damage class instead of the post accident samples that were input. If the spiking I
e factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results. y /. t 178
5 1 { 1 Number f (~N ( ,/
@uNUCLEAR TM, Radiological Controls Procedure 6610-PLN-4200.02 l Title Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 105 of 164 . l Station Vent j COLA Calculation i i Using RMS i i e Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications. l
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field j teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if I additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
Periodically verify the release duration and pathway with the ED. {}.
\ f
%'. Consider getting an effluent sample from this pathway (MAP-5 or pre-filter marine;:i) when conditions appear to have stabilized.
. Consider performing manual'What If" dose projections for increases in damage class.
. If the charcoal filters become degraded and are believed to be operating at less than normal efficiency, the manual code RMS calculation should be used for this pathway. The TSC can provide information on charcoal filter efficiency and degradation.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually cot..d be? This type of calculation should be a pretty good approximation of what the doses could be.
I V 179
I I
- Number U TMI
\. j NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 I Title Revsson No.
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 106 of 164 Station Vent COLA Calculation Using Samples How it works:
. Effluents leaving via Station Vent are sampled using a pre-filter marinelli sample
. The user enters the positive noble gas and iodine isotope concentrations from the analysis into the Station Vent Sample Edit Screen.
. If a MAP-5 sample is used, the manual code must be used to perform the calculation, since the COLA will not produce a noble gas source term.
. Once the COLA determines the isotopic concentrations leaving via Station Vent, it uses the flow rate from the plant computer to determine the uCl/sec leaving the plant (source term),
e Having developed the source term, the COLA uses the meteorological dispersion model to calculate the fg concentrations and doses at distences from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor buildin0 purge exhaust and the meteorological conditions present. l User inputs Needed to Perform the Calculation
- Sample results need to be entered on the Station Vent Sample Edit Screen e Release dt, ration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad 1
I
)
180
i l l Number l
\ U TMI l Q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 l TMie Revision No )
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 107 of 164 Station Vent COLA Calculation i Using Samples ' l Problems to Watch Fof
. The user must ensure that the sample results are representative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should make the Station Vent Sample Edit Screen inactive and resume COLA RMS calculations.
. MAP-5 samples cannot be used solely with the COLA code since noble gases are not measured. The COLA code will not produce a noble gas source term from a MAP-5 sample. MAP-5 samples must be calculated using the manual codes to get the correct TEDE dose. The MAP-5 sample should still be input into the COLA code since the thyroid doses produced by the COLA will still be accurate.
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up pI i
. with TEDE projections. Roughly 3% of the CDE dose is included in tne TEDE dose. Ensure field teams get
'd iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the *LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for increases in damage class using a manual code contingency calculation.
. Always communicate the uncertainty of the dose projection is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
C\ (-) 181
r
~
Number
/ U TMI
( (/ NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Titie Revtsson No. l TMI Emergency Dose Calculation Manual (EDCM) 9 1 l EXHIBIT 1 Page 108 of 164 Station Vent - ! Manual Code Calculation l Using RMS - How it wcrks:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is a combination of primary system iodine retention (0.4) and charcoal filtration efficiency specified by the user (E). The totaliodine reduction factor is then (0.4)(1-E),
normally = 0.004.
. The RMS monitor (RM-A-4,6, or 8 Gas, RM-A-4,6, or 8 lodine, or RM-A-8 high) determines the activity leaving via this pathway. These monitors will respond to both increases in RCS activity and increased leakage into the auxiliary / fuel handling building.
The user selects whether RM-A-4, RM-A-6, RM-A-8 or RM-A-8 High is used for the calculation: [ .
\__ O If RM-A-8 or RM-A-8 High is selected, the code wili ask the user if RM-A-8 lodine is on scale. If the answer is yes, the code will calculate the iodine source term base on the current and previous iodine channel reaoings specified by the user. If RM-A-8 iodine is offscale, the iodine source term is scaled in from the noble gas to iodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway.
e if RM-A-4 or RM-A-6 are selected. The code assumes the iodine channels are onscale and iodine i channels must be input to get an iodine source term. ! O If RM-A-4 or RM-A-6 are selected, orif RM-A-8 lodine is offscale, the code will ask the user if the charcoal filters are operational. If they are, the code will prompt the user to input their efficiency. The efficiency should be input as 0.99 unless information frem the TSC indicates another value should be used.
. Once the code determines the isotopic concentrations leaving via station vent, it uses the flow rate to determine the uCl/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated re! ease, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
tD , L) 182
gg ,
'Q NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No.
TMi Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 109 of 164 Station Vent Manual Code Calculation Using RMS User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 38 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class I
. The monitor to be used and the monitor reading.
. The time in minutes since reactor shutdown
. Station Vent flowrate
. The Reactor building purge flowrate (for adjacent plume rise) f\ j. Charcoal filter operability and efficiency j
t,,%./ !
. Met Data (wind speed, wind direction, and delta t)
]
l
. Release duration l Problems to Watch For
. Station Vent monitor increases can result from increased RCS activity or increased RCS leakage to the Aux /FHB, or a combination of the two. If a sudden increase occurs:
O Check for signs of increased RCS activity. Coincident increases on RM-L-1, RM-G-22, RM-G-23, RM-G-6, or RM-G-7 may indicate increased RCS activity from fuel damage. Also check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. O Check for signs of increased leakage to the Auxiliary / Fuel Handling Building. Check with STA or TSC to determine if leak rate changes have been observed.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class. 9 If this occurs, the user must specify the damage class instead of entering pressure and temperature.
/ 9 After doing this, the user must watch RCS temperature and pressure to ensure that damage class
( does not increase beyond the damage class assumed. V 183
r-Number
' N TMl j bl MN Radiological Controls Procedure 6610-PLN-4200.02 l
! Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 110 of 164 Station Vent Manual Code Calculation Using RMS
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to t' 1 TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of' .e CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates aie low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading
(] should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual"What If" dose projections for increases in damage class using a manual code contingency calculation.
. Consider getting an effluent sample from this pathway (MAP-5 or pre-filter marinelli) when conditions appear to have stabilized.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
l O, t
\j 184
Number [ U TMl G NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 111 of 164 Station Vent Manual Code Calculation Using Samples How it works:
. Effluents leaving via station vent are sampled using a MAP-5 sample or a pre-filter marinelli sample e if a pre-filter mannelli sample is used, the user enters the positive noble gas and iodine isotope concentrations from the analysis when prompted by the code.
. If a MAP-5 sample is used, the user enters the positive iodine isotope concentrations from the analysis when prompted by the code. Since MAP-5 samples do not provide noble gas results, manual code will scale in the noble gases. It assumes that the noble gas to iodine ratio leaving Station Vent is the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway.
. The iodine reduction factor for this pathway is a combination of primary system iodine retention (0.4) and (p) charcoal filtration efficiency specified by the user (E). The total iodine reduction factor is then (0.4)(1 E),
normally = 0.004.
%/
. Once the code determines the isotopic concentrations leaving via Station Vent, it uses the Station Vent flow rate input by the user to determine the uCi/sec leaving the plant (source term),
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
User inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 38 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The code will prompt the user to input values for RCS temperature and pressure or Fuel Damage Class values. These must be input but will have no bearing on the calculation since effluent concentrations leaving the plant are being determined by isotopic sample analysis.
. Sample results from Station Vent and the time of the sample
. Station vent flow rate
. The reactor building purge flow rate (t adjacent plume rise)
. Charcoal filter operability and efficiency
. Met Data (wind speed, wind direction, and delta t)
-. Release duration 185
Number l U TMl NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Titie Revson No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 1'.2 of 164 Station Vent Manual Code Calculation Using Samples Problems to Watch Foj
. The MAP-5 sample panel may present a dose rate hazard to personnelif the release rato is high or if it has been running for a long time with the same cartridges.
l
.- The user must ensure that the sample results are rep:3sentative of current pl&at conditions. If plant conditions change significantly after the sample was obtained, the user should begin using COLA or manual cede RMS calculations.
If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles
. Rei nnber that the CDE contnbutes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get p iodine samples even if E-520 dose rates are low.
5 V . Depending on the physical form of iodine in the field, the *LLD" for thyroid dose rates measured by field teams could be as Ngh as 5 mrem /hr based on a 10 minute air sample. Consider 5nger sampling times if additional sensitivityis desired.
. Periodically verify the release duration and pathway with the ED.
. Consider performiag manual %Vhat If" dose projection for increases in damage class.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that i know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
l O U 186
U TMI WUCLEAR t Radiological Controls Frocedure 6610-PLN-4200.02 Title RevrSon No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 113 of 164 Station Vent Manual Code Calculation Contingency Calculatica How it works: The manual code does not suppod this calculation .11 th.S time. I l
\
187
Number (d Tme II NUCLEAN TMI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBlY 1 Page 114 of 164 Station Veni Quick Code Calculation How it works:
- This code provides a quick, but very conservative, calculation. As such it should only be used by the on-shift RAC, during the first hour of an emergency,if the COLA Code is not available. . This code requires an RMS reading forits t alculation. If an RMS reading is not available, a manual code .
contingency calculation must be performed. l i
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of !
activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The RCS mixture is assumed to be a 50 to 1 noble gas to iodine ratio and a radiciodine spike of 50. The resulting 1 to 1 noble gas to iodine ratio is y_e.y er conservative for situations involving core damage.
. The iodine reduction factor for this pathway is a combination of primary system iodine retention (0.4) and charcoal filtration (0.01). The total iodine reduction factor should be (0.4)(0.01) = 0.004. However, this code ,
p is using a primary system retention of 0.1. As a result, thyroid doses for this pathway should be multiplied by ' \j
. The RMS monitor (RM-A-8 or RM-A-8 High) determines the activity leaving via this pathway. These monitors will respond to both increases in RCS activity and increrssed leakrate into the Aux / Fuel Handling Building. The user selects which mon v will be used for the calculation. Generally, the highest range monitor reading > 100 should be used. . The code assumes a Station Vent flow of 80,000 cfm. Dose results may t'e ratioed by the actual cfm (i.e. if the actual flow is only 40,000 cfm, the dose projection is reduced by %). . The code assumes an 8 hour release duration. . Once the code determines the isotopic concentrations leaving via Station Vent, it uses the flow rate to determine the uCi/sec leaving the plant (source term). . Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant. . This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the meteorological conditions present.
User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 38 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA) The monitor to be used and the monitor reading. Y ),.
. Met Data (wind speed, wind d:'ction, and delta 1) 188
l l l Number [
\ /
\ U NUCLEAR TMI F'~4iological Controls Procedure 6610-PLN-4200.02 Tdle Rtvision No.
TMI Emergency Dese Calculation Manual (EDCM) 9 EXHIBIT 1 Page 115 of 164 Station Vent Quick Code Calculation Problems to Watch For
. Since the COLA is not available, all readings must come Tom control room instrumentation.
. The IREO RAC should not use this code. The full manual code should be used when the IREO organization is activated.
. The iodine reduction factors for this code are non-conservative. Thyroid doses produced by this code should be multiplied by 4, particularly for RCS iodine spike situations.
. This calculation produces a bounding estimate of offsite dose, particularly for core damage situations due to the i to 1 noble gas to iodine ratio assumed in the RCS. When communicating the results of this code to the ED, the conservative nature of the resub:, should be emphasized.
. The user should check the sum of the actual flow at FR-149 and FR-150 to determine if the dose needs to
[ N. be adjusted based on the 80,000 cfm flow assumed in the calculation. kd I
. As soon as time is available, a manual code RMS calculation should be performed so a dose estimate using I mcre representative plant parameters can be obtained. l l
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /br based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Label all Quickcode calculations as "Quickcode" l
O O 189
I l t Number U TMl ( NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 116 of 164 Reactor Building Pr i Exhaust COLA Calci on Using F S How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution. . The iodine reduction factor for this pathway is a combination of reduction by natural processes (0.4) or reduction by reactor building spray (0.03) and charcoal filtration (0.01). The total iodine reduction factor with no reactor building spray is then (0.4)(0.01) = 0.004. The total iodine reduction factor with reactor building spray is then (0.03)(0.01) = 0.0003. . The RMS monitor (RM-A-9 Gas, RM-A-9 lodine, RM-A-9 High, or RM-G-24) determines the activity leaving via this pathway. These monitors will respond to both increases in RCS activity and increased leakage into the reactor building.
O
. The logic the COLA uses in picking which monitor to perform the calculation is as follows:
O If RM-A-9 High and RM-G-24 read s100, the COLA is using the reading on RM-A-9 Gas to compute the noble gas source term. 8 If RM-A 9 High and RM-G-24 read s100, the COLA is using the difference between the current reading on RM-A-9 lodine and the previous reading on RM-A-9 iodine compute the iodine source term 9 if RM-A-9 High reeds > 100 cpm and RM-G-24 reads s100 mR/hr, the COLA is using RM-A-9 High to compute the noble gas and iodine source terms. O If RM-G-24 reads > 100 mR/hr, the COLA is using RM-G-24 to compute the noble gas and iodine source temis
. Once the COLA determines the isotopic concentrations leaving via the reactor building purge exnaust, it uses the flow rate to determine the uCi/sec leaving the plant (source term). . Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant. . This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
User Inouts Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours . Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad 190
Number J
\ N TMI Q NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02
- Title Revision No.
l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 117 of 164 Reactor Building Purge Exhaust l COLA Calculation Using RMS Problems to Watch For i e Fuel Damage Class using the Parameter Edit screen the damage class is believed to be other than that indicated by current RCS temperature and pressure. l
. Reactor Building Purge Exhaust monitor increases can result from increased RCS activity or RCS 'eakage in j the reactor building, or a combination of the two. If a sudden increase occurs:
O Check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. 8 Check for signs of increased leakage to the Reactor Building. Check with STA or TSC to determine if leak rate changes have been observed.
. if RM-A-9 High is s 100 cpm, RM-A-9 lodine is used to ca.culate the iodine source term. The iodine source p)
( v term is not ratioed in from the RCS noble gas to iodine ratio. The isotopic distribution of the iodines is the same as in the RCS.
. The RM-A 9 iodine calculation assumes that 4 minutes have elapsed since the current and previous iodir.e I channel readings. Due to differences in the timing of PPM files and COLA calculations, there can be variability in thyroid dose estimate (factor of 2) during each 15 minute period. Use the highest of the three j estimates performed each 15 minutes. l
= If RM-A-9 High is s 100 cpm and RM-A 9 lodine is offscale high, the iodine source term will not be calculated correctly. In this situation, the manual RMS calculation for the reactor building purge exhaust should be performed.
. if RM-A 9 High or RM-G-24 is > 100, the iodine source term is scaled in from the noble gas to iodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred, but RCS temperature and pressure improve, the COLA will reduce the Fuel Damage Class. 9 If this occurs, the user must edit the damage chss using the Parameter Edit Screen to lock in the higher damage class. O After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. if post accident RCS results are to be input into the Reactor Coolant Baseline Activity Edit Screen, ensure the Fuel Damage Class is set to 1 on the Parameter Edit Screen and the spiking factors in the Sample Edit /p\ Screen are set to 1. If the COLA is using a damage class greater than 1, the COLA will use the default i
l ( isotopic activities for that damage class instead of the post accident samples that were input. If the spiking factors are not set to 1, the COLA will erroneously apply the spiking factors to the input sample results. 191
Number U TMI h WCN Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 118 of 164 l Reactor Building Purge Exhaust l COLA Calculation l Using RMS
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Venfy met data with EACC and/or Control Room indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical fonn of iodine in the field, the "LLD' for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
e if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles. (]. Periodically verify the release duration and pathway with the ED. (O e 1 Consider performing manual *What If' dose projections for increases in damage class.
. Consider getting an effluent sample from this pathway (MAP-5, pre-fiaar marinelli, or CATPASS as applicable) when conditions appear to have stabilized.
e If the charcoal filters become degraded and are believed to be operating at less than normal efficiency, the manual code RMS calculation should be used for this pathway. The TSC can provide information on charcoal filter efficiency and degradation.
. if no significant releases a:a occurring through station vent, consider maximizing station vent flow to increase dispersion from adjacent plume momentum.
. Always communicate the uncertainty of the dose projeciion. Is it a bounding calculation such that I know doses cani be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
i v) 192
r Number
) $UNUCLEAA TMl Q Radiological Controls Procedure 6610 PLN-4200.02 Title Revision ;40.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 119 of 164 Reactor Building Purge Exhaust COLA Calculation Using Samples How it works:
. Effluents leaving via the reactor building purge exhaust are samp!ed using a pre-filter marinelli sarnple )
. The user enters the positive noble gas and lodine isotope concentrations from the analysis into the Reactor Building Purge Exhaust Sample Ed:t Screen.
. If a MAP-5 sample is used, the manual code must be used to perform the calculation, since the COLA will not produce a noble gas source term.
. Once the COLA determines the isotopic concentrations leaving via the reactor building purge exhaust, it uses the flow rate from the plant computer to determine the uCi/sec leaving the plant (source term).
l
. Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
]
. This pathway can be a ground level re! ease, an elevated release, or a mixture of the twu depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
User Inouts Needed to Perform the Calcu,at;6n
. Sample results need to be entered on the Station Vent Sample Edit Screen
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad L
() 193
Wumber U (Q) NuctrAA TMI Radiological Controls Procedure 6610-PLN-4200.02 Tdie Revson No TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 120 of 164 I Reactor Building Purge Exhaust COLA Calculation l Using Samples J Problems to Watch For
. The user must ensure that the sample results are representative of current plant conditions. If plant l cot. - ~ ns change significantly after the sample was obtained, the user snould make the Reactor Building l Purrexhaust Sample Edit Screen inactive and resume COLA RMS calculations.
. MAP-5 samples cannot be used solely with the COLA code since noble gases are not measured. The i COLA code will not produce a noble gas source term from a MAP-5 sample. MAP-5 samples must be calculated using the manual codes to get the correct TEDE dose. The MAP-5 sample should still be input i into the COLA code since the thyroid doses produced by the COLA will still be accurate. )
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up f].j with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get (V iodine samples even if E-520 dose rates are low.
. Depending on th6 physical form of lodine in the field, the "LLD" for thyroid dose rates measured by field l teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if '
additional sensitivity is desired.
. If dose projec' ions show PAG's are exceeded at 10 miles, validation of the dose project on from field reading !
should be performed prior to issuing e PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED. !
. Consider performing manual *What If" dose projections for increases in damage class using a manual code contingency calculation.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know !
doses can't be any higher than this? Or,is this the best 1pproximation if what I believe offsite doses actually t.ould be? This type of calculation should be a pretty good approximation of what the doses could be.
. If no significant releases are occurring through station vent, consider maximizing station vent flow to !
increase dispersion from adjacent plume momentum. 1 NY 5 194
I' l l Number
- /N I
(V l TMI "N Radiological Controls Procedure 6610-PLN-4200.02 D' Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 121 of 104 Reactor Building Purge Exhaust Manual Code Calculation Using RMS How it works:
. The isotopic distribution of activity leaving tne plant is assumed to be Lie same as the isotopic distobution of activity in the RCS after adjusting for iodine losses (redudion factors) applicable to this pathway. The total activity in the RCS is not important to this calculation, just the RCS isotopic distribution.
. The iodine reduction factor for this pathway is a combination of reduction by natural processes (0.4) or reduction by reactor building spray (0.03) and charcoal filtration (E) specified by the user. The total iodine s reduction factor with no reactor building spray is then (0.4)(1-E), normally 0.004. The totaliodine reduction factor with reactor building spray is then (0.03)(1 E), normally = 0.0003.
. The RMS monitor (RM-A-9 Gas, RM-A-9 lodine, RM-A-9 High, or RM-G-24) determines the activity leaving via this pathway. These monitors will respond to both increases in RCS activity and increased leakage into the reactor building.
\. The user selects whether RM-A 9, RM-A-9 High, or RM-G-24 is used for the calculation. The logic the code
\ uses in picking which monitor to perform the calculation is as follows:
O If RM-A-9, RM-A 9 High, or RM-G-24 is selected, the code will ask the user if RM-A 9 lodine is on scale. If the answer is yes, the code will calculate the lodine source term based on the current and previous iodine channel readings specified by the user. If RM-A-9 iodine is offscale, the iodine source term is scaled in from the noble gas to iodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway, e if RM-A-9, RM-A-9 High, or RM-G-24 are selected, and RM-A-9 lodine is offscate, the code will ask ; the user if the charcoal filters are operational. If they are, the code will prompt the user to input their l efficiency. The efficiency (E) should be input as 0.99 unless information from the TSC indicates ' another value should be used.
. Once the code determines the isotopic concentrations leaving via the reactor building purge exhaust, it uses the flow rate input by the user to determine the uCi/see leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
O' 195
y Numtnr U TMI l C/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 l Title Revision No. l TMI Emergency Dose Calculation Manual (EDCM) 9 1 EXHIBIT 1 Page 122 of 164 Reactor Building Pu 'ge Exhaust Manual Code Calculation Using RMS User Inouts Needed to Perform the Ca!culation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. Purge valve position (open if performing this calculation)
. The monitor to be used and the monitor reading.
. The time in minutes since reactor shutdown
. Station vent flowrate (for adjacent plume rise)
A) . (V The reactor building purge flowrate e RB spray on or off
. Charcoal filter operability and efficiency
. Met Data (wind speed, wind direction, and delta t)
. Release duration Problems to Watch For
. Reactor Building Purge Exhaust monitor increases can result from increased RCS activity or RCS leakage in the reactor building, or a combination of the two. If a sudden increase occurs: j 0 Check the ED/ESD Screen for changes in Fuel Damage Class based on thermal hydraulic conditions in the core. )
I 9 Check for signs of increased leakage to the Reactor Building. Check with STA or TSC to determine I if leak rate changes have been observed.
. Remember that if fuel damage has occurred, it cannot get better. .
1 O If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class.
- p. 9 If this occurs, the user must specify the damage class instead of entering pressure and temperature.
196
[ l l l ! Number l
\ U TMI - (j NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Tale Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 123 of 164 Reactor Building Purge Exhaust Manual Code Calculation Using RMS ; 9 After doing this, the user must watch RCS temperature and pressure to ensure that damage class does not increase beyond the damage class assumed.
. If post accident RCS results are to be input into code, specify that the Fuel Damage Class as i and there has been no power transient when the code asks. If the code is using a damage class greater than 1, the ,
code will use the default isotopic activities for that damage class Instead of the post accident samples that I were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if A additional sensitivity is desired.
(
\ . If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Consider performing manual "What If" dose projections for increases in damage class using a manual code contingency calculation
. Consider getting an effluent sample from this pathway (MAP-5, pre-filter marinelli, or CATPASS as applicable) when conditions appear to have stabilized.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
I 197
I i l Number l U TMI
\j NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 124 of 164 Reactor Building Purge Exhaust Manual Code Calculation Using Samples How it works:
. The MAP-5 sample panel may present a dose rate hazard to personnelif the release rate is high orifit has been running for a long time with the same cartridges.
. Effluents leaving via the reactor building purge exhaust are sampled using a MAP-5 sample or a pre-filter marinelli sample.
. If a pre-filter marinelli sample is used, the user enters the positive noble gas and iodine isotope concentrations from the analysis when prompted by the code.
. If a MAP-5 sarnple is used, the user enters the positive iodine isotope concentrations from the analysis when prompted by the code. Since MAP-5 samples do not provide noble gas results, manual code will scale in the noble gases. It assumes that the noble gas to lodine ratio leaving the reactor building purge exhaust is p the same as thu isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway.
( N
. The iodine reduction factor for this pathway is a combination of reduction by natural processes (0.4) or reduction by reactor building spray (0.03) and charcoal filtration (E) specified by the user. The total iodine reduction factor with no reactor building spray is then (0.4)(1-E), normally 0.004. The totaliodine reduction factor with reactor building spray is then (0.03)(1-E), normally = 0.0003
. Once the code determines the isotopic concentrations leaving via the reactor building purge exhaust, it uses the reactor building purge exhaust flow rate input by the user to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
O V I 1 198
Number ! U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tale Reymon No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 125 of 164 Reactor Building Purge Exhaust Manual Code Calculation Using Samples User inputs Needed to Perfom1 the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The code will prompt the user to input values for RCS temperature and pressure or Fuel Damage Class values. These must be input but will have no bearing on the calculation since effluent concentrations leaving the plant are being determined by isotopic sample analysis.
. Sample results from the reactor building purge exhaust and the time of the sample
. Station Vent flowrate (for adjacent plume rise)
. The Reactor Building purge flowrate (g
Charcoal filter operability and efficiency (normally 0.99) ( Q ). l
. Reactor building spray status
. Met Data (wind speed, wind direction, and delta t)
. Release duration Problems to Watch For
. The user must ensure that the sample results are representative of current plant conditions. If plant conditions change significantly after the sample was obtained, the user should begin using COLA or manual code RMS calculations.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD' for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles
. Periodically verify the release duration and pathway with the ED.
O \.
! Consider performing manual What If' dose projection for increases in damage class.
V 199
1 i Number TMl NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 { Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 . EXHlBIT 1 Page 126 of 164
. Always communicate the uncettainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what i believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
a j l i
\
200 I
Number i E8 TMl
'Q NUCLEAfr Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 127 of 164 Reactor Building Purge Exhaust Manual Code Calculation Contingency Calculation How it works:
. This calculation is very useful for performing "What If" calculations, whera the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is equal to the product of the totalleakage into the reactor building and the activity concentration in :
the RCS, after adjusting for iodine losses (reduction factors) applicable to this pathway. 1
. The iodine reduction factor for this pathway is a combination of reduction by natural processes (0.4) or l reduction by reactor building spray (0.03) and charcoal filtration (E) specified by the user. The total iodine I reduction factor with no reactor building spray is then (0.4)(1-E), normally 0.004. The total iodine reduction l factor with reactor building spray is then (0.03)(1-E), normally = 0.0003, Unlike RMS calculations, this calculation is very dependent on RCS total activity. As a result, the code has f].
r the ability to use RM-L-1 high or low readings to estimate total RCS activity. The user may also specify a V. total RCS activity.
. Once the code determines the isotopic concentrations in the reactor building, it uses the reactor building purge exhaust flow rate to determine the uCi/sec leaking from the building (source term).
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. RCS temperature and pressure or the user specified Fuel Damage Class
. RM-L 1 high or low readings if letdown is not isolated.
. An estimate of total RCS activity,if known.
. The total RCS leakage into the reactor building, if known. The STA or TSC should provide.
. The time in minutes since reactor shutdown m
I . Charcoal filter efficiency (normally 0.99) d
. The status of reactor building spray (on or off) 201
Nurnber
/ N TMI l j NUCLEAR Radiological Controls Procedure 6610 PLN 4200.02 Title Revision No.
l TMi Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 128 of 164 Reactor Building Purge Exhaust l Manual Code Calculation Contingency Calculation
. The status of the purge valves (open if doing this calculation)
. Met Data (wind speed, wind direction, and delta t) and release duration Problems to Watch For
. Of all calculations available, contin 0ency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available. Contingency calculations are ideal for performing "What If" calculations, where the precise response of an RMS monitor to a change in plant conditicos cannot be predicted.
. This type of calculation is very dependent on total RCS activity. The code will adjust RCS total activity if an RM-L-1 reading is input. Prior to inputting an RM-L-1 reading, verify that RCS letdown is not secured. If it has been secured (and it typically is during this type of accident), RM-L-1 readings should not be used unless the user is sure that the last reading is representative of current RCS conditions. Note that due to travel time in the sample lines, RM-L-1 may take up to 30 minutes before responding to a change in RCS (VO) activity.
. If RM-L-1 readings are not available, the code will permit the user to enter a total RCS activity (uCi/cc)if this is known. It generally is not. An example of when this could be used is if an RCS sample has just been pulled but has not been analyzed. If the dose rate on the sample is a factor of 10 higher than a normal RCS sample, it can be inferred that the RCS activity is 10 times higher than it was prior to the incident. Be aware, entering a total RCS activity will replace damage class default RCS activities and spiking factors.
. The code will prompt the user for the total RCS leakage into the reactor building. If the leakage is less than 63,000 gallons (one RCS volume), the actualleakage should be input. If the leakage is more than 63,000 gallons, a maximum of 63,000 gallons should be used. Inputting more than 63,000 gallons will overestimate the source term because it will put more activity in the building than is available for release from the core.
. If the total RCS leakage is not known, the code will assume the entire RCS volume (63,000 gallons) has been released to the building using a damage class 4. If total RCS leakage is not known and the damage class is other than 4, the user should respond that the totalleakage is known and specify the leakage as 63,000 gallons.
. Remember that if fuel damage has occurred, it cannot get better.
O If damage has occurred but RCS temperature and pressure improve, and the new values are input into the code, the code will reduce the Fuel Damage Class. O If this occurs, the user must specify the damage class instead of entering pressure and temperature. p S After doing this, the user must watch RCS temperature and pressure to ensure that damage class I i does not increase beyond the damage class assumed. t 202
Nun ar i EI TMI []
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02
% tie Revision No l TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 129 of 164 i
Reactor Building Purge Exhaust Manual Code Calculation l Contingency Calculation e if post accident RCS results are to be input into code, specify that the Fuel Damage Class as 1 and there l has been no power transient when the code asks. If the code is using a damage class greater than 1, the code will use the default isotopic activities for that damage class instead of the post accident samples that were input.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading n, should be performed prior to issuing a PAR beyond 10 miles.
I \ kj . Periodically verity the release duration and pathway with the ED.
. When performing "What If" dose projections, clearly label them as "What If" calculations
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? Depending on the inputs, this type of calculation could go either way
. Consider getting a MAP-5, pre-filter /marinelli, or CATPASS sample from this pathway to increase the accuracy of the dose projection. The CATPASS system takes 45 minutes to warm up so direct the OSC to start the warm up as soon as you think you may want this type of sample.
I l v) 203
F Number U TMl
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revision No.
TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 130 of 164 l Reactor Building Purge Exhaust Quick Code Calculation l How it works:
. This code provides a quick, but very conservative, calculation. As such it should only be used by the on-shift RAC, during the first hour of an emergency, if the COLA Code is not available.
. This code requires an RMS reading for its calculation. If an RMS reading is not available, a manual code contingency calculation must be performed.
. The isotopic distnbution of activity leaving the plant is assumed to be the same as the isotopic distribution of activity in the RCS after adjusting for iodine losses (reduction factors) applicable to this pathway. The RCS mixture is assumed to be a 50 to 1 noble gas to iodine ratio and a radioiodine spike of 50. The resulting 1 to 1 noble gas to iodine ratio is ve_ry conservative for situations involving core damage.
. The iodine reduction factor for this pathway is 0.1 for charcoal filtration. This is less than the 0.01 factor used by the COLA and manual code. This code also does not use the reduction by natural processes (0.4)
[O 4 used by the COLA and manual code. Reactor building spray is assumed to be off.
. The RM3 monitor (RM-A-9, RM-A-9 High, or RM-G-24) determines the activity leaving via this pathway.
These monitors will respond to both increases in RCS activity and increased leakrate into the reactor building. The user selects which monitor will be used for the calculation. Generally, the highest range monitor reading > 100 should be used.
. The code assumes a reactor building purge exhaust flow of 10,000 cfm. Dose results may be ratioed by the actual cfm (i.e. if the actual flow is only 5,000 cfm, the dose projection is reduced by %).
. The code assumes an 8 hour release duration.
. Once the code determines the isotopic concentrations leaving via the reactor building purge exhaust, it uses the flow rate to determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the code uses the meteorological dispersion model t, calculate the concentrations and doses at distances from the pisnt.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the meteorological conditions present.
User Inouts Needed to Perform the Calculatign I Plant data inputs are available on Area 38, Group 37 of tiie PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
g > Met D'ata (wind speed, wind direction, and delta t)
%./
. Since the COLA is not available, all readings must come from control room instrumentation.
204
Number N TMI V NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 131 of 164 Reactor Building Purge Exhaust Quick Code Calculation Problems to Watch For
. The IREO RAC should not use this code. The full manual code should be used when the IREO organization is activated.
. This calculation produces a bounding estimate of offsite dose, particularly for core damage situations due to the 1 to 1 noble gas to iodine ratio assumed in the RCS and the lower iodine reduction factors used by this code. When communicating the results of this code to the ED, the conservative nature of the results should be emphasized.
. The user should check the actual flow at FR-909 or FR-148A to determine if the dose needs to be adjusted based on the 10,000 cfm flow assumed in the calculation.
. As soon as time is available, a manual code RMS calculation should be performed so a dose estimate using more representative plant parameters can be obtained.
Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up (O 'd
). with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the *LLD" for thyroid dose rates measured by field t teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times i!
additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
- Periodically verify the release duration and pathway with the ED.
. Label all Quickcode calculations as "Quickcode" f%
t V) 205
I i l Number ( bu NUCL EAR
,M ,
Radiological Controls Procedure 6610-PLN-4200.02 Title Recsson No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 132 of 164 ESF Fuel Handling Building COLA Celculation Using RMS How it works:
. The isotopic distribution of activity leuing the plant is assumed to be the same isotopic distribution of activity as a fuel assembly 72 hours after shutdown from full power, after adjusting for iodine losses (reduction factors) applicable to this pathway. The total activity in the fuel assembly is not important to this calculation, just the isotopic distribution.
. The iodine reduction factor for this pathway is a combination of reduction by decontamination in the spent fuel pool (0.01) and charcoal filtration (0.01). The total iodine reduction factor is then (0.01)(0.01) = 0.0001.
. The RMS monitor (RM-A-14) determines the activity leaving via this pathway.
. If the COLA sees that the normal fuel handling building ventilation system is < 3000 cfm,it assumes that the shutdown of the ventilation system was caused by a fuel handling accident (interlock with RM-G-9 or RM-A-14). It then assumes the ESF Ventilation is operating at a constant flow rate of 7000 cfm.
]
. Once the COLA determines the isotopic concentrations leaving via the ESF Ventilation, it uses the 7000 cfm flow rate to determine the uCl/sec leaving the plant (source term).
. Having developed the source term, thJ COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always a ground level release.
User inputs Needed to Perform the Calculation
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad
/%
N 206
Number ! P (
@u NUCLEAN Tui Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No.
l TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 133 of 164 ESF Fuel Handling Building COLA Calculation Using RMS Problems to Watch For
. The assumed isotopic mix is the same as a spent fuel assembly 72 hours after shutdown from full power. It is likely to be conservative for CDE if the actual fuel assembly is significantly older than that. Air samples from this pathway should be obtained as soon as possible to run a COLA or manual calculation using samples.
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, the;e may be a problem with the data. Verify met date with EACC and/or Control Room indications.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up .
with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get ! iodine samples even if E-520 dose rates are low.
- Depending on the phys l cal form of iodine in the field, the 'LLD* for thyroid dose rates measured by field m
teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if
, ) additional sensitivrty is desired. %.) '
e if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
- Consider getting an effluent sample from this pathway (radiciodine or pre-filter marinelli) as soon as possible.
e if the charcoal filters become degraded and are believed to be operating at less than normal efficiency, the manual code RMS calculation should be used for this pathway. The TSC can provide information on charcoal filter efficiency and degradation.
- Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses cant be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could b67 This type of calculation will produce conservative thyroid doses if the fuel assembly is significantly greater than 72 hours old.
,n v
207
7: Number l $$ 1 (Q NUCLEAR TML Radiological Controls Procedure 6610-PLN-4200.02 Title Revision Nc. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 134 of 164 ESF Fuel Handling Building COLA Calculation Using Samples How it works:
. Effluents leaving via the ESF ventilation system are sampled using a pre-filter marinelli sample
. The user enters the positive noble gas and iodine isotope concentrations from the analysis into the ESF l Ventilation Sample Edit Screen. i e if just the RM-A-14 iodine sample is used, the manual code must be used to perform the calculation, since the COLA will not produce a noble gas source term.
. If the COLA sees that the normal fuel handling building ventilation system is < 3000 cfm,it assumes that the shutdown of the ventilation system was caused by a fuel handling accident (interlock with RM-G-9 or RM-A-14). It then assumes the ESF Ventilation is operating at a constant flow rate of 7000 cfm.
Q. i Once the COLA determir s the isotopic concentrations leaving via the ESF Ventilation,it uses the 7000 cfm f flow rate to determine the t.Ci/sec leaving the plant (source term). G
= Having developed the source term, the COLA uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always a ground level release. .
I i User inputs Needed to Perform the Calculation
. Sample results need to be entered on the ESF Vent Sample Edit Screen
. Release duration using the Parameter Edit screen if other than 8 hours
. Met Data using the Parameter Edit screen if the data being picked up from the Met Tower is bad e) t, v
208
F-Number , /m ' ( NUCLEAR TMI
's Radiological Controls Procedure 6610-PLN-4200.02 Title Rewslon No.
TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 135 of 164 l l ESF Fuel Handling Building ' COLA Calculation Using Samples Problems to Watch For
. RM-A 14 iodine samples cannot be used solely with the COLA code since noble gases are not measured.
The COLA code w:ll not produce a noble gas source term from a RM-A-14 iodine sample. RM-A-14 iodine samples must be calculated using the manual codes to get the correct TEDE dose. The RM-A-14 iodine sample should still be input into the COLA code since the thyroid doses produced by the COLA will still be accurate.
. Monitor met data. If met data shows absolutely no changes for a 30 - 45 minute period, there may be a problem with the data. Verify met data with EACC and/or Control Room indications.
. Remember that the CDE contribuies to the TEDE. Do noi expect field team dose rate readings to match up '
with TEDE piojections. Roughly 3% of the CDE dose is incluc'ed in the TEDE dose. Ensure field teams get lodine samples even if E-520 dose rates are low. C's) . l Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field V teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If do:,e projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed psior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
0 b 209
Number > O G "" Radiological Co trols Procedure 6610 PLN-4200.02 Title Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 136 of 164 ESF Fuel Handling Building Manual Code Calculation Using RMS How it works:
. The isotopic distribution of activity leaving the plant is assumed to be the same isotopic distribution of activity as a fuel assembly 72 hours after shutdown from full power, after adjusting for lodine Insses (reduction factors) applicable to this pathway. The total activity in the fuel assembly is not important to this calculation, just the isotopic distribution.
. The iodine reduction factor for this pathway is a combination of reduction by decontamination in the spent fuel pool (0.01) and charcoal filtration (E). The totaliodine reduction factor is then (0.01)(1-E), normally =
0.0001.
. The RMS monitor (RM-A-14) determines the activity leaving via this pathway.
. The code defaults to a ventilation flow rate of 7000 cfm.
r
't . Once the code determines the isotopic concentrations leaving via the ESF Ventilation, it uses the flow rate to f()h determine the uCi/sec leaving the plant (source term).
. Having developed the source tem), the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always a ground level release.
UJer Inouts Needed to Perform the CalculatioD e The monitor to be und and the monitor reading.
. The time in minutes since the accident
. Charcoal filter operability and efficiency (normally 0.99)
. Met Data (wind speed, wind direction, and delta t)
. Release duration G
210 I l
r ( ,/ NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBli 1 Page 137 of 164 ESF Fuel Handling Building Manua! Code Calculation Using RMS Problems to Watch For
. The assumed isotopic mix is the same as a spent fuel assembly 72 hcurs after shutdown from full power. It is likely to be conservative for CDE if the actual fuel assembly is significantly older than that. Air samples from this pathway should be obtained as soon as possible to run a COLA or manual calculation using samples.
. The code prompts the user for the time since reactor shutdown to perform decay corrections. The time since the fuel handling accident occurred should be entered in this field.
- Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
p)e ( (_,/ Depending on the physical form of iodine in the field, the 'LLD* for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired. e if dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
- Periodically verify the release duratim and pathway with the ED.
. Consider getting an effluent sample from this pathway radiciodine or pre-filter marinelli as soon as possible.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be. This type of calculation will produce conservative thyroid doses if the fuel assembly is significantly greater than 72 hours old.
i
/% r 211 l
E l Number U (y) NUCLEAR TMI Radiological Controls Procedure 6610-Pl.N-4200.02 1 1 Tdie Revision No. ' TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 138 of 164 ESF Fuel Handling Building
< Manual Code Calculation ;
Using Samples i How it works: I
. Effluents leaving via the reactor building purge exhaust are sampled using an RM.A 14 iodine sample or a pre-filter marinelli sample.
. If a pre-filter marinelli sample is used, the user enters the positive noble gas and iodine isotope concentrations from the analysis when prompted by the code.
. If a radiolodine sample is used, the user enters the positive lodine isotope concentrations from the analysis when prompted by the code. Since radioiudine samples do not provide noble gas results, manual code will scale in the noble gases. It assumes that the noble gas to iodine ratio leaving the ESF ventilation is the same as the isotopic distribution of activity in the damaged fuel assembly after adjusting for iodine losses ,
(reduction factors) applicable to this pathway.
- p. The iodine reduction factor for this pathway is a combination of reduction by decontamination in the spent l
> 1 fuel pool (0.01) and charcoal filtration (E). The totaliodine reduction factor is then (0.01)(1 E), normally =
\d 0.0001 l
. The code assumes a ventilation flow rate of 7000 cfm.
. Once the isotopic concentrations leaving via the ESF Vent are entered, the code uses the ESF Vent flow ,
raf e te determine the uCi/sec leaving the plant (source term).
. Having developed the source term, the _ code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway is always a ground level release.
User inputs Needed to Perform the Calculation
. Sample results from ESF Ventilation System and the time of the sample
. Met Data (wind speed, wind direction, and delta t)
. Release duration l
l
\
V 212
r Number n (hbu k,./ WCEN Tu, Radiological Controls Procedure 6610-PLN-4200.02 Tale Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 139 of 164 ESF Fuel Handling Building Manual Code Calculation Using Samples Problems to Watch For
. Remember that the CDE contributes to the TEDE. Do not expect ficid team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the *LLD* for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. Periodically verify the release duration and pathway with the ED.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles (A
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? This type of calculation should be a pretty good approximation of what the doses could be.
~N I
)
/
f 213 j 1
Nurnber
/
(3 C/ i NUCLEAR TMI Radiological Controls Procedure 6610-PLN-4200.02 Tate Rension No i TMI Emergency Dose Calculation Manual (EDCM) 9 ) EXHlBIT 1 Page 140 of 164 ESF Fuel Handling Building Manual Code Calculation Contingency Calculation How it works: i
. This calculation is very useful for performing "What if' calculations, where t.1e precise response of an F',MS l monitor to a change in plant conditions cannot be predicted. It should be used when RMS data is not available. I
- This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is assumed to be equal to the gap activity in 56 fuel pins, 72 hours after reactor shutdown.
. The iodine reauction factor for this pathway is a combinatiun of reduction by decontamination in the spent fuel pool (0.01) and charcoal filtration (E). The total iodine reduction factor is than (0.01)(1-E), normally =
0.0001, l
. Once the code determines the isotopic concentrations in the fuel handling building, it uses ESF Vent flow rate input by the user to determine the uCi/sec leaking from the building (source term).
{}Y .
\
v Having developed the source term, the code uses th3 meteorological dispersion model to calculate the concentrations and doses at distances from the plant. i
. This pathway is always a ground level release. j User 1,[1 outs Needed to Perform the Calculation
. Charcoal filter operability and efficiency (normally 0.99)
. Flow rate out the ESF Ventilation System (normally assumed to be 7000 cfm)
. Met Data (wind speed, wind direction, and delta t)
. Re!aase duration s
V 214 s
Number i U TMl l / WUCLEA" Radiological Controls Procedure 6610-PLN-4200.02 t Tme Revrsion No l TMl Emergency Dose Calculation Manual (EDCM) 9 j EXHIBIT 1 Page 141 of 164 ESF Fuel Handling Building Manual Code Calculation Contingency Calculation Problems to Watch For I
. Of all calculations available, contingency calculations are the least accurate. They should only be used i when RMS data or effluent samples are not available. Contingency calculations are ideal for performing i
*What If" calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This type of calculation is very dependent on the activity in the damaged fuel assembly and the extent of damage to the fuel assembly. The calculation assumes 56 of 208 pins are broken during the accident, it assumes the assembly involved was shutdown from full power 72 hours ago. Conditions that vary from these assumptions will decrease the accuracy of the dose projection. ,
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up
. with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading j should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
. When performing "What If" dose orojections, clearly label them as "What If" calculations.
. Always communicate the uncertainty of the dose projection, is it a bounding calculation such that I know doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses actually could be? Depending on how closely actual conditiors match the stated assumptions, this type of calculation could go either way. However, it will typically be a bounding type calculation.
. Consider getting an effluent sample from this pathway (radiciodine or pre-filter /marinelli) as soon as possible.
215
NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tate Revision No TMI Emergency Dose Calculation Manual (EDCM) 9 F.XHIBIT 1 Page 142 of 164 s l ESF Fuel Handling Building Quick Code Calculation ; How it works: This calculation is not performed 0y the Quick Code. 216
Number C'; GPU TMi (j NUCLEAff Radiological Controls Procedure 6610-PLN-4200.02 True Rewsion No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 143 of 164 Reactor Building Fuel Handling Accident ! Manual Code Calculation l Using RMS How it works: l
. The isotopic distribution of activity leaving the plant is assumed to t> t. 'same as the isotopic distribution of '
activity fuel assembly 72 hours after shutdown from full power, afte -
- ting for iodine losses (reduction factors) applicable to this pathway. The total activity in the fuel assei. t is not important to this calculation, just the isotopic distribution.
. The iodine reduction factor for this pathway is a combination of reduction by decontamination in the fuel l transfer canal (0.01) and charcoal filtration (E). The totaliodine reduction factor is then (0.01)(1-E), normally
= 0.0001.
. If the purge valves are open, the RMS monitor (RM-A-9 Gas, RM-A-9 lodine, RM-A-9 High, or RM-G-24) determines the activity leaving via this pathway.
+ The user selects whether RM-A 9, RM-A 9 High, or RM-G-24 is used for the calculation. The logic (G)
N.J the coae uses in picking which monitor to perform the calculation is as follows: O If RM-A-9, RM-A-9 High, or RM-G-24 is selected, the code will ask the user if RM-A-9 lodine is on scale. If the answer is yes, the code will calculate the iodine source term based on the current and previous iodine channel readings specified by the user. If RM-A-9 iodine is offscale, the iodine source term is scaled in from the noble gas to iodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway. O If RM-A-9, RM-A-9 High, or RM-G-24 are selected, and RM-A-9 lodine is offscale, the code will ask the user if the charcoal filters are operational. If they are, the code will prompt the user to input their efficiency. The efficiency (E) should be input as 0.99 unless information from the TSC indicates another value should be used.
+ Once the code determines the isotopic concentrations leaving via the reactor building purge exhaust, it uses the flow rate to determine the uCi/sec leaving the plant (source terrn).
. If the purge valves are closed, the RMS monitor (RM-A-2) determines the airbome activity in the reactor building. The manyal code does not use RM-G-22 or RM-G-23. If RM-A-2 is isolated (>4 psiin RB), this calculation cannot be used. Use manual contingency calculation.
+ Once the code determines the isotopic concentrations in the reactor building, it uses the leakage flow rate from the buil ding, based on building pressure input by the user, to determine the uCi/sec leaving the plant (source term).
. Having developed the source teim, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the pnt.
( if the purge valves are closed, this pathway is always a ground level release. If the purge valves ara open, V) e this pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present. 217
I humber j U TMI
,/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Remon No TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 144 of 164 Reactor Building Fuel Handling Accident Manual Code Calculation Using RMS
.t)her inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (St.e STA)
Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
. The time in minutes since the accident
. Purge valves open or closed l l
. Charcoal filter operability and efficiency (normally 0.99)
. Met Data (wind speed, wind direction, and delta t) 1 0 I Release duration (d .
Problems to Watch For
. The assumed isotopic mix is the same as a spent fuel assembly that was shutdown from full power 72 hours ago 11 is likely to be conservative for CDE if the actual fuel assembly is significantly older than that. For this type of accident in the reactor building, an age of 72 hours is probably a pretty good assumption. Air samples from this pathway should be obtained as soon as possible to run a COLA or manual calculation using samples.
. The code prompts the user for the time since reactor shutdown to perform decay corrections. The time since the fuel handling accident occurred should be entered in this field.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is included in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed prior to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathwar m"h the ED.
- p. Consider getting an effluent sample from this pathwa PAP-5, pre-filter marinelli, or CATPASS as applicable) as soon as possible.
( 218
Number k U TMI NUCLEAN Radiological Controls Procedure 6610-PLN-4200.02 Td;e Revis#on No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 145 cf 164 Reactor Building Fuel Handling Accident ' Manual Code Calculation Using RMS
- . Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know doses can't be any higher than this? Or,is this the best approximation of what I believe offsite doses actually could be? This type of calculation will produce conservative thyroid doses if the fuel assembly is significantly greater than 72 hours old, V(%
1 1 J 219
- z. .
l Number U TMI NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 j Title Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 146 of 164 Reactor Building Fuel Handling Accident Manual Code Calculation Using Samples liqw it works: e If the purge valves are open, this calculation is identical to a reactor building purge exhaust manual code calculation using samples.
. if the purge valves are closed, this calculation is identical to a reactor building design basis leakage exhaust manual code calculation using samples.
l l 220
l l Number [ N NUCLEAR TMI (.,/ Radiological Controls Procedure 6610-PLN-4200.02 Title Revnion No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 147 of 164 Reactor Building Fuel Handling Accident Manual Code Calculation j ConUngency Calculation l How it works:
. This calculation is very useful for performing "What if' calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted. It should be used when RMS data is not 1 available.
. This calculation uses no RMS data. The isotopic distribution and total activity being released via this pathway is assumed to be equal to the gap activity in an entire fuel assembly (208 fuel pins),72 hours after reactor shutdown. The user may modify the number of pins that were believed to be damaged.
. The iodine reduction factor for this pathway is a combination of reduction by decontamination in the fuel transfer canal (0.01) and charcoal filtration (E). The total iodine reduction factor is then (0.01)(1-E), normally
= 0.0001,
, If the purge valves are open, the code uses reactor building purge exhaust flow rate input by the user to (Q) L/
. determine the uCi/sec leaking from the building (source term).
. If the purge valves are closed, the code uses reactor building leak rate based on the building pressure l input by the user to determine the uCi/sec leaking from the building (source term)
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. If the purge valves are closed, this pathway is always a ground level release. If the purge valves are open, this pathway can be a ground levelielease, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
User Inouts Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 37 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. Charcoal filter operability and efficiency (normally 0.99)
. Estimated number of fuel pins damaged
. Flow rate out the reactor building purge exhaust and the station vent if the purge valves are open
. Reactor b'uilding pressure if the purge valves are closed.
. Met Data (wind speed, wind direction, and delta t)
!' e. Release duration
'%)
221
l l Number
/ U TMI
( / NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Title Revmon No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 ' Page 148 of 164 Reactor Building Fuel Handling Accident Manual Code Calculation Contingency Calculation Problems to Watch For
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are not available Contingency calculations are ideal for performing "What if* calculations, where the precise response of an RMS monitor to a change in plant conditions cannot be predicted. ,
. This type of calculation is very dependent en the activity in the damaged fuel asiambly and the extent of damage to the fuel assembly. The calculation assumes as a default that 208 pns are broken during the j accident, it assumes the assembly involved was shutdown from full power 72 hours ago. Conditions that {
vary from these assumptions will decrease the accuracy of the dose projection.
]
l e Consider getting an effluent sample from this pathway (MAP 5 pre-filter /marinelli, or CATPASS as i applicable) as soon as possible. j
. Remember that the CDE contributes to the TtIDE. Do not expect field team dose rate readings to match up N with TEDE projections. Roughly 3% of the CCE dose is included in the TEDE dose. Ensure field teams get I iodine samples even if E 520 dose rates are low.
. Depending on the physical form of iodine in the field, the 'LLD" for thyroid dose rates measured by field j teams could be as high as 5 mrem /hr based on a 10 minute air sample Consider longer sampling times if l additional sensitivity is desired. 1
. If dose projections show PAG's are exceeded at 10 miles, va idation of the dose projection from field reading l should be performed prior to issuing a PAR beyond 10 miles. j i
. Periodically verify the release duration and pathway with the ED.
. When performing "What If" dose projections, clearly label them as "What If" calculations.
i
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know i doses can't be any higher than this? Or, is this the best approximation of what I believe offsite doses l actually could be? Depending on how closely actual conditions match the stated assumptions, this type of calculation could go either way. However, it will typically be a bounding type calculation.
1 i I 222
Number b {V T Ue "CN TMI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT 1 Page 149 of 164 l Waste Gas Decay Tank Rupture Accident ' Manual Code Calculation Using RMS How it works: l
. The isotopic distribution of activity in the tank is assumed to be the same as the isotopic distribution of I activity in TMI-1 FSAR Table 14.2 21 for the design basis waste gas decay tank rupture. The total activity in the tank is not important to this calculation, just the isotopic distribution. . The iodine reduction factor for this pathway is solely charcoal filtration (E). The totaliodine reduction factor is normally (1-E) = 0.01. . The RMS monitor (RM-A-4,6, or 8 Gas, RM-A-4,6, or 8 lodine, or RM-A-8 high) determines the activity leaving via this pathway. These monitors will respond to a waste gas decay tank rupture as well as increases in RCS activity and increased leakage into the auxiliary / fuel handling building. . The user selects whether RM-A-4, RM-A-6, RM-A 8 or RM-A-8 High is used for the calculation: logic the A COLA uses in picking which monitor to perform the calculation is as follows:
O If RM-A-8 or RM-A-8 High is selected, the code will ask the user if RM-A-8 lodine is on scale, if the answer is yes, the code will calculate the iodine source term base on the current and previous iodine channel readings specified by the user. If RM-A-8 iodine is offscale, the iodine source term is scaled in from the noble gas to iodine ratio after adjusting for iodine losses (reduction factors) applicable to this pathway. e !f RM-A-4 or RM-A-6 are selected. The code assumes the iodine channels are onscale and iodine channels must be input to get an iodine source term. 9 If RM-A-4 or RM-A-6 are selected, orif RM-A-8 lodine is offscale, the code will ask the user if the charcoal filters are operational. If they are, the code will prompt the user to input their efficiency. The efficiency (E) should be input as 0.99 unless information from the TSC indicates another value should be used. s
. Once the code determines the isotopic concentrations leaving via station vent, it uses the flow rate input by the user to determine the uCi/sec leaving the plant (source term). . Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant. . This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
l N b 223 1
Humber [ } U TMI (/ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Reesion No. TMI Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 150 of 164 Waste Gas Decay Tank Rupture Accident Manual Code Calculation Using RMS User inputs Needed to Perform the Calculation Plant data inputs are available on Area 38, Group 38 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. The monitor to be used and the monitor reading.
. The station vent and r retor building purge exhaust flow rates
. The time in minutes since the accident
. Charcoal filter operability and efficiency
. Met Data (wind speed, wind direction, and delta t)
O \ Release duration (v) = Problems to Watch For
. The assumed isotopic mix is the same as the design basis accident in the FSAR. It is likely to be very conservative for CDE since historically, there is much less iodine in the tank than was assumed in the FSAR accident. Air samples from this pathway should be obtained as soon as possible to run a COLA cr manual calculation using samples.
. The code prompts the user for the time since reactor shutdown to perform decay corrections. The time since the waste gas decay tank rupture occurred should be entered in this field.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections Roughly 3% of the CDE dose is included !a the TEDE dose. Ensure field teams get i
. iodine samples even if E-520 dose rates are low. l l
. Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if additional sensitivity is desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading should be performed pnor to issuing a PAR beyond 10 miles.
. Periodically verify the release duration and pathway with the ED.
l l . Consider getting an effluent sample from this pathway (MAP-5 or pre-filter marinelli) as soon as possible. t f\ !. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that I know V doses can't be any higher than this? Or,is this the best approximation of what i believe offsite doses l actually could be? This type of calculation will produce conservative thyroid doses. 224
1 3 @u NUCLEAN Tui Radiological Controls Procedure 6610-PLN-4200.02 Ytie Revision No TMI Emergency Dose Calculation Manual (EDCM) 9 EXHlBIT1 Page 151 of 164 Waste Gas Decay Tank Rupture Accident Manual Code Calculation Using Samples How it works;
. This calculatiori is identical to a station vent manual code calculation using samples.
{ t I 5 225
r Number
/'
OU "N TMl Radiological Controls Procedure 6610-PLN-4200.02 Tale Remion No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 152 of 164 Waste Gas Decay Tank Rupture Accident Manual Code Calculation Contingency Calculation How it works:
. This calculation is very useful for performing "What If" calculations, where the precise response of an RMS manitor to a change in plant conditions cannot be predicted. It should be used when RMS data is not aveilable.
. This calculation uses no RMS data. The user inputs the isotopic distribution and total activity being released via this pathway. Two selections are available; a " typical" tank inventory or a " worst case" inventory. Both of these inventories are very conservative based on operational history.
. The lodine reduction factor for this pathway is solely charcoal filtration (E). The total iodine reduction factor is normally (1-E) = 0.01.
. The code uses duration of the accident input by the user to determine the uCi/sec leaking from the building (source term).
\v/ . The code uses station vent flow rate input by the user to determine the effective release height.
. Having developed the source term, the code uses the meteorological dispersion model to calculate the concentrations and doses at distances from the plant.
. This pathway can be a ground level release, an elevated release, or a mixture of the two depending on the flow rate from station vent and the reactor building purge exhaust and the meteorological conditions present.
User inputs Needed to Perfom) the Calculation Plant data inputs are available on Area 38, Group 38 of the PPC (See STA) Met Data is available on Area 19, Group 2 of the PPC (See STA)
. Charcoal filter operability and efficiency (normally 0.99)
. " Typical" or " worst case" tank
. Flow rate out the reactor building purge exhaust and the station vent
. Met Data (wind speed, wind direction, and delta t)
. Release duration l
226
Number [ U TMI () NUCLEAA Radiological Controls Procedure 6610-PLh -4200.02
~
Tme Revision No. ~ TMi Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 153 of 164 Waste Gas Decay Tank Rupture Accident Manual Code Calculation Contingency Calculation Problems to Watch For
. Of all calculations available, contingency calculations are the least accurate. They should only be used when RMS data or effluent samples are noi available. Contingency calculations are ideal for performing "What If" calculations, whera the precise response of an RMS monitor to a change in plant conditions cannot be predicted.
. This type of calculation ir very dependent on the activity in the damaged waste gas decay tank. The assumptions for the ackity in the tank are very conservative.
. Remember that the CDE contributes to the TEDE. Do not expect field team dose rate readings to match up with TEDE projections. Roughly 3% of the CDE dose is inc;uded in the TEDE dose. Ensure field teams get iodine samples even if E-520 dose rates are low.
- p. i Depending on the physical form of iodine in the field, the "LLD" for thyroid dose rates measured by field v teams could be as high as 5 mrem /hr based on a 10 minute air sample. Consider longer sampling times if
\/ additional sensitivityis desired.
. If dose projections show PAG's are exceeded at 10 miles, validation of the dose projection from field reading l should be performed prior to issuing a PAR beyond 10 miles. j
. Periodically verify the release duration and pathway with the ED.
. When performing "What If" dose projections, clearly label them as "What If" calculations.
. Always communicate the uncertainty of the dose projection. Is it a bounding calculation such that i know thses can't be any higher than this? Or, is this the best approxirnation of what I believe offsite doses l tctually could be? Depending on how closely actual conditions match the stated assumptions, this type of l calculation could go either way. However, it will typically be a bounding type calculation.
l
. Consider getting an effluent sample from this pathway (MAP-5 or pre-filter /marinelli) as soon as possible.
[ (D
%/
)
227
Number TMl
) NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tme Revision No.
TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 154 of 164 PSLR vs COG RMS Reading Graphs Graphs are presented in this section which can be used to provide a rough estimate of PSLR based on condenser off g,s monitor readings and estimated RCS activities. The graphs are referenced in RAF 6612-97-011. The following assumptions were used in the graphs:
. The condenser off gas flow rate was fixed at 50 cfm to simplify the correction of the PSLR for the actual COG flow during use. The correction is made by taking the Factor value from the graph and multiplying it by the actual flow divided by 50 cfm.
. Total RCS activity for th" purposes of these graphs are the totaliodine and noble gas activity. Tritium is not to be included in the to- RCS activity.
. For RCS activities 10 uCi/cc and below, the isotopic mix is assumed to be the same as TMI Cycle 9, which represents a core with fuel pin defects that are present prior to the PSLR.
. Do Dol adjust total RCS activity for iodine spiking. The RCS total activities listed on the graphs are not intended to be used with iodine spiking fr.ctors since most of the iodine transferred to the secondary side D \
does not leave via the condenser off gas. For exampia, if the RCS activity prior to shutdown was 10 uCi/cc ( and an iodine spike of 50 would be expected upon shutdown (Damage Class 1 A), the iodine spiking would nql be used to adjust the RCS activity for the graph. The 10 uCi/cc line would still be used to determine the leak rate.
. For RCS activities above 10 uCi/cc up to Damage Class 2 the isotopic mix is assumed to be the same as a Damage Class 2. The total RCS activity specified for each diagonalline was multiplied by the normalized Damage Class 2 isotopic distribution to develop an isotopic distribution for that activity. This is meant to represent the RCS isotopic mix that could occur as gap activity enters the RCS during the accident.
. The lines for Damage Classes 2 - 5 are the default RCS activities found in the TMI EDCM.
. Graphs were generated for each range monitor for times of 0,4, and 8 hours post shutdown to account for isotopic distribution changes resulting from radiological decay.
. The graphs assume there are no direct to atmosphere releases (main steam relief valve, atmospheric dump valve, or EF-P-1) occurring. If such a release is occuning, the estimation of PSLR using condenser off gas monitor is more complex since all of the activity released into the secondary side does not reach the condenser. To correct for this loss of activity, the following equation can used:
PSLR, = (PSLRm)/(1 - F - F. - F.) Where: PSLR i = Total primary-to-secondary leakage (gpm)
= Primary-to-secondary leakage estimated from the graph (gpm)
PSLRcoa
;O. Fm = Estimated fraction of main steam flow estimated to be leaving through a main steam (j safety (dimensionless) 228 i
i
F Number U TMI ( NUCLEAN Radiological Controls Procedure 6610 PLN-4200.02 Title Revision No-TMi Emergency Dose Calculation Manual (EDCM) 9 l EXHIBIT 1 Page 155 of 164 Fe = Estimated fraction of main steam flow estimated to be leaving through EF-P-1 A j (dimensionless) Fe = Estimated fraction of main steam flow estimated to be leaving through the atmospheric dump valves (dimensionless) The user of these graphs must be aware that the PSLR results they produce are only as accurate as the estimate of RCS activities. The uncertainty in the PSLR derived should be essentially the same as the uncertainty of the RCS activity estimate. The graphs for 0,4, and 8 hours post shutdown are attached in the following pages. U O 229
Number
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TMl
/ Radiological Controls Procedure 6610-PLN-4200.02 i Tale Revson No TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 156 of 164 l rwm a ResAwie Countresse we Factor For PSLR Damage Casse 1 RCS teene Phee Neue one ActMuee (ucucc)
S to SS I le_ o_ t
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t ecteot - - - - - - - -- i l l t ooE*0S - - - - - I / u l ooteet . p
# / . .
j 1 O Hours Post Shutdown e oot *os ,
-[
f y
! , i
, , ee, e to 100 1000 0 00 Si Feeeer PSLR (gpm) e candenaar Denna Fknar f reni X Factor 50 The Pata seues W er one ywh em mer es menen u ow esm et acs newer N vanaway en the PSLR esmed showW es seseneasy se some es W esteeneney el the RCS erwey esemene l
I l 230
(8 TM1 NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Tdle Revision No TMl Emergency Dose Calculation Manual (EDCM) 9 , EXHIBIT 1 Page 157 of 164 F6gure 2 RetA 8#18 Countreens we Factor For PSUI Demago Class 1 RCS lodine to Plu.s NeWe s Gee ActMe,lese (ucikc) es o, J [ i m.ee 1 _. _ _ . . _ _ _ _ _ . l [ t i i i . i } . i m.a.
]
I, 7 i i , l q gg , h I p essi sw.e.n O
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PSLR (gpm) = Condenaar 08paa Famet Icffn1 X Facter so The Pt'Jt maae produced by Wue emeh am eWy se someets as One eetpues of RCS esseny. The uneenowny a sie PatJt eenses ensued en eenenese eis r some se se emsmetumey of sie RC8 enesey seemses 1 0 231
Number i U TMi G NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 M' Revrsion No TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 158 of 164 Mews 3 RNA-Siit Countreepe vs Factor For PSLR Demage Claes 1 RCS leens Pgp Ney Gas assvealee $Uce),, i I i i t 00E4 ---- - - - - - - -- -- - i i i g ooE+06 -- ------------4 4 I i l
' tcoe.04 -
O g ioca+os l s Hows Post Shutdown - 1 e oct+o2 0s 1 poseer to 100 1o00 0.01 PSLR (gam) = casunanaar Camas Fkeer fatal X Factor 50 The PSLR reemes amassed by sw yuan m eney as eesume se su eseue of RCs scsway The esionneewy at tw PSLR (eseres ehened be essenesgy 9tp eene te One encengewy of Wie ACS actuey eslumees 1 l l l 232
1 I Number O Titie NUCLEAR TMI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 159 of 164
~
Fleuse 4 muAme couneresse ve resser ree rota veryonemesietsineg,oe. Aeievgueucey ,, w, j j ; '10 i t On.a _ e iges e e . I I l I :
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l --- -i -- P
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t OBE*01 0 01 et 1 Poets, to 100 1000 PSLR (SPm) e F.andanner Danan Flamr fren) X Famer to the Psua muuas peousse ey em peon are on, as assueste se om ese une et acs essney Tw useneteney a sie PSLR deaved esieuW to e . Ihe esem as ein ennenamtp of om Ac4 anney esanuas I i O 233 I
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EI TMt Os NUCLEAR Radiological Controls Procedure 6610 PLN-4200.02 Trtle Revision No. TMI Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 160 of 164 Piews s R48A4844 Countrates we Facter For PSLR Verying RCS lodine plus gg Aegvities (ge) f l w _ _ i, t.00144 - - - ~ - - -- j 5 i l l l t m.os __ 4 _ . . _ . . I y .,_. - O
- t. _.
I I I ; I 100E.09 e et et t Pesear 10 100 4000 PSLR (gem) e candanner osrama FlearIct.nl X Factor 50 the Peut mesne pommes my me puen es awy a sammen as em esoneen er aca eenway Tae unestensy a me PeLa aanvas enume ne eenenauer se esme as me unenenwy et se secs asaw, esemate 234 1 l I
Number
\ NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02 Revision No.
TMl Emergency Dose Calculation Manual (EDCM) g EXHIBIT 1 Page 161 of 164 Peewee ROSA s HI C4Natrates vs Pector Fet PSLR Verying RCS ledene Pim g Gas Activites (gc) , too i cos.es e
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f _ 6' t est *03 -- 0 -
, , l2 , ,
l -~& t 00E*01 0 01
// et 1 Feener to PSLR (egen) = candanaar OAmma Flms isent X Factor 100 'M 50 fns Psin reeuse W ey su emen me any se answam es ow esanas er acs eene, % uneenemer e sw PSLR deseed eheuid he emmenesty me some se De W elsie ACS esihdly eserm6e.
O 235
O\ U TMi G NUCLEAR Radiological Controls Procedure 6610-PLN-4200.02
- Remion No.
TMI Emergency Dose Calculation Manual (EDCM) g EXHIBIT 1 Page 162 of 164 Pigewe 7 oca Rm0 28 Does Reese we Factor For PSLR verytne RCS ledene Mus gp lucroc)
; . . f
, ces a
q cog.06 .- - - - - - - - -- +- - sono 100 i , s t.00i!43 -- I i 1 pot.e3
- t l
! O Hours Post Shutdog l l
l 10o641 Peeter 10 100 1000 O ct 01 1 PSLR (gpm) e Candansar Ottons Fhnur frtn) X Facter 80 The PSLR soeums preemed by me graph em ordy as eenmee se tw seamens of RCS estvey ihe uncerteeig a Se PSLA densed esessed be eeneredy the esau as tie uneestaner ette ACS menusy estenets J O ; 236 l l i
Number O U TMI l WUcLEAN Radiological Controls Procedure 6610-PLN-4200.02 ) Tstle Reymon No. TMl Emergency Dose Calculation Manual (EDCM) 9 EXHIBIT 1 Page 163 of 164 Pagure & A000-24 Does Retes we Feeter For PSLR Verytng RCS leelne Plus Noble Get Activleles (uCyg
, ,,e .as ~a-- - -
i DC2 I 5000
, 0,
/ . ,00
; W t I # 100 h esta04 l //
5 100043 ----+----/ i
/ I O t e0E42 i
l 4 Hours Pest Shuldeten l i 1est 01 - 0 01 Ol i Feeeer 10 100 1000 PSLR (gem) e cansanmar Qanaa Flan fden) X Fesser 50 Tw Patn emme poense my me pesa me av, as eeneas no em seeines of acs omvey few smama, e to PSLA ennved ensued to semeneasy en esem es se . _ et se RCs eeder estones i I I 1 l I i I O 237
v 1 Numbsr O Title E8 NUCLEAR 1MI Radiological Controls Procedure 6610-PLN-4200.02 Revision No. TMl Emergency Dose Calculai on Manual (EDCM) 9 EXHIBIT 1 Page 164 of 164 Plgtere 9 manG4s Deee meses we Peeter Per PsLR very8ae acs ledene Poue menee see ocs Aeg gyee) ,Q I I i g 8000 I I f -
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l t90E*e4 _A_ . s b n oot.es yI , , O ' '
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8 Hours Poet Shutdown i i i PSLR (0pm) . e -- na rwe,$,*ai X Facter " 50 N PeLR reasse peeused e, sus puem mer ene es senseen es ou seemene et Res nuever N eneminner m Gio PeLR esaved emessa he essensesy se esse es one eneseemer of se RCs sesser seemste , l i i O 238}}