NRC-95-0032, Annual Radioactive Effluent Release Rept for Period of Jan-Dec 1994
| ML20087G208 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/31/1994 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-95-0032, CON-NRC-95-32 NUDOCS 9504040003 | |
| Download: ML20087G208 (270) | |
Text
{{#Wiki_filter:- i DrugIls R, Gipson Senior V;ce PresiJent Nuclear Genefahon i h Fermi 2 6400 North Dine Highway Newport, %chigan 48166
. (313)ES249 I
March 31,1995 7 NRC-95-0032 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D.C. 20555 ,
References:
- 1) Fermi 2 [
NRC Docket No. 50-341 NRC License No. NPF-43
- 2) Appendix A, Facility Operating License No.
NPF-43, Technical Specification 6.9.1.8 (
Subject:
Annual Radioactive Effluent Release Report )\ The Annual Radioactive Effluent Release Report for Fermi 2 is attached. This report is being transmitted in compliance with Reference 2 and Regulatory Guide 1.21, Revision 1. The attached report covers the period from January I through December 31,1994. t Please direct any questions or requests for additional information to Lynda Craine Supervisor, Radiological Health, at (313) 586-1388. Sincerely, a Enclosure cc: T. G. Colburn J. B. Manin , M. P. Phillips A. Vegel Region III l 9504040003 941231 PDR R ADOCK 05000341 [ / i PDR I ! L I
4 % i l l FERMI 2 NUCLEAR POWER PLANT DETROIT EDISON COMPANY OPERATING LICENSE NO. NPF - 43 , a N ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT , for the period of January 1,1994 through December 31,1994 I i r i 1 i
1994 Annu:1 R:dioactiva ! Effluent Release Report TABLE OF CONTENTS TABLE OF CONTENTS 2
SUMMARY
' 4
- 1. INTRODUCTION 7 ,
- 2. REGULATORY LIMITS 'S A. Gaseous Effluents 8-B. Liquid Effluents 9
- 3. AVERAGE ENERGY 9
- 4. MEASUREMENTS AND APPROXIMATIONS OF TOTAL ACTIVITY 10 A. Gaseous Effluents 10
- 1. Fission and Activation Gases 10
- 11. Radiciodines 10 lit. Particulates 11 IV. Tritium 11 V. Gross Alpha 12 B. Liquid Effluents 12
- 1. Fission and Activation Products 12
- 11. Tritium 13 ,
Ill. Dissolved and Entrained Gases 14 IV. Gross Alpha 14
- 5. ABNORMAL RELEASES 14
- 6. BATCH RELEASES 15
- 7. LIQUID EFFLUENT
SUMMARY
15 - A. Fission and Activation Products 15 B. Tritium 16 C. Dissolved and Entrained Gases 16 D. Gross Alpha Radioactivity 16 E. Waste Volume Released Pre-Dilution 16 F. Waste Volume Released Post-Dilution 17 G. Total Volume Dilution Water Released 17 H. Total For Each Nuclide Released (curies) 17
- 8. GASEOUS EFFLUENT
SUMMARY
17 A. Fission and Activation Gases 18 B. Radiciodines 18 C. Particulates 18 D. Tritium 19 E. Particulates: Totals for Each Nuclide Released (curies) 19 F. Fission and Activation Gases: Summary for Each Nuclide Released 19 G. lodines: Summary for Each Nuclide Released 19 2
3 1994 Annu 1 R:dicentiva Effluent Release Report
- 9. SOLID WASTE AND IRRADIATED FUEL SHIPMENTS 20 A. Solid Waste Shipped Offsite for Burial or Disposal (not irradiated fuel): 20 B. Irradiated Fuel Shipments 20
- 10. RADIOLOGICAL IMPACT ON THE PUBLIC 20 A. Dose Due to Liquid Effluents 20 B. Dose Due to Gaseous Effluents 21 C. Dose Due to Direct Radiation and Compliance with 40CFR190 21 D. Dose to Members of the Public on Site due to Effluents 22 E. Population Dose 23 F. Site Boundary Air Dose 23
- 11. RADIATION INSTRUMENTATION 23
- 12. METEOROLOGICAL DATA
SUMMARY
24
- 13. CHANGES TO DOSE CALCULATION AND ENVIRONMENTAL MONITORING LOCATIONS 24
- 14. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM) 24
- 15. MAJOR CHANGES TO RADIOACTIVE WASTE SYSTEMS 25
- 16. LIQUID HOLDUP TANKS EXCEEDING LIMITS 25 Appendix A: Meteorological Data Tables Appendix B: Changes to Offsite Dose Calculation Manual 1
1 3 i i
1 1994 Annu;l R: dis ctiva l Effluent Release Report
SUMMARY
The Fermi 2 Nuclear Power Plant maintains a comprehensive program of , monitoring and controlling the release of radioactive material from the site. The releases covered in this report are of three types: liquid releases, gaseous ' releases, and radioactive waste shipments. In a liquid release, a tank containing radioactive water is sampled for analysis prior to release. Based on the analysis, the amounts of radioactivity in the tank, , and the potential radiation dose to a member of the public are determined, and these figures are compared to federallimits. Conservative assumptions are used in calculating the radiation dose. For example, it is assumed that an individual eats 46 pounds of fish per year from Lake Erie caught directly offshore of the Fermi 2 plant discharge point. The tank will be released only after it is determined that federal limits are not exceeded. The contents of the tank are diluted by clean water (from.the circulating water decant line) as the tank is released in a ratio of approximateiy 400 gallons of clean water to one gallon of tank water for tanks located in the Radwaste Building basement (the normal release pathway). In the case of a Condensate Storage Tank release (an alternate release pathway approved in January 1994 by the plant Onsite Review Organization), this ratio is , approximately 50 gallons of clean water to one gallon of tank water. The release is continuously monitored by radiation detectors. As a result of the December 25,1993 turbine incident, the radwaste system and , the normal liquid release pathway were not available. In 1994, there were 3 liquid - releases from the Condensate Storage Tank, and no releases from the tanks in the Radwaste Building basement, in accordance with ALARA (As Low As Reasonably Achievable) principles, the three Condensate Storage Tank batches ; were thoroughly demineralized and filtered using portable equipment in order to ensure that the most restrictive Fermi 2 liquid effluent limit (monthly projected - dose to the total body) would not be exceeded. The Condensate Storage Tank release pathway and associated monitoring equipment were designed to be as similar as possible to the normal release configuration. A new radiation monitor was purchased and installed in the discharge path; the entire discharge path was , hard-piped; and the release was routed to the circulating water decant line and ! monitored by the circulating water decant line radiation monitor, as with a normal release.
1994 Annu:1 Rrdirctiva Effluent Release Report The three 1994 liquid releases contained 2.4 curies of tritium and 0.011 curies of other radioactive material. This amount of tritium is high compared with previous l years, simply because a larger volume of water was released. The tritium concentration in plant water is essentially constant, and it cannot be removed. 3 However, the dose impact of tritium is relatively low. Due to the thorough water cleanup, the amount of other radioactive material is fairly low compared with previous years, despite the large volume released. This is due to Detroit Edison's l I commitment to maintain all radioactive releases As Low As Reasonably Achievable (ALARA). The calculated 1994 liquid effluent radiation dose to the maximally exposed l individual due to 1-131,1-133, tritium, and particulates with half lives greater than 8 days, was 0.064 mrem to the total body and 0.085 mrem to the maximany l exposed organ. (These values are 2.2% and 0.85% of federallimits, l respectively.) Fermi 2 is continuing to work toward minimizing or eliminating liquid releases. l Between the last Condensate Storage Tank release in April 1994 and the date of this report (March 1995), there were no liquid radioactive releases from Fermi 2. Radioactive gaseous releases occur as part of the normal operation of Fermi 2. There are six ventilation system release points, or " stacks", each of which is monitored by a sophisticated and sensitive radiation monitor which continuously extracts a sample from the stack effluent. Since any gaseous radioactive material is diluted by the building ventilation air flow, the stack concentrations are I small. In fact, radioactive material is not detected in most stack samples. All I sample results are compared with federal limits to ensure they are not exceeded. If the amount of radioactivity in the effluent of any stack approaches a federal limit, an alarm will be activated in the Fermi 2 control room to alert operations personnel. After evaluating the situation, the operators may choose to' order increased sampling, shut down building ventilation, or divert the effluent stream to a special gaseous treatment system so that federal limits are not exceeded. In 1994, the amount of radiciodines and particulate radionuclides with half lives greater than 8 days in gaseous releases was 0.00027 curies. The amount of noble gases released in 1994 was 0.49 curies. These quantities are considerably less than those seen in previous years due to the fact that the plant did not operate above low power levels in 1994. Estimated annual radiation doses due to gaseous effluents from Fermi 2 are included in this report. Noble gas doses to air at the site boundary in 1994 were 0.00015 mrad gamma and 0.000075 mrad beta. These doses are 0.0015% and 0.00038%, respectively, of federallimits. Dose to the maximally exposed organ 5
s 1994 Annual R:dioactiva Effluent Release Report of the maximally exposed individual due to 1-131,1-133, tritium, and particulates - with half lives greater than 8 days, was 0.0028 mrem. This dose is 0.019% of the federal limit. Radioactive shipments of solid waste from the Fermi 2 site consist of waste generated during water treatment, radioactive trash, and irradiated components. Federal regulations govcining these shipments are extensive, and Fermi 2 complies with these regulations and with internal procedures. Shipment destinations are either licensed burial sites or intermediate processing facilities. In 1994, Fermi 2 did not ship any radioactive waste for final disposal due to the exclusion of Michigan licensees from the burial sites. 6
4
'1994 Annu21 R: disc: tivi I Emuent Release Report
- 1. - INTRODUCTION i
The Detroit Edison Fermi 2 Nuclear Power Plant is designed and operated in - ! a manner which strictly controls and monitors the release of radioactive. ; material to the environment in accordance with Nuclear. Regulatory ! Commission (NRC) and Detroit Edison Company requirements. This Annual- i Radioactive Effluent Release Report, for the January through December 1994 ' period, is submitted in accordance with Fermi 2 Technical Specification 6.9.1.8 and NRC Regulatory Guide 1.21. This report provides the following i information required by those references: A. Summation of the quantities of radioactive material (in the form of gases. and liquids) released from the plant (Sections 7 and 8). .! I B. Summation of quantities of radioactive material contained in solid waste - ; packaged and shipped for off-site disposal (Section 9). . 1 C. Changes to the Offsite Dose Calculation Manual (ODCM) (Section 14). D. A list and description of any unplanned releases of radioactive ma'terials to unrestricted areas (Section 5). E. A list of any new locations for dose calculation or environmental monitoring ; identified by the land use census (Section 13). l ! F. A list of effluent monitors which were inoperable for a period longer than I that specified in ODCM Controls 3.3.7.11 and 3.3.7.12, and an explanation i of why the time limit was exceeded (Section 11). i 1 G. A description of events leading up to any liquid holdup _ tanks exceeding the l limit of Technical Specification 3.11.1.4 (Section 16). l H. A description of any major changes to radioactive waste treatment systems (Section 15).
- l. - An assessment of the radiological impact on the 'public in terms of dose i due to liquid and gaseous effluents, both to the maximally exposed t individual and to the population within a 50 mile radius of the plant (Section- -
10). ., i J. A summary of 1994 meteorological data (wind speed and wind direction for l' different stability classes) which was used in calculating gaseous '
' dispersion factors (Section 12).
l ; 7
+ e ., , , - - . - . . , , - - - , , , , n .- n.., . , . . .
r 1994 Annu:1 R:dir ctiva-Effluent Release Report
- 2. REGULATORY LIMITS The Nuclear Regulatory Commission limits on liquid and gaseous effluents are incorporated into the Fermi 2 Offsite Dose Calculation Manual. These limits prescribe the maximum doses and dose rates due to radioactive effluents resulting from normal operation of Fermi 2. The limits are defined in several ways to limit the overall impact on persons living near the plant. The limits are described in the following sections.
A. Gaseous Effluents I. Dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to the following: a) Noble gases Less than or equal to 500 mrem / year to the total body Less than or equal to 3000 mrem / year to the skin b) lodine-131, lodine-133, tritium, and for all radionuclides in particulate form 'with half lives greater than 8 days Less than or equal to 1500 mrem / year to any organ. II. Air dose due to noble gases to areas at and beyond the site boundary shall be limited to the following: a) Less than or equal to 5 mrad for gamma radiation Less than or equal to 10 mrad for beta radiation
- During any calendar quarter b) Less than or equal to 10 mrad for gamma radiation Less than or equal to 20 mrad for beta radiation
- During any calendar year Ill. Dose to a member of the public from lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:
8
[k >, v 1994 Annu 1 Rrdit;citiva Effluent Release Report a
~
a): Less than or equal to 7.5 mrem to any organ
- During any calendar quarter g b)' Less than or equal to 15 mrem to any organ
- During any calendar year. !
B. Liquid Effluents !
- 1. The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations -
specified in Title 10 of the Code of Federal Regulations (10 CFR) Part-- ! 20 (Standards for Protection Against Radiation), Appendix B, Table 2' Column 2 for radionuclides other than dissolved or entrained noble
. gases, as required by the Fermi 2 Offsite Dose Calculation Manual. !
For dissolved or entrained noble gases, the concentration shall be ; limited to 2E-4 (.0002) microcuries/mi total activity. This limit is based ! on the Xe-135 air submersion dose limit converted to an equivalent . . concentration in water as discussed in the International Commission on ' i Radiological Protection (ICRP) Publication 2. ,
- 11. The dose or dose commitment to a member of the public from :
radioactive materials in liquid effluents released to unrestricted areas , shall be limited to the following: a) Less than or equal to 1.5 mrem to the total body , Less than or equal to 5 mrem to any organ :
- During any calendar quarter b) Less than or equal to 3 mrem to the total body :
Less than or equal to 10 mrem to any organ 1
- During any calendar year ;
- 3. AVERAGE ENERGY The calculated site boundary dose rates for Fermi 2 are based on -
identification of individual isotopes and on use of dose factors specific to each identified isotope or a highly conservative dose factor. Average' energy values i are not used in these calculations, and therefore need not be reported. i 9
1994 Annu:1 Rcdiscctiva Effluent Release Report ,
- 4. MEASUREMENTS AND APPROXIMATIONS OF TOTAL ACTIVITY As required by NRC Regulatory Guide 1.21, this section describes the methods used to measure the total radioactivity in effluent releases and to estimate the overall errors associated with these measurements. The effluent monitoring systems are described in Chapter 11.4 of the Fermi 2 Updated Final Safety Analysis Report (UFSAR). -
A. Gaseous Effluents
- l. Fission and Activation Gases i
Samples are obtained from each of the seven plant radiation monitors which continuously monitor the six ventilation exhaust points and from the Offgas Vent Pipe which carries the gland seal condenser exhaust, mechanical vacuum pump exhaust, and treated offgas streams. The Offgas Vent Pipe effluent is released through one of the six ventilation exhaust points (the reactor building exhaust plenum). The fission and - activation gases are quantified by gamma spectroscopy analysis of periodic samples. The values reported in Section 8 are the sums of all fission and activation gases quantified at all monitored release points. Considering the inherent variability in radiation measurement, the variability in effluent stream composition, and the uncertainties in effluent flow rate and instrument calibration, Detroit Edison estimates that the one sigma uncertainty of the fission and activation gas total release figures is plus or minus 30 percent. II. Radiolodines Samples are obtained from each of the seven plant radiation monitors which continuously monitor the six ventilation exhaust points. The radiciodines are entrained on charcoal and then quantified by gamma spectroscopy analysis. For each sample the duration of sampling and continuous flow rate through the charcoal are used in determining the concentration of radiciodines. From the flow rate of the ventilation system a rate of release can be determined. 10
.~. . - -
i f 3; *
- E 1994'Annu21 R disastiva
' Effluent Release Report !
4 . . -
. t
. The values reported in Section 8 are the sums of a!! radioiodines ';
quantified at all continuously monitored release points. j Considering the inherent variability in radiation measurements, the l variability in effluent stream composition, and the uncertainty in sample l
- and effluent flow rates, Detroit Edison estimates that.the one sigma : !
uncertainty of the total radioiodine release figures is plus or minus 17' ' percent.
-l Ill. Particulates j
. Samples are obtained from each of the seven plant effluent radiation monitors which continuously monitor the six ventilation exhaust points.-
The particulates are collected on a filter and then quantified by gamma spectroscopy' analysis. For each sample, the duration of sampling and continuous flow rate through the filter are used in determining the .l concentration of particulates. From the flow rate of the ventilation ; system a rate of release can_ be determined. , Quarterly, the filters from each ventilation release point are composited and then radiochemically separated and analyzed for Strontium (Sr)- l 89/90 using various analytical methods. If found, these radionuclides ! are reported as total particulate activity. ; The values reported in Section 8 are the sums of all particulates quantified at all monitored release points, q Considering the inherent variability in radiation measurements, the l variability in effluent stream composition, and the uncertainties in instrument calibration and in sample and effluent flow rates, Detroit Edison estimates that the one sigma uncertainty of the total particulate - release figures is plus or minus 16 percent. IV. Tritium Samples are obtained for each of the seven plant effluent radiation monitors which continuously monitor the six ventilation exhaust points. I The sample is passed through a bottle containing water and the tritium is " washed" out to the collecting water. Portions of the collecting water are analyzed for tritium using liquid scintillation counting techniques. For each sample, the duration of sample and sample flow rate is used l. l 11 i
I l 1994 Annu I Radioactiva Effluent Release Report to determine the concentration. From the flow rate of the ventilation ! system a release rate can be determined. . l The values reported in Section 8 are the sums of all tritium quantified at all monitored release points. Considering the inherent variability in radiation measurement, the variability in effluent stream composition, and the uncertainties in instrument calibration, sample and effluent flow rates, and collection efficiency, Detroit Edison estimates that the one sigma uncertainty of the total gaseous tritium release figures is plus or minus 30 percent. V. Gross Alpha The gaseous particulate filters from the seven plant effluent radiation - monitors are stored for one week to allow for decay of naturally occurring alpha emitters. These filters are then analyzed for gross alpha radioactivity by gas proportional counting, and any such radioactivity found is assumed to be plant related. The quantity of alpha emitters released can then be determined from sample flow rate, sample duration, and stack flow rate. The values reported in Section 8 are the sums of all alpha emitters , quantified at all monitored release points. Considering the inherent variability in radiation measurements, the variability in effluent stream composition, and the uncertainties in instrument calibration and in sample and effluent flow rates, Detroit Edison estimates that the one sigma uncertainty of the total gaseous gross alpha release figures is plus'or minus 16 percent. B. Liquid Effluents The liquid radwaste processing system and the liquid effluent monitoring system are described in the Fermi 2 UFSAR.
- 1. Fission and Activation Products Before the contents of each holding tank are discharged to the environment, a representative sample of the tank's contents is taken and retained. The sample allows for the determination of radioactive 12 i
1994 Annu:1 R:di:=tiva Effluent Release Report-material concentrations and establishes the rate at which the ' radioactive material can be discharged to the environment. At the end of the calendar quarter a composite sample is made of all discharge samples taken during the quarter. This composite sample consists of portions of each discharge sample which are proportional to the volumes discharged. The composite sample is analyzed for Iron (Fe)-55 and Strontium (Sr)-89/90. Radiochemical separations and various analytical methods are used to quantify the amounts of Sr-89/90 and Fe-55. The values reported in Section 7 are the sums of all fission and activation products found in all batch releases. Also reported in Section 7 are the pre-dilution waste volume (the total volume of liquid waste tanks released), the post-dilution waste volume (the total tank volume released plus the volume of circulating water released while the tanks were being released), and the total dilution volume discharged (the total volume of circulating water released during the reporting period). Considering the inherent variability in radiation measurement and the uncertainties in volume measurements and instrument calibration, Detroit Edison estimates that the one sigma uncertainty in total liquid fission and activation product release figures is plus or minus 15 percent. t II. Tritium , Before the contents of each holding tank are discharged to the environment, a representative sample of the tank contents is taken and retained. At the end of the calendar month a composite sample is made of all discharge samples taken during the month. This composite sample consists of portions of each discharge sample which are proportional to the volumes discharged. The composite sample is analyzed for tritium by liquid scintillation counting. l The values reported in Section 7 are the sums of all tritium quantified from all batch releases. Considering the inherent variability in radiation measurement and the certainties in volume measurement and instrument calibration, Detroit Edison estimates that the one sigma uncertainty in total tritium release figures is plus or minus 7 percent. 13
7
- 1 1994 Annu:1 Rrdioactiva '
Effluent Release Report Ill. Dissolved and Entrained Gases . Prior to releasing liquid radioactive waste to the environment, a sample ! is taken from the radwaste holding tank. This sample is representative - of the tank's contents. The sample is examined using gamma . spectroscopy to determine the concentration of dissolved and entrained
' noble gases.
The values reported in Section 7 are the sums of all radioactive gases found for all batch releases. Considering the inherent variability in radiation measurement and the uncertainties in instrument calibration and volume measureme 1ts,- ! Detroit Edison estimates that the one sigma uncertainty.in total dissolved and entrained gas release figures is plus or minus 33 percent. IV. Gross Alpha ' i Before the contents of each holding tank are discharged to the l environment, a representative sample of the tank's contents is taken- ? and retained. At the end of the calendar month a composite sample is made of all discharge samples taken during the month. This composite a sample consists of portions of each discharge sample which are l proportional to the volumes discharged. The composite sample is l analyzed for gross alpha radioactivity by gas proportional counting. .i The values reported in Section 7 are the sums of the gross alpha i I radioactivity from all batch releases. i Considering the inherent variability in radiation measurement and the uncertainty in volume measurements and instrument calibration, Detroit - Edison estimates that the one sigma uncertainty in total liquid gross - : alpha release figures is plus or minus 54 percent. i
' 5. ABNORMAL RELEASES l For the purpose of this report, an abnormal release is any release of radioactive material not performed in accordance with the Fermi 2 license and implementing procedures, No abnormal releases occurred during the reporting period. However, as discussed in the preface to this report, three liquid releases were made from the Condensate Storage Tank. Releases 14 . ~ , - - . _ .
l 1994 Annu:.1 Rcdiscctivo Effluent Release Report
]
from this tank had not been made previously and are not expected to be made , in the future. To allow these releases, safety evaluations were performed, l temporary modifications were approved and implemented, and changes to the Offsite Dose Calculation Manual and plant procedures were made.
- 6. BATCH RELEASES As required by Regulatory Guide 1.21, a summary of data for batch releases must be provided in this report. The following batch liquid releases from the Condensate Storage Tank to the circulating water decant line occurred between January 1,1994 and December 31,1994 (these releases occurred during February, March, and April 1994).
Number of releases: 3 Total time for all releases: 4609 minutes Maximum time for a release: 1668 minutes Average time for a release: 1536 minutes Minimum time for a release: 1423 minutes The only batch gaseous releases from Fermi 2 are the venting or purging of the primary containment (drywell or torus) atmosphere. These venting or purging releases pass through the reactor building ventilation or standby gas treatment system and are monitored by the final effluent monitors for these pathways. Separate data on these venting or purging releases are not reported because the associated data are already included in the gaseous effluent release data (Section 8). The amount of radioactive material released during venting and purging has been a small fraction of the amount released as continuous gaseous effluents.
- 7. LIQUID EFFLUENT
SUMMARY
A. Fission and Activation Products Total Release 6.58E-03 4.21 E-03 0.00E+00 0.00E+00 except Tritium, Gases, and Alpha (curies) Average Diluted 6.82E-10 1.67E-09 0.00E+00 0.00E+00 Concentration During Period ( Ci/ml)* 15 , l I l
1994 Annual R diosctivo Effluent Release Report Maximum Percent 1.64E-01 1.56E-01 0.00E+00 0.00E+00 of ODCM Control Limit for a Single Release (%) B. Tritium Total Release 1.70E+00 7.32E-01 0.00E+00 0.00E+00 l (curies) l Average Diluted 1.76E-07 2.90E-07 0.00E+00 0.00E+00 Concentration During Period ( Ci/ml)* l 1.06E-01 7.35E-02 0.00E+00 0.00E+00 l Control Percent of ODCM Limit (%)
- Dilution water volumes from part G were used in calculating these values.
C. Dissolved and Entrained Gases I ll Total Release { *<8.0E-08 ll *<8.0E-08 ll 0.00E+00 ll 0.00E+00 ll D. Gross Alpha Radioactivity Total Release 2.38E-05 *< 1.1 E-08 0.00E+00 0.00E+00 l (curies) E. Waste Volume Released Pre-Dilution Pre-Dilution 3.80E+06 l 2.11 E+06 l 0.00E+00 0.00E+00 l Volume (liters) l l l 16 _. . . - . y -
l 1994 Annuri R:dio:ctivo l Effluent Release Report F. Waste Volume Released Post-Dilution Post-Dilution 1.91 E+08 9.98E+07 0.00E+00 0.00E+00 Volume (liters) i G. Total Volume Dilution Water Released i i Total Volume 9.65E+09 2.52E+09 l 4.10E+09 l 3.19E+09 ; l (liters) l l l l 1 Less than the Lower Limit of Detection (LLD), i.e. the maximum sensitivity of measurement, in units of microcuries per milliliter (pCi/ml). H. Total For Each Nuclide Released (curies) (All are batch releases.) ll H-3 Il 1.70E+00 ll 7.32E-01 ll 0.00E+00 ll 0.00E+00 l Cr-51 l 5.26E-04 l *<3.8E-07 l 0.00E+ 00 l 0.00E+00 l Mn-54 l 4.03E-04 l 1.47E-03 l 0.00E+00 l 0.00E+00 l Co-58 l 2.49E-04 l 1.16E-04 l 0.00E+00 l 0.00E+00 l Co-60 l 2.52E-03 l 1.12E-03 l 0.00E+00 l 0.00E+00 l Sr-89 l 4.94E-04 l 5.06E-04 l 0.00E+00 l 0.00E+00 l Sb-125 l 7.68E-04 l *<1.7E-07 l 0.00E+00 l 0.00E+00 l l-131 l 1.33E-04 l *<4.8 E-08 l 0.00E+00 l 0.00E+00 l Cs-134 l 8.38E-04 l 5.54E-04 l 0.00E+00 l 0.00E+00 l Cs-137 l 6.51 E-04 l 4.48E-04 l 0.00E+00 l 0.00E+00 l Ce-141 l *<3.7E-08 l *<3.7E-08 l 0.00E+00 l 0.00E+00 l Ce-144 l *< 1. 5E-07 l *<1.5E-07 l 0.00E+00 l 0.00E+00 l Total l 1.71 E+00 l 7.36E-01 l 0.00E+00 l 0.00E+00 __
- 8. GASEOUS EFFLUENT
SUMMARY
(Mixed mode releases as defined in NRC Regulatory Guide 1.111) l 17
I 1994 Annu:1 Radiorctivo Effluent Release Report A. Fission and Activation Gases Total Release 0.00E+00 0.00E+00 0.00E+00 4.90E-01 l (curies) l Average Release 0.00E+00 0.00E+00 0.00E+00 6.16E-02 Rate for Period (pCi/sec) B. Radiolodines Total 1-131 1.26E-04 *<3.4 E-14 *<3.4 E-14 *<3.4 E-14 l curies l 1.62E-05 0.00E+00 0.00E+00 0.00E+00 l Average Release Rate for Period (pCi/sec) Less than the Lower Limit of Detection (LLD), i.e. the maximum sensitivity of measurement, in units of microcuries per milliliter ( Ci/ml). C. Particulates l l Particulates with 1.05 E-05 2.18E-05 8.88E-06 9.93E-05 half lives > 8 days (curies) Average 1.35E-06 2.77E-06 1.12E-06 1.25E-05 Release Rate for Period (pCi/sec) Gross Alpha 1.39E-06 1.61 E-06 2.74E-06 1.49E-06 i Radioactivity (curies) 18 I
1994 Annuni Rcdio::ctivo Effluent Release Report D. Tritium , Quader 1 Quader 2 Quader 3 Quader 4 Total Release *<5.0 E-08 *<5.05_-08 *<5.0E-08 *<5.0 E-08 E. Particulates: Totals for Each Nuclide Released (curies) i ll Cr- 51 0 3.95E-06 Q *<2.5E-13 ll *<2.5E-13 Q *<2.5E-13 l Mn-54 l *<4.8E-14 l 2.89E-06 l *<4.8E-14 l 1.13E-05 l Co-57 l *<1.4 E-14 l 1.20E-06 l *<1.4E-14 l *<1.4 E-14 l Co-60 l 5.47E-06 l 1.77E-05 l 8.49E-06 l 8.29E-05 - l Ba-139 l *<2.0 E-09 l *<2.0E-09 l *<2.0E-09 l 1.41 E-03 l Cs-138 l *<3.5E-11 l * <3.5 E-11 l *<3.5E-11 l 6.71 E-03 l Sr-89 l 1.05E-06 l *<2.0E-15 l *<2.0E-15 l 3.76E-06 ; l Sr-90 l *<3.0E-16 l *<3.0E-16 l 3.93E-07 l *<3.0E-16 1 l l Cs-134 l *<4.2E-14 l *<4.2E-14 l *<4.2E-14 l *<4.2E-14 l Cs-137 l *<5.0E-14 l *<5.0E-14 l *<5.0E-14 l *<5.0E-14 ) l Ce-141 l *<3.9E-14 l *<3.9E-14 l *<3.9E-14 l 1.38E-06 { l Ce-144 l *<1.3E-13 l *<1.3E-13 l *<1.3E-13 l *<.1.3E-13 l l Total l 1.05E-05 l 2.18E-05 l 8.88E-06 l 8.22E-03 Less than the Lower Limit of Detection (LLD), i.e. the maximum sensitivity of measurement, in units of microcuries per milliliter ( Ci/mi). F. Fission and Activation Gases: Summary for Each Nuclide Released i Xe-138 *<2.0E-07 *<2.0E-07 *<2.0E-07 4.90E-01 curies G. lodines: Summary for Each Nuclide Released 1-131 1.26E-04 *<3.4E-14 *<3.4 E-14 *<3.4E-14 l 4 curies l l Less than the Lower Limit of Detection (LLD), i.e. the maximum sensitivity of measurement, in units of microcuries per milliliter (pCi/ml). 19
1994 Annu:1 R:discctiva : Effluent Release Report
- 9. SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. Solid Waste Shipped Offsite for Burial or Disposal (not irradiated fuel):
No shipments in this reporting period., B. Irradiated Fuel Shipments No shipments in this reporting period. i
- 10. RADIOLOGICAL IMPACT ON THE PUBLIC
( A. Dose Due to Liquid Effluents l As discussed in Section 6.5.1 of the Fermi 2 ODCM, compliance with , ODCM Control 3.11.1.2, which limits dose to a member of the public to any i organ and to the total body due to liquid effluents, is evaluated by i calculating the dose to a hypothetical individual who both eats fish from Lake Erie and drinks water extracted from Lake Erie at the water intake for-the city of Monroe. Conservative assumptions from Regulatory Guide 1.109 are made about the quantity of fish and water consumed. The i individual organ and total body doses for 1994 to this hypothetical ! individual were calculated according to Section 6.5.1 of the ODCM and are i listed below. l i Organ 1994 Liquid Effluent Dose Bone . < 4.66E-2 mrem ^
' Liver ' J ;'c j 8.51E-2 mrem Thytoldf n j 6.88E-4 mrem
.K,ldney ; . i 2.81E-2 mrem Lurig , , ' '] 9.36E-3 mrem
,Gi-LLI i,y 3 4.30E-3 mrem Total body '+l 6.44E-2 mrem i
l 20 l
1994 Annu:1 R:dl=ctiva Effluent Release Report The highest organ dose,8.51E-2 mrem to the liver, is 0.85% of the ODCM Control 3.11.1.2 annual organ dose limit (10 mrem); the total body dose, 6.44E-2 mrem, is 2.15% of the ODCM Control 3.11.1.2 annual total body dose limit (3 mrem). B. Dose Due to Gaseous Effluents As discussed in Section 7.8.1 of the Fermi 2 ODCM, compliance with ODCM Control 3.11.2.3, which limits dose due to 1-131,1-133, H-3, and ' particulates with half lives greater than 8 days in gaseous effluents to any organ of a member of the public, is evaluated by calculating the dose to an individual in an age group which would receive the highest single organ dose of any member of the public. This individualis a child who is assumed to live at an offsite location which is known to have a garden based on the Land Use Census,. This child is assumed to eat food from this garden, and to also be exposed by the inhalation and ground plane pathways. The individual organ and total body doses to this individual due to 1-131,1-133, H-3, and particulates with half lives greater than 8 days were calculated according to Section 7.8.1 of the ODCM and are listed below. Organ 1994 Gaseous Effluent Dose to Receptor with Highest Single Organ Dose Bone' 1 1.14E-3 mrem - I.lver' " ' q .d 1.01E-3 mrem Thyroid f' j 2.81E-3 mrem
';j 1.01E-3 mrem
~
Kidney : Lung. ' . n. j 1.02E-3 mrem
,Gi-LLi- ~l 4;; 1.05E-3 mrem
- Totalbody ' : 1.05E-3 mrem The highest single organ dose to the maximally exposed receptor,2.81E-3 mrem to the thyroid, is 0.019% of the ODCM Control 3.11.2.3 annual dose limit (15 mrem).
C. Dose Due to Direct Radiation and Compliance with 40CFR190 Title 40, Part 190 of the Code of Federal Regulations requires that dose to an individual in the unrestricted area from the uranium fuel cycle be limited to 25 mrem /yr to the total body and 75 mrem /yr to the thyroid. The sources of fuel cycle dose not analyzed above are due to other fuel cycle 21
P 1994 Annu::1 R disastivo
~
g Effluent Release Report . facilities and dose due to direct radiation.' As discussed in Section 8.2 of the Fermi 2 ODCM, no other fuel cycle facilities contribute significantly to dose in the vicinity of Fermi 2. With respect to direct radiation, none of the _, offsite TLD locations listed in Table 10.0-1 of the ODCM showed 1994 TLD - 4 readings which were consistently greater than the TLD readings at the . : control locations. Since other facilities and direct radiation did not l
- contribute significantly to offsite dose, and since the preceding sections of .
this report show compliance with the more restrictive requirements of 10CFR50 Appendix 1, Fermi 2 was in compliance with'40CFR190 in 1994. O D. Dose to Members of the Public on' Site due to Effluents ,- Members of the public may receive dose on site as visitors or as _ 4 employees (non-radiation workers). As discussed in Section 8.0 of the ' s Fermi 2 ODCM, " visitors" to the Fermi 2 site may receive dose due to their . ,
- activities within the site boundary. For purposes of this analysis, visitors.
are members of the public who spend time within the site boundary and.- who do not do work associated with the operation of Fermi 2c The ODCM considers'two categories of visitors: persons ice fishing on Lake' Erie and ! persons spending time in the Fermi 2 Visitors Center. Employees (non- 1 radiation workers) may receive dose from various activities and at various - 3 locations (other than radiation areas) on site. ; Table 8.0-1 of the ODCM lists the maximum amount of time a member of ; the public is likely.to spend on site, the likely locations of exposure, and ; the effluent exposure pathways $which apply. A visitor is assumed to " spend 240 hours per year ice fishing near the site or 4 hours per year at the Visitors Center; an employee is assumed to spend 2500 hours per g year on site at various locations. Exposure by immersion in noble geses ; and by inhalation of radioactive particulates, lodines, and tritium are ; considered. In the case of employees, ingestion of potable water from the ; onsite water plant is also considered. The doses given below do not include dose due to the pathways already considered in parts A and C of ; this section, namely dose due to water ingestion from the Monroe water ; intake, fish ingestion, and direct radiation. ; i Based on these assumptions, the maximum dose in 1994 to a visitor at the Visitors Center is 1.02E-7 mrem to total body and 2.51E-7 mrem to the maximally exposed organ (thyroid). The maximum _ dose in 1994 to an ice fisherman is 9.37E-6 mrem to the total body and 2.78E-5 mrem to the maximally exposed organ (thyroid). The maximum dose in 1994 to an employee (non-radiation worker) on site is 1.49 E-3 mrem to the total body and 1.77 E-3 mrem to the maximally exposed organ (thyroid). o i 22 l l H
. . . - -- . I
y . , , I
. 1994 Annuil Rcdi:cctivo
. Effluent Release Report 1 j E. . Population Dose Dose to the population within a fifty mile radius of Fermi 2 due to 1994 gaseous and liquid effluents was calculated.
For liquid effluents, the fish ingestion and drinking water pathways were - considered. Since there is no significant commercial fishery in the Michigan waters of Lake Erie, the dose due to fish ingestion was assumed . : to be due to ingestion by the local population of the entire sport fish catch in these waters. In calculating dose due to fish ingestion, parameters from ! Regulatory Guide 1.109 were used, as was the UFSAR dilution factor of ! 100. The dose due to water ingestion was determined by assuming that all residents served by the Monroe water intake drink at the average rate given by Regulatory Guide 1.109, and by using the UFSAR dilution factor , to the intake of 77. The population total body dose due to drinking water was estimated to be 3 mrem, and the total body dose due to fish ingestion - was estimated to be 49 mrem, for a total estimated population total body . dose due to liquid effluents of 52 mrem. For gaseous effluents, the code MICROAIRDOS was used to estimate the , population dose. Inputs to the code were 1994 gaseous release data, wind direction and wind speed frequencies for each stability class, population in each of 10 segments of each of 16 sectors, stack release , specifications, etc. The estimated 1994 collective effective dose due to , gaseous effluents is 21 mrem. F. Site Boundary Air Dose Gamma and beta dose to air at the site boundary due to noble gases must be calculated to evaluate compliance with ODCM Control 3.11.2.2. In ' 1994, gamma rir dose was 1.46E-4 mrad and beta air dose was 7 53E-5 mrad. These doses represent 0.0015% and 0.00038% of the ODCM Control 3.11.2.2 gamma and beta annual air dose limits, respectively. : (The gamma dose limit is 10 mrad and the beta dose limit is 20 mrad.) _3 i t
- 11. RADIATION INSTRUMENTATION i
Fermi 2 ODCM Controls 3.3.7.11, Radioactive Liquid Effluent Monitoring Instrumentation, and 3.3.7.12, Radioactive Gaseous Effluent Monitoring - Instrumentation, require that those monitors which exceed the time specified for out of service status be reported in the next Annual Effluent Release
- Repod.
1 23
N i 1994 Annuni Rrdirectiva 3 - '~ Effluent Release Report- !
. 1 LOn December 25,1993,' the Liquid Radwaste Effluent monitor became *
' inoperable when it was submerged as a result of a turbine incident which
* . flooded the Radwaste Building basement. - This monitor remained submerged 1
until February 1994 due to the difficulty of removing water from the Radwaste_- i Building basement, thereby exceeding the 30 day out of service time specified i in ODCM section 3.3.7.11. It was subsequently restored to full operability. - No j liquid releases via the pathway associated with this monitor have taken place ; since the turbine incident. : r
- 12. METEOROLOGICAL DATA
SUMMARY
l The meteorological monitoring system is described in the Fermi 2 UFSAR. In ' accordance with Regulatory Guide 1.21, data recorded by that system is l provided here to permit the NRC to assess the radiological impact of Fermi 2 -i releases independently. The data format required by Regulatory G'uide 1.21- l is used. Appendix A contains the meteorological data tables. Specifically, these are joint frequency tables of wind speed versus wind direction for each ; atmospheric stability class for the 10 meter monitoring level. These data were ; use'd to derive annual average dispersion and deposition factors.
- 13. CHANGES TO DOSE CALCULATION AND ENVIRONMENTAL' l
. MONITORING LOCATIONS ;
During 1994, due to a request from local residents, one TLD and one air - sampler at the same location were moved to a nearby location 0.66 miles j west of Doty Road in the west sector at 270 degrees. These are control
- j locations. Also, a new milk sampling location was added. -It is in the north l sector at 6 degrees, and is 4.2 miles from the plant. .!
- 14. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM)
In January 1994, sections 0.0,3.0, and 6.0 of the ODCM were revised to allow I discharge of the Condensate Storage Tank to the Circulating Water Decant j Line and to institute proper controls on such releases, including use of a 4 Condensate Storage Tank discharge monitor in April 1994, after Condensate Storage Tank releases were completed, sectionc 0.0 and 3.0 of the ODCM were revised to delete operability and surveillance requirements for the ] !. Condensate Storage Tank discharge monitor. Appendix B contains a copy of i the entire ODCM as approved in April 1994 with revision bars showing the April 1994 revisions in sections 0.0 and 3.0, and the January 1994. revisions in section 6.0. Also included in Appendix B are sections 0.0 and 3.0 as - approved in January 1994 with revision bars showing the January 1994 changes. Plant documentation supporting these revisions is also included. L l 24 l t m
1994 Annu 1 R:di:cctiva Effluent Release Report
- 15. MAJOR CHANGES TO RADIOACTIVE WASTE SYSTEMS 1
D! ring 1994, there were no major changes to the liquid, gaseous, or solid radioactive waste treatment systems, e
- 16. LIQUID HOLDUP TANKS EXCEEDING LIMITS Fermi 2 Technical Specification 3.11.1.4 requires that the quantity of radioactive material contained in any outside temporary tank shall be limited to 10 curies, excluding tritium and dissolved or entrained noble gases. During ,
this reporting period, January through December of 1994, this activity limit for such tanks was not exceeded. End of Text i 25
k 1994 Annu 1 R discctiva Efflu;nt R:l::ca R p::rt Appendix A: Meterological Data Tables I l l l t
9 =o i i I 1 JOINT FREQUENCY DISTRIBUTION (JFD) AT THE 10-METER LEVEL 1994
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Nelliburton NUS Corporation Afr/Rediological Programs Department '1/12/1995 PAGE 1 TIME OF DAT: 16: 7:35 PROGRAM: JFD WERSION: PC-1.2 PRINTOUT OF INPUT CONTROL DATA i , TITLE: DECO FERM12 JFD at 10-Meters for 1994 SEGIN DATE: ' 94 1 1 1 END DATE: 94 12 31 24 OPTION TO PRINT MONTMLY JFDS: NO OPTION TO PRINT SEASONAL JFDS: NO OPTION TO PRINT STASILITY BY NOUR OF DAY: NO OPTION TO PLACE JFD IN FILE FORMATTED FOR PAVAN /N00D00: YES OPTION TO USE 12 WIND SPEED CLASSES INPUTTED WIND SPEED CLASSES IN MPN .75 2.50 4.50 6.50 8.50 11.50 14.50 18.50 23.50 30.50 39.50 .00 PRIMARY MASUREMENTS BASED 04: WIND SPEED MEASURED AT 10.0 METERS IN MPN BAD WIND SPEED DATA CODED: 999.90 WIND SPEED TMRESNOLD: .75 MPN WIND DIRECTION MEASURED AT' 10.0 METERS BAD WIND DIRECTION DATA CODED: 999.0 - STABILITY BASED ON 1=A,2=9,...,7=G BAD STAtlLITY CODED: 9.0 BACK-UP MEASUREMENTS BASED ON: NO SACKUP WIND SPEED MEASUREMENTS NO SACKUP WIND DIRECTION MEASUREMENTS NO BACKUP STABILITY MEASUREMENTS WIND SPEED HEIGNT TO BE USED FOR JFO: 10.00 METERS CONVER$104 FACTOR TO CONVERT SIGMA RANCE TO SIGMA TMETA: 6.0 FORMAT TO READ INPUT DATA: (412, F5.1, F3.0,727, F 1.0,155, A6,711, A3, A3, T27, A1) FIRST DATA RECORD READ: FERMI 2 94 1 1 1 9.0 193.0 4.0 m ma-. a , , , - - - , n - - - ~ m - --- -. _ - - - -__.
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- I I Nottiburton NUS Corporetten Air /Rediological Progreme Department 1/12/1995 PAGE 3 TIME 0# DAT: 16: 7:35 PROGRAM: JFD VERSION: PC-1.2 DECO FERMI 2 JF0 et 10-Meters for 1994 SITE IDENTIFIER: FERMI 2 DATA PERIOD EXAMINED: 1/ 1/94 - 12/31/94
*** ANNUAL ***
STAtlLITY CLASS- C STASILITY BASED ON 1=A,2=S,...,7=0 WIe MEASURED AT: 10.0 METERS WIND THRESNOLD AT: .75 MPN JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION IN NGURS AT 10.00 METERS SPEED i (MPN) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WWW NW NNW TOTAL CALM 0
.76- 2.50 1 1 0 0 1 0 0 2 0 J 0 0 0 2 0 1 8 2.51- 4.50 4 1 4 6 1 6 1 3 3 2 6 4 15 7 11 4 78
' 4.51- 6.50 2 7 16 8 8 17 16 3 15 to 5 14 22 27 15 21 210 6.51- 8.50 9 7 19 5 11 10 23 22 32 14 15 10 9 7 9 17 219
. 8.51-11.50 8 13 27 6 10 8 21 13 18 24 21 11 10 9 4 12 215 11.51 14.50 3 5 7 4 9 2 6 2 4- 8 15 6 2 5 3 4- 85 14.51-18.50 2 0 0 3 6 4 0 0 0 0 3 2 0 1 1 0 22 18.51 23.50 0 0 0 0 1 1 0 0 0 0 0 0 0 0 0 0 2
. 23.51-30.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 30.51-39.50 0 0 0 0- 0- 0 0 0 0 0 0 0 0 0 0 0 0
>39.50 0 0 0 0 0 0 0- 0 0 0 0 0 0 0 0 0 0 TOTAL 29 34 73 32 47 48 67 45 72 62 65 47 58 58 43 59 7 STA81LITY CLASS D STABILITY SASED 04 1=A,2=5,...,7=G WIND MEASURED AT: 10.0 METERS WIND T MESNOLD AT: .75 MPN JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION IN MOURS AT 10.00 METERS SPEED (MPN) N- NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NWW TOTAL CALM 1
.76- 2.50 4 8 3 5 1 3 3 6 5 2 11 13 9 6 8 6 93 2.51- 4.50 26 17 19 15 ' 18 17 19 11 19 17 36 38 38 36 39 33 398 4.51- 6.50 41 45 76 39 31 50 51 58 47 60 92 88 61 53 53 47 892 6.51- 8.50 35 56 94 38 32 . 39 58 39 84 70 108 56 22 36 34 48 849 8.51-11.50 48 55 79 48 46 54 55 30 29 99 142 53 38 35 32 33 8 76 11.51 14.50 27 50 26 16 25 12 12 8 to 33 64 28 9 23 6 9 358 14.51-18.50 9 -6 8 6 12 9 3 3 0 11 27 7 2 4 0 1 108 18.51-23.50 0 0 4 1 3 6 0 1 0 0 12 2 0 0 0 0 29 23.51 30.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0.
30.51-39.50 0- 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 m39.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 TOTAL 190 237 309 168 168 190 201 156 194 - 292 492 285 7 193 7 1 77 3604 __ - -_____ -_____ . . - . .-. - - - - _ . - - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ ~
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Nottiburton NUS Corporetton Air /Redlological Programs Department 1/12/1995 PAGE 4 TIME OF DAT: 16: 7:35 PROGRAM: JF0 VERSION: PC-1.2 DECO FERMl2 JFD at 10-Meters for 1994 SITE IDENTIFIER: FERM!2 DATA PERIOD EXAMINED: 1/ 1/94 - 12/31/94
*** ANNUAL ***
i STABILITY CLASS E STASILITY BASED ON 1=A,2=5,...,7=G WIND MEASURED AT: 10.0 METERS WINO THRESNOLD AT: .75 MPH JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION IN IKRMtS AT 10.00 METERS SPEED (MPN) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WWW NW .NNW TOTAL CALM 0
.76- 2.50 7 3 5 1 1 6 3 7 18 22 19 17 32 41 12 13 207 2.51- 4.50 33 12 14 5 to 8 8 27 34 41 79 74 68 63 52 44 572 4.51- 6.50 31 23 25 16 19 14 21 27 55 72 90 61 39 25 26 31 3 6.51- 8.50 9 to 12 15 14 20 24 31 50 92 53 29 9 18 4 u 404 8.51-11.50 5 1 4 1 6 18 16 21 40 95 48 7 3 6 2 275 11.51-14.50 1 3 0 1 2 3 12 9 10 . 38 23 0 1 1 0 S 104 14.51-18.50 0 0 0 0 1 0 2 5 2 9 4 0 0 0 0 0 23 18.51-23.50 0 0 0 0 0 0 0 1 1 2 0 0 0 0 0 0 4 23.51-30.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 30.51 39.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 m39.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 total 86 52 60 39 53 69 86 128 210 371 316 188 152 15I 96 104 2164 STA51LITY CLASS F STA81LITY BASED ON 1=A,2=3,...,7=G WIND MEASURED AT: 10.0 METERS WIND TNRESNOLD AT: .75 MPN JOINT FREQUENCY DISTRIBUTION OF WIND SPEED ANO DIRECTION IN NOURS AT 10.00 METERS SPEED (MPN) N NME NE ENE E ESE SE SSE S SSW SW WSW W WNW NW :eNW TOTAL CALM 3
.76- 2.50 5 6 3 1 2 0 1 6 12 14 9 24 32 27 22 17 181 2.51- 4.50 39 9 1 1 2 4 6 9 9 16 29 46 41 55 50 44 361 4.51- 6.50 17 5 0 2 1 7 5 8 10 20 2 1 5 5 1 5 94 6.51- 8.50 1 0 0 0 1 3 8 8 18 25 1 0 0 0 0 1 66 8.51-11.50 0 0 0 0 2 3 5 5 11 26 1 0 0 0 0 0 53 11.51-14.50 0 0 0 0 0 0 1 1 4 3 0 0 0 0 0 0 9 14.51-18.50 0 0 0 0 0 0 0 2 1 0 0 0 0 0 0 0 3 18.51-23.50 0 0 0 0 0 0 0 0 0 0- 0 0 0 0 0 0 0 23.51-30.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 30.51-39.50 0 0 0 0 0 0 0 0 0- 0 0 0 0 0 0 0 0
>39.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 TOTAL 62 20 4 4 8 17 26 39 65 104 42 71 78 87 7 67 770
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. y Nettlburton NUS Corporetten Air /Rediologicet Programs Deportsent 1/12/1995 PACE 5 TIME OF DAY: 16: 7:35 PROGRAM: JFD VERSION: PC-1.2 DECO FERMI 2 JFD at 10-Meters for 1994 SITE IDENTIFIER: FERMI 2 SATA PERIOD EXAMINED: 1/ 1/94 - 12/31/94
*** ANNUAL ***
STABILITY CLASS G STASILITY BASED ON 1=A,2=S,...,7=G WIND MEASURED AT: 10.0 METERS WIND THRESNOLD AT: .75 WM JOINT FREQUENCY DISTR 19UTION OF WIND SPEED AND DIRECTION IN NOURS AT 10.00 METERS SPEED (WM) N NME NE ENE E ESE SE SSE S SSW SW USW W WWW NW NNW TOTAL CALM 1
.76- 2.50 3 2 1 2 2 0 2 1 3 2 5 11 15 21 27 13 110 2.51- 4.50 16 1 0 0 1 1 4 0 6 1 8 10 17 28 27 29 149 4.51- 6.50 6 0 0 0 1 4 2 2 5 1 0 0 0 2 0 3 26 6.51- 8.50 0 0 0 0 0 0 0 5 5 0 0 0 0 0 0 0 10 i 8.51-11.50 0 0 0 0 0 0 2 8 2 1 0 0 0 0 0 0 13 11.51 14.50 0 0 0 0 0 0 1 3 0 0 0 0 0 0 -0 0 4 14.51-18.50 0 'O O O O O O O O O O O 0 0 0 0 0 18.51 23.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 23.51 30.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 30.51-39.50 'O O O O 0 0 0 0 0 0 0 0 0 0 0 0 0
>39.50 0 0 0 0 .0 0 -0 0 0 0 0 0 0 0 0 0 0 TOTAL 7 3 1 2 4 5 11 19 21 .5 13 21 32 51 54 45 313 STAstLITY CLASS ALL STABILITY BASED ON 1=A,2=8....,7=G WIND MEASURED AT: 10.0 METERS vim TNRESNOLD AT: .75 MPH JOINT FREQUENCY DISTRl9UTION OF WIND SPEED AND DIRECTION IN NOURS AT 10.00 METERS SPEED (MPN) N NNE NE ENE E ESE SE SSE S SSW SW WSW V WNW NW. WNW TOTF1 CALM 5
.76- 2.50 20 20 12 9 7 9 9 22 38 40 44 65 89 97 69 50 600 2.51- 4.50 119 44 44 34 35 42 39 50 74 78 158 178 185 201 .183 158 1622 4.51- 6.50 104 86 140 89 74 113 112 102- 139 173 190 183 159 153 '114 114 2065 6.51- 8.50 62 82 156 71 77 110 129 110 204 208 188 118 72 95 86 100 1868 8.51-11.50 76 84 134 62 96 113 111 83 119 254 225 106 74 81' 49 58 1725-11.51-14.50 34 63 40 24 43 20 34 23 32 85~ 109 37 29 49 23 22 667 14.51-18.50 14 6 8 9 21 15 5 10 '3 20 35 '9 7 13 1 4 180 18.51-23.50 0 0 4 1 4 7 0- 2 1 2 12 2 1 0 0 0 36 '
23.51-30.50 0 0 0 0- 0 0 0 0 0 0 0 0 0 0 0 0 0 30.51 39.50 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0
>39.50 0 0 0 0 0 0 0 0 -0 0 0 0 0 0. 0 0 0 TOTAL 429 385 538 299 7 429 439 402 610 860 961 698 616 689 7 506 N
= w -, . . . g , . , y 1- . . - .
<I
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.g Nottiburton NUS Corporation Air /Redfological Programe Department 1/12/1995 PAGE 6 TIME OF DAY: 16: 7:35 PROGRAM: JFD VERSION: PC-1.2 DECO FERMl2 JFD at 10-Meters for 1994 SITE IDENTIFIER: FERMI 2 DATA PERICD EXAMINED: 1/ 1/94 - 12/31/94
*** ANNUAL ***
STASILITY BASED ON 1=A,2=0....,7=G j WIND MEASURED AT: 10.0 METERS l. Vlie TNRESNOLD AT: .75 WH TOTAL NLmgER OF CBSERVATIONS: 8760 TOTAL NLNWER OF WALID OBSERVATIONS: 8748 TOTAL NL9WER OF MISSING OBSERVATIONS: 12 PERCENT DATA RECOVERY FOR TMIS PERIOD: 99.9 % MEAN WIND SPEED FOR TNIS PERIOD: 7.0 MPM TOTAL NUNGER OF CBSERVATIONS WITH BACKUP OATA: O PERCENTAGE OCCURRENCE OF STASILITY CLASSES j A B C D E F G j 6.92 5.18 9.59 41.20 24.74 8.80 3.58 =1 DISTRIBUTION OF WIND DIRECTION VS STASILITY N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM A 21 24 60 27 57 63 19 1 18 6 13 37 72 100 59 28 0 8 16 15 31 27 20 37 29 14 30 20 20 49 45 46 28 26 0 C 29 34 73 32 47 48 67 45 72 62 65 47 58 58 43 59 0 0 190 237 309 168 168 190 201 156 194 292 492 285 179 193 172 177 1 E 86 52 60 39 53 69 86 128 210 371 316 188 152 154 96 104 0 F 62 20 4 4 8 17 26 39 65 104 42 71 78 87 73 67 3 G 25 3 1 2 4 5 11 19 21 5 13 21 32 51 54 45 1 TOTAL 429 385 538 299 357 429 439 402 610. 860 961 698 616 689 525 506 5
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9
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nettiburton sus Corporetten Alr/nodletogicot Progrees Department 1/12/1995 PAGE 7-
. TIME OF DAY: 16: 7:35 P90eman: JFD WERSION: PC-1.2 4 LAST DATA HECORD READ: FERMf2 94 12 31 24 3.2 10.0 4.0 OUTPUT FILE NAME tjfd10m94.out i
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r 1994 Annu 1 Ridistctiva Effluent Release Report E i L Appendix B: Changes to Offsite Dose Calculation Manual . l r b 9 e f s f t
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EDP Tech Spec PDC Procedure , ABN SE (Attached) 44- ooo3 (4m ,u.(e T ' DER PE (Attached) N/c( Test Drawing No. / Effectivene .s Review (Attached) S Yes O No Other: (Drawings, Design Calculations, Correspondence, etc.) PART 2: OPERATING UCENSE CHANGES ;.4 NA~%W "MiEA' >Af ~ A) Document O Operating 1.ieense Q Tech Specs O Environmental Protection Plan - B) Section(s), Table (s), Figure (s), etc. Affected C) Reference and Source Documents Attached O NA O Other O Marked up pages O Significant Haurds Consideration O Envi.- u.;.1 Evaluation O EnvironmentalImpact/ Categorical Exclusion O Justification D) is change to UFSAR and/or approved plana / programs required? O Yes O No O IfRNo(s) El Priority O NA NRC approval required by (date): Explanation F) NRC Letter No. G) Amendment No. PART 3: APPROVALS : 4.- Y g
, Date /-2-W ,n SD 8) Technic pe ._
C) Respo eS Organisational Unit Head an I ew' Sheet Date / I~ El Other Date F) Director, Nuclear Licensing
$ Q Attach Management Review Sheet 86 M Date /- f 97 l G) OSRO (Fire Protection, RERP, PCP, ODCM, SEP, TS) j O NA O Attach Management Review Stuet O . . e/ 4/ A, Date //JS/"1(
H) NSRG (OL, TS, EPP) E NA O Attach Management Review Sheet Date FormFIP-RA241 Att1P1/106063 DTC: O TDLCR Pile O 1701.(B O K1fR O 1735 O TMPLAN O 1715.05 1
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i CONTINUATION SHEET in lC713l-l0lyjoi.IolOlmj nc I I I-1 I I l
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t EFFECTIVENESS REVIEW Refe m a d -l l)l.ldl l h l- lhlh l Revision C Page"[of 7 PART 1: UFSAR $NA A) Quality Assurance Program O Yes O No O NA Does the change (s) cease to satisfy the criteria of 10CFR50, Appendix B or reduce UFSAR program commitments previously accepted by the NRC7 Provide the basis for each change on Attachment 2, Page 2. B) Fire Protection Program O Yes O No O NA Does the change (s) adversely affect the ability to achieve and maintain safe ' shutdown in the event of a fire? Provide the basis for each change on Attachment 2, Page 2. PART 2: RADIOLOGICAL EMERGENCY RESPONSE PREPAREDNESS PLAN ENA A) O Yes O No Does the change (s) decrease the effectiveness of the RERP Plan? O Yes O No Does the RERP Plan, as changed, cease to meet the standards of 10CFR50.47(b) and 10CFR50 Appendix E7 + Provide the basis for each change on Attachment 2, Page 2. ' PART 3: SECURITY PLANS lE-NA A) Document B) O Yes O No Does the change (s) decrease the effectiveness of the Physical Security Plan ; or Security Personnel Training and Qualification Plan prepared pursuant to 10CFR50.34(c) or 10CFR73? O Yes O No Does the change (s) decrease the effectiveness of the first four categories of Informational Background, Generic Planning Base, Licensee Planning Base, and/or responsiblity matrix of the Safeguards Contingency Plan prepared pursuant to 10CFR50.34(d) or 10CFR73? Provide the basis for each change on Attachment 2, Page 2. PART 4: PROCESS CONTROL PROGRAM 6NA A) O Yes O No Does the change (s) reduce the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations? (Technical Specification 6.13) , Provide the basis for each change on Attachment 2, Page 2. PART S: ODCM Q NA l A) O Yes O'No Does the change (s) reduce the level of radioactive effluent control required by 10CFR20.106 or 10CFR20.1302,40CFR Part 190,10CFR50.36a, and Appendix ! to 10CFR Part 50? (Technical Specification 6.14) O Yes %o Does the change (s) adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations? (Technical Specifications 6.14) Provide the basis for each change on Attachment 2, Page 2. PART 6: APPROVALS l mor- s Date /~2-W f, / o Ott Da e - C) Quality Assu[nceISecurity Plans, QA Pgam) f' Date D) OSRO (Fire Protection Pro 6 ram, RERP Plan, Security Plans, PCP, ODCM) A A 1[ PA e,.
/
Date //U~/f $L Form FIP RA2-01 Att 2 P1/2081193 DTC: O TDLCR File: O 1701.03 O TCLCR O 1735 O TMPLAN O 1715.03
, i EFFECTIVENESS REVIEW DOCUMENTATION l Reference LCR l 3 l 3 l - l 6 l f l C l- l C l D l/t/l Revision d Pagefof 7 Document - '
Listed below is each change by section and page, the reason for the change, and the basis for including that the revised plan or program continues to satisfy the criteria for that plan or program. Attach all appropriate analyses or evaluations justifyin; the change (s). Section/Page Change Basis r
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Form FIP-RA2-01 Att 2 P2 /2081193 DTC: O TDLCR File: O 1701.03 O TCLCR O 1735 O TMPLAN O 1715.03
~ ,
t EFFECTIVENESS REVIEW DOCUMENTATION T Reference LCR: 93-040-ODM Revision: O Page 5 of 7
' Section/Page 0.0/0.0-1 Change; Update revision numbers for sections 0.0,3.0, and 6.0.
i Basis: The sections pertaining to liquid effluent instrumentation, liquid effluent controls, and liquid effluent calculations are being revised. - 0.0/0.0-2 Change: Change title of Section 6.3.1 in Table of Contents. Basis: Make section applicable to setpoint calculation for both Radwaste Effluent Monitor and CST Discharge Monitor. Table Change: Add CST discharge path radiatson monitor and flow 3.3.7.11-1/ rate monitor, with footnote (1) and Action 113. 3.0-5 and 6 Footnote 1 requires redundant sampling and verification prior to CST discharge as well as continual monitoring of the CST discharge , monitor during discharge. . Action 113 requires suspension of CST - discharge when the monitor is inoperable. Basis: CST discharge is designed to be as similar as possible to Waste: Sample Tank discharge, with the CST Discharge Monitor being in an . equivalent portion of the waste stream to that of the Radweste ; Effluent Monitor. Since the Radwaste Effluent Monitor is effectively
- inoperable during a CST discharge, similar actions are taken prior to :
a CST discharge and when the Radwaste Effluent Monitor is inoperable. Continual surveillance of the CST discharge monitor during discharge also assures prompt termination of discharge if any abnormal conditions, e.g. alarm, arise; And, since CST s discharge is suspended when the CST Discharge Monitor becomes ' inoperable, the controls on CST discharge are more conservative than those on Waste Sample Tank discharge. I
' ible Change: Add surveillances of CST Discharge Monitor (3.7.11-1/ similar to those for the Circulating Water 'I 3.0-7 and 8 Reservoir Decant Line Monitor.
Basis: Surveillance requirements are the same except that the CST Discharge Monitor is only tested for a local alarm, since it does not have remote alarm capability. l Table Change: Add CST discharge to allowable batch release ' 4.11.1.1.1-1 types.
/3.0-1g Basis: CST discharge is adequately controlled per this evaluation. 1 I
Table Change: Add sentence in notation b limiting batch release ! 4.11.1.1.1-1 to one tank at a time. !
/3.0-20 Basis: This is consistent with current practice.
I h EFFECTIVENESS REVIEW DOCUMENTATION- j Reference LCR: 93-117-ODM .i Revision: 0 .Page 6 of.7 ' Section/Page l
- 6.1.1.1.a/ Change: Add phrase "at the aischarge to Lake Erie".
6.0-2 Basis: Clarity that MPC limit applies at discherce to take, not at the i monitor. 6.1.1.2.b/ Change: Add section on CST Discharge Monitor. 6.0-3 Basis: Provide description and requirements similar to those of other liquid effluent monitors. 6.1.1.3.c/ Change: Add section on CST discharge flow rate 6.0-3 monitoring. Basis: Conforms to flow rate monitors on other liquid effluent lines. 6.3/ Change: Delete " trip" from description of setpoint , 6.0-5 function. Basis: Setpoint formulas now also apply to CST Discharge Monitor, which does not have a trip function. 6.3.1/ Change: Change title of section and add reference to CST 6.0-5 Discharge Monitor. Basis: Make section generic for both types of batch effluent monitors. 6.3.1/ Change; a) Revise first paragraph to clarity that the 50 6.0-7 gpm maximum rate applies only to Waste Sample Tank releases. b) Revise definition of SP and SEN i to allow units used for CST Discharge Monitor. Basis: a) Maximum discharge rate for CST discharge path may be higher than 50 gpm. b) CST Discharge Monitor will read out in cpm, versus cps for Radwaste Effluent Monitor, 6.3.1/ Change: Provide single conservative sensitivity factor 6.0-8 for CST Discharge Monitor. Basis: This value is based on a conservative analysis of expected release mixes and primary calibration data. 6.3.2/ Change: Revise Equation 6-7 (and definition of MPCF). 6.0-8 Basis: Revision makes Circulating Water Decant Monitor setpoint dependent on activity in decent line regardless of source of this activity, whereas before revision the setpoint was dependent on - Wasta Sample Tank activity.
l EFFECTIVENESS REVIEW DOCUMENTATION Reference LCR: 93-117-ODM Revision: O Page 7 of 7 Section/Page 6.3.5/ Change: Add reference to CST Discharge Monitor in 6.0-10 requirements for evaluation when alarm is received. Basis: CST Discharge Monitor alarms must be responded to as well as other liquid affluent monitor alarms. 6.3.5 NOTE / Change: Change note to clarify use of equation 6-8 when 6.0-10 ' Circulating Water Decant Line Monitor alarms. Basis: Clarifies that DF terms is not applicable in this case but that beta factors still apply. 6.6/ Change: Allow special dose projection evaluation in tho' 6.0-16 case of CST discharges. , Basis: it may be possible to project doses from CST releases by a more accurate means than simply by assuming that CST releases will
, continue at the same rate. 1 Figure Change: Add note stating that figure does not 6.0-1/ include CST discharge path.
6.0-20 Basis: Figure applies only to processing liquid radwaste to the waste sample tanks. e a
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LICENSING CHANGE REQUEST LCR l l l . ] Ol l ! . l O l h hl Revision C , Page/of L/ PART 1: IITSAR, PLAN, OR PROGRAM REVISION Q NA
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f 0 $M / Nt 3s0"l D) Refersfice and Source Documenet (identify) f/ ! EDP Tech Spec : PDC Pror edure ABN SE (Attached) DER PE ( Attached) b Loduhr Test _ Drawing No. Effectn eness Review ( Attached) Des O No Other: (Drawmgs, Design Calculations. Correspondence. etc.) PART 2: OPERATING LICENSE CHANGES & Al Document O Operstmg License O Tech Specs O Environmental Protection Plan . B) Section(s), Table (s), Figure (s), etc. Affected C) Reference and Source Documents Attached O NA O Other O Marked-up pages O Significant Hazards Consideration O Environmental Evaluation i O Environmental Impact / Categorical Exclusion O 'listificatiool- n r n ,m D) is change to UFSAR and/or approved plans / programs required? " "-- V L. I V C. LJ t O Yes O No O LC_R No(s) LICENSINA ~ r I l El Priority 0 NA I I NRC approval required by (date): APR 211994 Explanation F) NRC Letter No. G) Amendn eR\4%f gg _
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CCO Al Originato \/ , B) Technical fx' pert C) Responsible Su sor r rational Unit Head Date ZIfTlf D) Plant Manager /Q p e/ /g' O NA O Attac anagement Revie cet Date / E) Other Date rF) Director, Nuclear Licensing - j j qq ) O Attach Management Review t r= ShSEP, TS) Date 6/ M a G OSRO (Fire Protection, RERP, PCP, ODCM, O NA O Attach Management Review Sheet b _ J//hh Date Jd4A/ H) NSRG (OL TS, EPP) F ' N NA O Attach Management Review Sheet Date Form FIP-RA2-01 Att 1 P1/10@l93 DTC: O TDLCR File: O 170LC3 O TCLCR O 1735 O TMPLAN O 1715.03
. ... ___ j
{' EFFECTIVENESS REVIEW n*== = l 'Il Yi - I 013171- IOlDMl Revision o Page/DI Y PART 2:155AR h Al Quality Assurance Program O Yes O No O NA Does the change (s) cease to satisfy the criteria of 10CFR50, Appendix B or reduce UFSAR program commitments previously accepted by the NRC7 - Provide the basis for each change on Attachment 2, Page 2.
- 3) Fire Protection Program O Yes O No O NA Does the change (s) adversely affect the ability to achieve and maintain safe shutdown in the event of a fire?
Provide the basis for each change on Attachment 2 Page 2.
- PART2: RADIOLOGICAL EMERGENCY RESPONSE PREPAREDNESS PLAN &
A) O Yes O No Does the change (s) decrease the effectiveness of the RERP Nn? O Yes O No Does the RERP Plan, as changed, cease to meet the standards of 10CFR50.47(b) and 10CFR50 Appendix E? Prmide the basis for each change on Attachment 2. Page 2. PART 3: STCUR11Y PLANS & l A) Document B) O Yes O No Does the change (s) decrease the effectiveness of the Physical Security Nn or Security Personnel Training and Qualification Nn prepared pursuant to 10CFR50.34(c) or 10CFR73? O Yes O No Does the changets) decrease the effectiveness of the first four categories of Informational Background, Generic Planning Base, Licensee Planning Base, and/or responsiblity matrix of the Safeguards Contingency Nn prepared pursuant to 10CFR50.34(d) or 10CFR73? Pmvide the basis for each change on Attachment 2, Page 2. PART 4: PROCESS CONTROL PROGRAM & Al O Yes O No Does the change (s) reduce the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable - regulations? (Technical Specification 6.13) Provide the basis for each change on Attachment 2. Page 2. PART5: ODCM D NA A) O Yes % Does the change (s) reduce the level of radioactive effluent control sequired by 10CFR20.106 or 10CFR20.1302,40CFR Part 190,10CFR50.36a, and Appendas I to 10CFR Part 50? (Techmcal Specification 6.14) O Yes - Does the change (s) adversely impact the accuracy or reliability of effluent, dose, or setpoent calculations? (Technical Specifications 6.14) Provide the basis for each change on Attachment 2. Page 2. -
, , PA}T 6: APPROVA,LS ,
Am M V Date D B) Technicatffspert' ' ' '# ' C) Quality Assurancef(5Scunty Plans, QA Program) I Date D) OSRO (Fire Protection Program, RERP Nn, Security Plans, PCP, ODCM) CLJ8fL Date 4lL4 k 6 l Form FIP.RA2 01 Att 2 P1/2081193 DTC: O TDLCR ' File: O 1701.03 j O TCLCR O 1735 O TMPLAN O 1715.03 L.__._..._.__..._ . . . _ __ _ . _ _ _ _ _ _ _
EFFECTIVENESS REVIEW DOCUMENTATION ; Reference LCR lhl l-lOl l l- ! Cl h f l Revision o Page3of Y C&le. Thrt (d?coldi+ $auf , Listed below is each change by section and page, the reason for the change, and the basis for including that the revised plan or program continues to satisfy the criteria for that plan or program. Attach all appropriate analyses or evaluations justifyin g the change (s). ! Section/Page Change Basis o.o / a o -l V<det mein ria se4., ~ tw : 4,,uInt GcrecAv efHAm,LiL 1
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EFFECTIVENESS REVIEW DOCUMENTATION ' Reference LCR l l l-lOl[l l- l O l h l$ Revision o Page%fY
'"' 0 6,f Dar<.~ GJculo-/;~ /1'7a>,ue Listed below is eac'i change by section and page, the reason for the change, and the basis for including that the revised plan or program continues to satisfy the criteria for that plan or program. Attach all appropriate analyses or evaluations justifyin g the change (s).
Section/Page Change , Basis Tdle. (d17. ll-l/ On Asu. 3,0-7, e@ G[07~qhdlae. h/iT h/N4/AJ 7.o-?. 7 0-9 ' l "ff)% ithnf 2.b U T. e . n N e/n e % be a swww/
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l PRELIMINARY EVALUATION Page1of '"2_ _ PART2: DESCRIPTTON OF OfANGE 0%pswr) A) Document" - C) PISNaaber c>Dcpf B) Revisionifdppnad # YocAt /r ,&? efth,a'-ides &u$m ta W 1
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# PART2: PRELIMINARYEVALUATION(Preparer, Approper)
A) Reviewof Commitment O N/AMeetsM- sCritena6.1.4.5. Appmvedby: '
$No comnutments O Commamersexit(list) - _
O Commitmentsmet-nonenegated
. O Commamentsneeddumgang(explain) l O Safety Engineering and/or Licensing have been contacted to make changes i B) on Ucanee, Plans,or Progrens ,
change to the Operating License (including the Technical "i+M =tions, the Environmental tection Plan, and the Technical W&ation Bases), the Core Operating Limits Report (COLR),
- f. the UPSAR, Fire Protection Fig.w. Quality Assurance Ti+w., Radiological E eger.g Response ,
Preparedness Plan, Physical Security Plan, Safeguards Conur ;..cy Plan, Security Personnel Training and Qualincation Plan, Offsite Dose Calculation Manual, Process Control Fi+w., or the Inservice * ' hopection on Inservice Testmg Programs. EiucenseChangeRapnet(IfR)requimd: LCR $ d. E .M k N C) en Enwhenment effect on environment 3 Einvironment affected - contact Feral Environmental Engineer and indicate resolution ; i D) Need for Safety Evaluation (check appropriate answer) l
.Yes N :
O 1. Is this a change to the facility, including z i-Ga-, as described in the UPSAR? l Q 2. Is this a change to a procedure, including r-plc--, as described in the UPSAR? j Q 3. Does this change constitute a Special Test? l If any are "Yes," initiate a Safety Evaluation unless an NRC Safety Evaluation Report exists F.ad..g the repairements of step 6.1.11. If all are "No," provide the basis to support the detenninstion on the continuation sheet apesquired by;teps 6.1.7.2,6.1.83, and 6.1.9. , steh7/ P Dat 7- , Form FIP-SR101 Att 1 P1/1052193 File: 0923.02
--- - -- - l
PRELIMINARY EVALUATION CONTINUATION SHEET Page2.of 2 B) Revision of Approved A) Document Identification o DCA s O Comments l iH C.e- "- W T r & ES Y R;rihk - Ik (JFSk{ kPC kV iMtv 0 SG eV A. W oh 1 M ft 9s' AEOCidz 24rW f' W AW /iCA- , inA c4 4 d 6M 'e Asmke tu $ UP592 . 2 . 0nce $ OFff/? b d incAA pecsb n W rr %e m/ 0Y W V/C2 h
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File: 0923.02 Form FIP-SR1-01 Att 2 P1/1052193
Nuclear ProductiIn - F rmi 2 CDCM-0.0 9 Offalte Do^o C:lcul:ti:n Manutt R:visi:n 8 .) P g3 0.0-1 1 i i DETROIT EDISON - FERMI 2 l OFFSITE DOSE CALCULATION MANUAL implementation Plan These revisions go into effect upon approval. Current Revision Status by Section 1 Section Number Revision Number Approval Date 0.0 8 04/26/94 l V/i/ 1.0 5 09/21/93 2.0 5 09/21/93 3.0 4.0 8 6 04/26/94 09/21/93 l Y/i f
- 5.0 5 09/21/93 6.0 6 01/25/94 7.0 6 09/21/93 8.0 5 09/21/93 #
9.0 5 *
. 09/21/93 10.0 5 09/21/93 Appendix A 4 08/07/92 Appendix B 4 08/07/92 ARMS - INFORMATION SERVICES Date approved: Yh hN Release authorized by: ( -
l Change numbers incorporated: 94-058-ODM [ _ DSN Od C/l- d.d Rev 8 Date APR 271994 8 DTC TMPLAN ' File 1715.02 Recipient [ dM ,
L UUCM-U.u R3vislen 8 Page 0.0-2 TABLE OF CONTENTS
'(- Pese Section 1.0-1
1.0 INTRODUCTION
PART I - RADIOLOGICAL EFFLUENT CONTROLS 2.0-2 2.0 DEFINITIONS 3.0-1 3.0 CONTROLS AND SURVEILLANCE REQUIREMENTS 3.0-2 3/4.0 Controls and Surveillance Requirements Applicability 3.0-4 3/4.3.7.11 Radioactive Uquid Effluent Monitoring instrumentation 3.0-9 3/4.3.7.12 Radioactive Gaseous Effluent Monitoring instrumentation 3.0-18 3/4.11.1.1 Uquid Effluents Concentration 3.0 3/4.11.1.2 Uquid Effluent Dose 3.0-23 3/4.11.1.3 Uquid Waste Treatment , 3.0-24 3/4.11.2.1 Gaseous Effluents Dose Rate 3.0-28 3/4.11.2.2 Gaseous Effluents Dose - Noble Gases 3.0-29 3/4.11.2.3 Gaseous Effluents Dose - lodine-131, lodine-133, Tritium, and Radionuclides in Particulate Form 3.0-30 3/4.11.2.4 Offges Treatment System 3.0-31 3/4.11.2.5 Ventilation Exhaust Treatment System 3.0-32 3/4.11.2.8 Venting or Purging 3.0-33 3/4.11.4 Radioactive Effluents Total Dose 3.0-34 3/4.12.1 Radiological Environmental Monitoring Program 3.0-46 3/4.12.2 Land Use Census 3.0-47 3/4.12.3 Interlaboratory Comparison Program 4.0-1 4.0 BASES 5.0-1 5.0 ADMINISTRATIVE CONTROLS 5.0-2 5.9.1.7 Annual Radiological Environmental Operating Report l 5.0-2 5.9.1.8 Annual Effluent Release Report 5.0-4 5.15 Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems PART ll - CALCULATIONAL METHODS 6.0-2 6.0 LiOUlD EFFLUENTS 6.0-2 6.1 Radiation Monitoring Instrumentation and Controls - 6.0-2 6.1.1 Offsite Dose Calculation Manual (ODCM) 3.3.7.11 Requirement 6.0-3 6.1.2 Non-ODCM Required Monitor
~ 6.0-3 6.2 Sampling and Analysis of Uquid Effluents 6.0-4 6.2.1 BATCH Releases 6.0-4 6.2.2 CONTINUOUS Releases 6.0-5 %.3 Uquid Effluent Monitor Setpoints 6.0-5 6.3.1 Tank Effluent une Monitors
~ ODCM-0.0 ' !
Revislan 8 - l l Page 0.0 ' TABLE OF CONTENTS (continued) f Page Section 4.0-8 6.3.2 Circulating Water Reservoir Decant une Radiation Monitor (D11-N402) 6.0-9 6.3.3 Generic, Conservative Alarm Setpoint for D11-N402 L 4.0-9 6.3.4 Alarm Setpoint for GSW and RHR System Radiation l l Monitors 6.0-10 6.3.5 Alarm Response - Evaluating Actual Release Conditions , 6.0-10 6.3.6 Uguld Radweste Setpoint Determination With l Contaminated Circulating Water Pond i 6.0-11 6.4 Contaminated OSW or RHR System - Quantifying and Controlling Releases 6.0-12 6.5 Uquid Effluent Dose Calculation - 10 CFR SO 6.0-12 6.5.1 MEMBER OF THE PUBUC Dose - Uguld Effluents 6.0-14 6.5.2 Simplified Uquid Effluent Dose Calculation 6.0-15 6.5.3 . Contaminated GSW System - Dose Calculation 6.0-15 6.6 Uquid Effluent Dose Projections 7.0-1 7.0 GASEOUS EFFLUENTS 7.0-1 7.1 Radiation Monitoring instrumentation and Controls 7.0-1 7.1.1 Effluent Monitoring - Ventilation System Releases 7.0-1 7.1.2 . Main Condenser Offgas Monitoring 7.0-2 7.1.3 Reactor Building Ventilation Monitors (Gulf Atomic) 7.0-2 7.2 Sampling and Analysis of Gaseous Effluents 7.0-2 7.2.1 Containment PURGE ' 7.0-2 7.2.2 Ventilation System Releases 7.0-3 7.3 Gaseous Effluent Monitor Setpoint Determination 7.0-3 7.3.1 Ventilation System Monitors 7.0-5 7.3.2 Setpoint Determination with No Nuclides Detected 7.0-5 7.3.3 Gaseous Effluent Alarm Response - Evaluating Actual , Release Conditions 7.0-6 7.4 Containment Drywell VENTING and PURGING 7.0-6 7.4.1 Release Rate Evaluation 7.0-7 7.4.2 Alarm Setpoint Evaluation 7.0-7 . 7.5 Guantifying Releases - Noble Gases 7.0-7 7.5.1 Sampling Protocol 7.0-8 7.5.2 Release Concentration Determination for Reactor !
~ Building Exhaust Plenum l 7.0-9 7.5.3 Calculation of Activity Released 7.0-9 7.6 Site Boundary Dose Rate - Radiolodine and Particulates 7.0-10 7.7 Noble Gas Effluent Dose Calculations - 10 CFR 50 ,
7.0-10 7.7.1 UNRESTRICTED AREA Dose - Noble Gases l I I
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Revisi:n 8 Pag 3 0.0-4 l
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TABLE OF CONTENTS (continued)
'Page Section 7A-11 7.7.2 Simplified Dose Calculation for Noble Gases
'7.6-11 7.8 Radiolodine and Particulate Dose Calculations - 10 CFR 50 7.0-11 7.8.1 UNRESTRICTED AREA Dose - Radiolodine and Particulates 7.0-13 7.3.2 Simplified Dose Calculation for Radiolodinos and :
Particulates 7.0-13 7.9 Gaseous Effluent Dope MiecFen 7.0-14 7.10 Waste Oil IncinereEon 8.0-1 8.0 SPECIAL DOSE ANALYSES 8.0-1 8.1 Doses Due to Activities inside the SITE BOUNDARY 8.0-1 S.2 Dosos to MEMBERS OF THE PUBLIC - 40 CFR 190 8.0-2 ' 8.2.1 Effluent Dose Calculations ' 8.0-3 S.2.2 Direct Exposure Dose Determination S.0-3 S.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data 9.0-1 9.0 ASSESSMENT OF LAND USE CENSUS DATA 9.0-1 9.1 Land Use Census as Required by ODCM 3.12.2 , 9.0-3 9.2 Land Use Census to Support Realistic Dosa Assessment ( 10.0-1 10.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 10.0-1 10.1 Sampling Locations 10.0-1 10.2 Reporting Levels 10.0-2 10.2 Interlaboratory Comparison Program APPENDICES A-1 A Technical Basis for Effective Dose Factors Liquid Effluent Releases 9-1 B Technical Basis for Effective Dose Factors Gaseous Radweste Effluents TABLES 2.0-6 2.1 Surveillance Frequency Notation 2.0-7 2.2 Operational Conditions 3.0-5 3.3.7.11-1 Radioactive Liquid Effluent Monitoring instrumentation ; 3.0-7 4.3.7.11-1 Radioactive Liquid Effluent Monitoring instrumentation l Surveillance Requirements l
. 3.0-10 3.3.7.12-1 Radioactive Gaseous Effluent Monitoring instrumentation e
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3.0-14 4.3.7.12-1 Radioactive Gaseous Effluent Monitoring instrumentation Survalliance Requirements 3.0-19 4.11.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program 3.0-25 4.11.2.1.2-1 Radioactive Gaseous Waste Sampling and analysis Program 3.0-36 3.12.1-1 Radiological Environmental Monitoring Program 3.0-42 3.12.1-2 Reporting Levels for Radioactivity Concentrations in ; Environmental Samples l 3.0-43 4.12.1-1 Detection Capabilities for Environmental Sample Analysis 1 6.0-17 6.0-1 Fermi 2 Site Specific Liquid ingestion Dose Commitment ' Factors, Ago 6.0-19 6.0-2 Bioaccumulation Factors (BH) 7.0-15 7.0-1 Values for Evaluating Gaseous Release Rates and Alarm Setpoints . 7.0-16 7.0-2 Dose Factors for Noble Gases ' 7.0-17 7.0-3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations 7.0-18 7.0-4 Gaseous Effluent Pathway Dose Commitment Factors 8.0-7 S.0-1 Assumptions for Assessing Doses Due to Activities insicio SITE BOUNDARY S.0-8 S.0-2 Recommended Exposure Rates in Lieu of Site Specific Data 10.0-3 10.0-1 Radiological Enviror: mental Monitoring Program, Fermi 2 Sample Locations sad Associated Media 10.0-14 10.0-2 Radiological Environmental Monitoring Program, Fermi 1 Sample Locations and Associated Media . t A-4 A-1 Relative Dose Significance of Radionuclides in Liquid Effluents B-4 B-1 Effective Dose Factors - Noble Gas Effluents FIGURES 3.0-48 3.0-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents 6.0-20 6.0-1 Liquid Radioactive Effluent Monitoring and Processing Diagram 7.0-33 7.0-1 Gaseous Radioactive Effluent Monitoring and Ventilation Systems Diagram 10.0-16 10.0-1 Radiological Environmental Monitoring Program Sampling Locations - Alto Area 10.0-17 10.0-2 Radiological Environmental Monitoring Program Sampling Locations - Greater than 10 Miles 10.0-18 10.0-3 Radiological Environmental Monitoring Program Sampling Locations - within 10 Miles 10.0-19 10.0-4 Radiological Environmental Monitoring Program Sampling Locations - Site Area (Lake Erie side) 10.0-20 10.0-5 Formi 1 Sampling Locations l END OF SECTION 0.0
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Nuclear Production - Fermi 2 OOCM-10 C Offatte Dese Calculation Manu:1
- Revi len 5 i Pces 1.0-1 !
,-- INTRODUCTION
1.0 INTRODUCTION
. Part i of the Formi 2 Offsite Dose Calculation Manual (ODCM), which includes Sections 2.0 through 5.0, contains the controls and surveillance requirements for radioactive effluents
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and radiological environmental monitoring. It also contains requirements for the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release ! Report. l l Part 11 of the ODCM describes the methodology and parameters used in a) determining -l radioactive material release rates and cumulative releases, b) calculating radioactive liquid l and gaseous offluent monitoring instrumentation alarm / trip setpoints, and c) calculating the corresponding dose rates and cumulative quarterly and yearly doses. Calculational methods different from those provided in the ODCM may be used only if they lead to more i conservative results than would be obtained using ODCM methods. ODCM data for ; dispersion factors, receptor locations, exposure pathways, ventilation flow rates, etc. are Intended to lead to conservative results. However, it is permissible to use plant i procedures which implement the ODCM and which contain different data due 'to changes in ; environmental and plant conditions. i The methodology provided in Part 11 of this manual is acceptable for use in demonstrating compliance with the dose limits for members of the public of 10 CFR 20, the cumulative i dose criteria of 10 CFR 50, Appendix 1 and 40 CFR 190, and the controls in Part i of this : manual. !
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Part II, Section 6.0 of the ODCM describes equipment for monitoring and controlling liquid i offluents, sampling requirements, and dose evaluation methods. Section 7.0 provides ; similar information on gaseous effluent controls, sampling, and dose evaluation. ! Section 8.0 describes special dose analyses required for compliance with Fermi 2 Offsite ;
. Dose Calculation Manual and 40 CFR 190. Section 9.0 describes the role of the annual land ;
use consus in identifying the controlling pathways and locations of exposure for assessing l potential off-site doses. Section 10.0 describes the Radiological Environmental Monitoring . Program. ! t The ODCM will be maintained at Fermi 2 for use as a listing of radiological effluent ~ , controls and surveillance requirements, as well as a reference guide and training document for accepted methodologies and calculations. Changes to the ODCM calculational ; methodologies and parameters will be made as necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 60.36a and Appendix 1 for - demonstrating that radioactive effluents are "As Low As Reasonably Achievable." l NOTE: Throughout this document words appearing all capitalized denote either definitions ! specified in the Fermi 2 Controls or common acronyms. mr0 EAT 10N ONU ARMS - INFORMATION SERVICES Date approved: f .L / - f3 Release authorized by: d hML , Change numbers incorporated: 93-117-ODM / _,( nm
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DSN Ob C /% - /. Y Rev 5 Date DTC TMPLAN File 1715.02 Recipient M/,b a
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- ' ODCN-2.0' Nuclear Productiin - Fcrat 2 -
-Offcite Dose Calculction Manual Pag] 2.0-1
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[ARTI g RADIOLOGICAL EFFLUENT CONTROLS ARMS . INFO W
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r: 1 Nuclear Producti:n - Fcrci 2 ODCH-2.0 Off;ite Dose Calculction Manual ~ Revicion 5 Page 2.0-2
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A l . SECTION 2.0 l DEFINITIONS ! t l AIBIS - IWORNATION SERVICES Date approved: 9-21-93 Release authorized by:T Vanderney /s/ - Change numbers incorporated: 93-117-ODN
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DSN ODCN-0.0 Rev 5 Date 11-4-93 DTC TMPLAN File 1715.02 Recipient ; 1 1 i I
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- Nuclear.Productica - Fcrai 2 ODCH-2.0 l
'Revicion 5 Offcite Dose Calculation Manual Page 2.0-3 2.0 BEFINITIONS l
ACTION 2.1 ACTION shall be that part of a specification which prescribes remedial j measures required under designated conditions. ; OIANNEL CALIBRATION. 2.4 A Nam- CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range'and accuracy , to known values of the parameter which the channel monitors. The mAM. ; CALIBRATION shall enoonpass the entire channel including the sensor and ' alarm and/or trip functions, and shall include the manuEL FUNCTIONAL TEST. The CHANNEL CALIBRATION any be performed by any series of ' sequential, overlapping, or total channel steps such that the entire channel is calibrated. Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors shall consist of
. verification of operability of the sens,ing element and adjustment,'as necessary, of the remaining adjustable devices in the channel.
CHANNEL CHECK 4 2.5 A N A8uRL CHECK shall be the qualitative assessment of channel behavior ; during operation by observation. This determination shall include, where 1 possible, comparison of the channel indication and/or status with other ( indications and/or status derived from independent instrument. channels measuring the same parameter. CHANNEL FUNCTIONAL TEST 2.6 A CHANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY .
including alara and/or trip functions and channel failure trips.
- b. Ristable channels - the injection of a simulated signal into the ,
sensor to verify-OPERASILITY including alars and/or trip functions. The CHANNEL FUNCTIONAL TEST any be performed by any sequential, overlapping, or total channel steps such that the entire channel is tested. DOSE GQUIVALENT I-131 2.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, alorocuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation 4 shall be those listed in Table III of TID-14844, " Calculation of Distance
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Factors for Pouer and Test Reactor Sites." l FREQUENCY NOTATION
' 2.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 2.1.
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Nuclear Producticn - Farci 2 ODCN-2.0 Offcite Dose Calculation Manual Revisien 5 l Pcg2 2.0-4 l 2.0 DEFINITIONS 4 IEMBER(S) 0F 1HE PUBLIC ' 2.21 MEMBER (S) 0F THE PUBLIC shall be an individual in a controlled or UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC during any period in ubich the individual receives an occupational dose. OFF-GAS TREATNENT SYSTEM. ' 2.23 An 0FF-GAS TREATMENT SYSTEM is any systes designed and installed to reduce radioactive gaseous effluents by collecting reactor coolant system offgases from the reactor coolant and providing for delay or holdup for i the purpose of reducing the total radioactivity prior to release to the environment. OFFSITE DOSE CALCULATIONAL MANUAL 2.24 The OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) shall contain the methodology
, and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid affluent monitoring alara/ trip setpoints, and in the conduct of the l
radiological environmental sonitoring program. The ODCN shall also contain (1) the Radiological Effluent Controls and Radiological Environmental Monitoring Program Controls required by Technical Specification 6.8.5, and (2) descriptions of the information that should I be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Reports required by Controls 5 9.1.7 and 5 9.1.8. 1 OPERABLE - OPERABILITY 2.25 A systes, subsystem, train, component, or device shall be OPERABLE or have ! OPERABILITY when it is capable of performing its specified function (s) and ' when all necessary attendant instrumentation, controls, electrical power, l cooling or seal water, lubrication or other auxilliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). OPERATION'AL CONDITION - CONDITION 2.26 An OPERATIONAL CONDITION, i.e., 00NDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 2.2. PURGE - PURGING 2 31 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. h 4 4 **
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Nuclear Production - Fcr:1 2 ODCM-2.0 Offcite Dose Calculstion Manual Revi::1cn 5 Paga 2.0-5 l k 2.0 DEFINITIONS RATED ThERNAL POWER 2 32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant equivalent to that stated in the Technical Specification definition of RATED THERMAL POWER. R'EPORTABLE EVENT
- 2 34 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
SITE BOUNDARY 2 38 the SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled, by the licensee. SOURCE CHECK 2.40 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. THERMAL POWER 2.42 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. g UNRESTRICTED AREA 2.45 The Fermi 2 Energy Center UNRESTRICTED AREA includes all areas outside the site boundary. VENTILATION EXHAUST TREATMENT SYSTEM 2.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any systes designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust streaa prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EIHAUST TREATMENT SYSTEM components. VENTING 2.47 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, hualdity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in systes names, does : not imply a VENTING process. we
Nuclear Production - Fcrai 2 ODCM-2.0 l Offeite Dose Calculation Manual Revielon 5 Page 2.0-6 k. TABLE 2.1 SURVEI11ANCE FREQUENCY NOTATION NOTATION FREQUENCY 3 At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. M At least once per.31 days. Q At least once per 92 days. SA At least once per'184 days. A At least once per 366 days. R At least once per 18 months (550 days). S/U Prior to each reactor startup.
-l P Prior to each radioactive release.
N.A. Not applicable. T e e l > 4
!*- . Nuclear Productica - Fcr0i 2~ . ODCN-2.0 Offcite Dose Calculstion Manual Revi:lon 5 Fag) 2.0-7 w - $ 2.0 DEFINITIONS .
TABLE 2.2- , , j t OPERATIONAL CONDITIONS ADDE SWITCH AVEPACE REACTOh CONDITION POSITION COOLANT TEMPERATURE
- 1. POWER OPERATION Run Any temperature
- 2. STARTUP Startup/ Hot Standby Any temperature 3 907 SHUTDOWN Shutdown #, *** > 200 degrees F
- 4. COLD SHUTDOWN. Shutdown #, ##, *** $ 200 degrees F
- 5. REFUELING
- Shutdown or refuel **,# $ 140 degrees F d The reactor mode switch may be placed in the.Run, Startup/ Hot Standby, e or Refuel position to test t.he switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator or'other. technically qualified aesbar of the unit technical staff. i di The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure '1 vessel per Technical Specification 3 9.10.1. 1
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
## See Special Test Exceptions 310.1 and 310 3 of Technical Specifications.
*** The reactor mode switch may be placed in the Refuel position uhile a single control rod is being recoupled or withdrawn provided that.the one-rod.out interlock is OPERABLE.
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Nuclear Production - Fcrai 2 ODCM-3 0 Off;ite Dose Calculation Manual Rev10 ion 8 Pago 3 0-1 1
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l e SECTION 3.0 CONTROLS AND SURVEILLANCE REQUIREMENTS k INFORM!Til0N ONLY ARNS - INFORMATION SERVICES Date approved: Y[2[N/ Release aut.horized by: Id r I Change numbers incorporated: j 94-058-0DM DSN. Odd M'8.M Rev 8 Date ADD 97'100I DTC TMPLAN File 1715.02 Recipient M M A , t' I
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! ' Nuclear Production - Fcrai 2 ODCN-3 0 Offcite Dose Calculation Manual Revicien 8 Pag > 3 0-2 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS
'3/4.0 APPLICABILITY ;
M 4rOLS 3 0.1 Compliance with the succeeding Controls is required during the OPERATIONAL CONDITIONS or other conditions F Mcified therein; except that upon failure to meet the control, the associated ACTION requirements shall be set. 3 0.2 Noncompliance with a Control shall exist when the requirements of the control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the Action reonirements is not required. 303 When a Ceatrol is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the control does not apply by placing it, as applicable, in:
- 1. At least STARTUP within the next 6 hours, t 2. At least HOT SHUTDOWN within the following 6 hours, and 3 At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Control. Exceptions to , these requirements are stated in the individual Controls. This Control is not applicable in OPERATIONAL 00NDITIONS 4 or 5. 3 0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall , i not be made when the conditions for the Controls are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with the ACTION requirements when conformance to then permits continued operation of the facility for an unlimited period of time. This provision shall l not prevent passage through or to OPERATIONAL CONDITIONS as required l to comply with ACTION requirements. Exceptions to these requirements l
' are stated in the individual Controls.
Nuclear Production - Forni 2 . ODCN-3 0 Offcite Dose Calculation Manual Revicicn 8 , Pago 3 0-3 f . , APPLICABILITY I SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATI0llAL 00NDITIONS or other conditions specified for individual Controls : unless otherwise stated in an individual Surveillance Requirement. l 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillanos interval with a maximum allouable extension not to exooed 25 percent of the specified surveillance interval. 4.0 3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Surveillance 4.0.2, shall constitute nonocapliance with the OPERABILITY requirements for a Control. The : time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours to permit the completion of the surveillance when the allouable outage time limits of the ACTION requirements are less than 24 hours. , Surveillance Requirements do not .5 ave to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable I condition shall not be made unless the Surveillance Requirement (s) associated with the Control have been performed within the applicable surveillance interval or as otherwise specified. This provision shall '
- not prevent passage through or to OPERATIONAL CONDITIONS as required l
to comply with ACTION requirements. l l \: f ! l f { __
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Nuclear Production - Fermi 2 ODCH-3 0 Offcite Dose Calculation Manual ~ Revision 8 - Pag) 3 0-4 TMSTRUMDf?ATION RADI0 ACTIVE LIQUID DTLUENT MONITORING INSTRUNDITATION suminULS 3 3 7 11 The radioactive liquid affluent monitoring instrumentation channels shown in Table 3 3 7 11-1 shall be OPERABLE with their alars/ trip setpoints set to ensure that.the limits of Control 3 11.1.1 are not exceeded. The alara/ trip setpoints of these channels shall be determined and adjusted in accordance with , the methodology and parameters.in the OFFSITE DOSE PAf Pin ATIONAL MANUAL (ODCM). ! APPLICABILITY: At all times, except as noted for Condensate Storage Tank Discharge Monitor and Flow Rate Measurement Devloe. g/g I I ACTION: '
- a. With a radioactive liquid effluent monitoring instrumentation channel alara/ trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3 3 7.11-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain why this inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.
- c. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 43711 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, essainRL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4 3 7 11-1. l s.
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0D08 3.0 Revtoten O Pope 3.0 TASLE 3.3.T.11-1
. RADI0 ACTIVE LIOt#ID EFFLUENT HONITORING INSTRUIRENTATION MINIt0Utt CHANNELS INSTRUnfENT DeERABLE ACTION
- 1. GROSS RADI0 ACTIVITY WONITORS PROVIDING ALARM AND AUTOIIATIC YERMINATION OF RELEASE
- e. Lleguld Redweste Effluent Line Ott-NOO7 1 110
- 2. GROSS RADIOACTIVITY te0NITORS PROVIDING ALARM BUT te0T PROVIDING AUT0 STATIC TERMINATION OF RELEASE
- e. Ctreutettne Water Reservetr Decent Line D11-N402 1 111
- b. Condensate $tersee Tank Otecheree Path (1) (2) 1 113- 9/T'f i
- 3. FLON RATE ISEASUREteENT DEVICES
- e. Lisputd Redweete Effluent Line G11-R703 1 112
- b. Ctreutettne Water Reservetr Decent Line N71-R802 1 112
- c. Condensate Storese Tank Diecherse Path (2) 1 112 f y[TY i
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~ Nuclear. Production - Fcrai 2 ODCM-3 0 Offcite Dose Calculaticn Manual Revisicn 8
-Paga 3 0-6 TABLE 3.3.7.11.1 (Continued)
TABLE NOTATIONS ACTION 110 - With the number of channels OPERABLE less than required by the Minimus Channels OPERABLE requirement, effluent releases from , this pathway may continue provided that prior to initiating a release:-
- a. At least two independent samples are analysed in accordance with Control 4.11.1.1.1, and
- b. At least two technically qualified individuals independently verify the release rate calculations and discharge line valving; ,
Otherwise, suspend reluase of radioactive effluents via this pathway. ACTION 111 - With the number of channels OPERABLE less than the Minimus Channels OPERABLE requirement, radioactive affluent releases via this pathway any continue provided that grab samples are collected and analysed at least once per 12 hours for' gross radioactivit at least 10~y' alcrocurie/al, (beta or gaana) at a lower for Cs-137. limit of detection Otherwise, suspend of release of radioactive effluents via this pathway. ACTION 112 - With the number of channels OPERABLE less than required by the Minimus Channels OPERABLE requirement, effluent releases via this pathway any continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. Otherwise, suspend release of radioactive affluents via this pathway. ACTION 113 - suspend release of radioactive effluents via this pathway (1) - Prior to initiating a release, a) at least two independent samples are analysed in accordance with Control 4.11.1.1.1, b) the release rate calculations are reviewed by two technically qualified individuals, and c) the discharge line valving is independently verified by a technically qualified individual. In addition, the Condensate Storage Tank Discharge Monitor will be continually monitored to ensure that the tank is isolated promptly if the monitor alarms. (2) - The Condensate Storage Tank Discharge Path Monitor and Flow Rate Measurement Device need be OPERABLE only when the Condensate YhY Storage Tank discharge path is in use. l
ODCW 3.0
- Rew t a t en O ' -
Pese 3.C+T TABLE 4.3.7.11-1 *
. RADIGACTIVE LIGUID EFFLUESIT NOMITORING INSTRtmIENTATION SURVEILLANCE REGUIREM CHANNEL l FtMcCTIONAL i
CHANNEL. SOURCE CHAT #4EL INSTRUBIENT CHECK CHECK CALIBRATION TEST l 1. CROSS RADI0 ACTIVITY IBONITORS PROVIDING l AL7RM AND AUT00sATIC TERMINATION 8 l OF RELEASE
- e. Lleguld Redweste Effluent Line P P R(3) O(t)(2)
- 2. OROSS DETA OR SAtmeA RADIOACTIVITY RSONITORS PROVIDING ALAmes BUT NOT PROVIDING AUTOe4ATIC TERMINATION OF RELEASE t e. Ctreutettne Water Reservoir Decent Line D N R(3) O(5) l Ott-N402 i
l
- b. Condensate Storese Tank Diecherpe Peth (7) P P R(3) O(5) WTY
- 3. PLOW RATE NEASUREIBENT DEVICES (4)
- e. Ltquid Redweste Effluent Line O(4) N.A. R 0
- h. Circulottne Water Reservoir Decent Line D(4) M.A. R 0
- c. Condeneste Storese Tank Diecherge Path (T) O(4) M.A. R 0 '//"I Y l
4 5 4
. _ _ _ _ _ . . _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ , _ _ _ . _ _ . _ _ _ . _ _ - - _ _ ., . - . _ .. . . .- __-_-__m _ _ _ _ _ _
i Nuclear Productica - Fcr:1 2 ODCH-3 0 Offcite Dose Calculation Manual Revicicn 8 - Pag 3 3 0-8 TABLE 4.3.7.11-1 (Continued) TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if any of the following conditions i exists: 1., Instruarirt indicates measured levels above the alars/ trip setpoint.
- 2. Circuit failure.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alara ! annunciation occurs if any of the following conditions exists: 1
- 1. Instrument indicates measured levels above the alars s'etpoint.
- 2. Circuit failure.
3 Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode. l 1
(3) The initial CHANNEL CALIBRATION shall be performed using National i Institute of Standards and Technology traceable sources. These standards ; shall permit calibrating the systes over the range of energy and ! acasurement expected during normal operation and anticipated operational i occurrences. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration or are National Institute of Standards and Technology traceable shall be used. { t (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continous, periodic, or batch releases are made. l (5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alars i annunciation occurs if any of the following conditions exists: , i
- 1. Instrument indicates measured levels above the alars setpoint. l
- 2. Circuit failure.
{. 3 Instrument indicates a downsesle failure. ;
- .t (6) The CHANNEL FUNCTIONAL TEST shall also demonstrate that local alars occurs if any of the following conditiona exists: 1) instrument indicates i sensured levels above the alara setpoint, 2) circuit failure, or 3) instrument indicates a downscale failure. )
(7) Surveillance requirements for the Condensate Storage Tank Discharge Path Monitor and the Condensate Storage Tank Discharge Path Flow Rate Nessurement Device shall be met when the Condensate Storage Tank discharge '/[9Y path is in use.
Buolear Production - F rci 2. ODCM-3 0 . Off;ite Dose Calcul0 tion Manual 'Revicicn 8 l Pa33 3 0-9 j I INSTRUMENTATION RADIDACTIVE GASEQUS EITLUENT MONITORING INSTRUMENTATION vum e- *
^
'3 3 7 12 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3 3 7 12-1 shall be OPERABLE with their alara/ trip setpoints set i to ensure that the limits of Control 3 11.2.1 are not exceeded. The alarm / trip i satpoints of these channels, with the exception of the offgas monitoring {
system, shall be determined and adjusted in accordance with the methodology and ; parameters in the ODCN. j APPLICABILITY: As shown in Table 3 3 7.12-1 l ACTION: i
- a. With a radioactive gaseous effluent monitoring instrumentation channel i alarm / trip setpoint less conservative than required by the above Control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative. l l
- b. With less than the minimum number of radioactive gaseous effluent .
! monitoring instrumentation channels OPERABLE, take the ACTION shown in !
Table 3 3 7 12-1. Restore the inoperable instrumentation to OPERABLE l status within 30 days and, if unsuccessful, explain why this : l inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.
- c. 'the provisions of Controls 3 0 3 and 3 0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS j l 4.3 7.12 gach radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE i CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the l frequencies shown in Table 4.3 7.12-1. i i 1 l l I
-. _ ,_. - _. l
. ~ w.
00CM 3.0 Revteten S Page 3.0-10 l TABLE 3.3.7.12-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS i INSTRUMENT OPERABLE APPLICABILITY ACTION 1 REACTOR BUILOING EMMAUST PLENUM EFFLUENT MONITORING SYSTEM
- e. Lee Range Noble Gas Acttwity Montter -
Providtog Alarm t
- 121
- b. Iedine Someter 1
- 122
- c. Particulate Sempler 1
- 122
- d. Sampler Flee Rate Monttor 1
- 123
- 2. OFFGAS MONITORING SYSTEM (At the 2.2 minute celey piping)
- e. Noble Gee Acttwity Monitor - 1 *** 126 Providing Alarm
- 3. STAND 9V GAS TREATMENT SYSTEM
- e. Lew Range Neble Gas Acttwity Monttor ## -
Prowlding Alarm 1 # 125 f b. todine Sampler 1 s 122
- c. Perticulate Sampler 1 # 122
- d. Seaster Flow Rete Monitor 1 # 123
- a. .m OOCN-3.G Cow 131en 3
-Pese 3.0-11 .-
'a TA9tt 3.3.7.12-1 (Continued)
RADIOACTIVITV GASEOUS EFFLUENT MONITORING INSTRUMENTATION minimum CHANNELS j INSTRUutNT OPERABLE APDLICA9ILITY ACTION
- d. TUR91NE BLOG.' VENTILATION WONITORING SYSTEM
- e. Low Range Noble Gas Activ9ty monitor - 1
- 121 Providing Alarm *
- b. Iodine Sampler 1
- 122
- c. Particulate Sampler 1
- 122
- d. Sempler Flow Rote Monitor 1
- 123
- 8. SERVICE SUILDING VENTILATION MONITORING SYSTEM
- e. Low Range Noble Gee Activity Monitor - 1
- 121 Providing Alarm
- b. Iodine Seseler 1
- 122
- c. Particulate Sampler 1 * -
122
- 6. Sampler Flow Rete Montter 1
- 123 I
- - - - _ _ _ . - - - _ - - - - _ _ - - - - - - _ - - - - - - -e..
r OoCm-3.0 - Revicien S
' Pope 3.0-12 i
TABLE 3.3.T.12-1 (Continued) RADIOACTIVE GASEOUS EFFLUENT WONITORING INSTRUNDITATION. MINIMUM CHANNELS INSTeUntNT OPERABLE APPLICABILITY ACTION.
- 6. RADNASTE BUILDING VENTILATION MONITORING .
Sv5 TEM ! e. Low Range Noble Gas Activity Montter -
- 1 121 1 Providing Alarm i
- b. Iodine Sempler i
- 122
- c. Particulate Sampler 1
- 122
- d. Sempler Flow Rete Monitor 1
- 123 T. ONSITE STORAGE SUILDING VENTILATION .
EXHAUST RADIATION MONITOR
- e. Low Range Nebte Gee Activity Montter - 1
- 121 Providing Alarm
- b. Iedtne Segler 1
- 122
- c. Particulate Sempler i
- 122
- d. Someier Flow Rete monitor 1
- 123
Nuclear Production - Fcrai 2 ODCH-3 0 Offaite Dose Calculetion Manual Revicion 8 ) Page 3 0-13 l TABLE 3.3.7.12-1 (Continued) TABLE NOTATIONS
- At all times.
es Not used. l
*** During operation of the main condenser air ejector. i i During operation of the standby gas treatment system.
di Also included in Technical Specifications Table 3 3 7.5.1 Item 13.a. ACTION STATEMENTS
' ACTION 121 - With the number of channels OPERABLE less than required by the Minimus Channels OPERABLE requirement, effluent releases via l this pathway may continue provided grab samples are taken at l 1 east once per 12 hours and these samples are analyzed for gross l activity within 24 hours. Otherwise, suspend release of ;
radioactive effluents via this pathway. ACTION.122 - With the number of channels OPERABLE one less than required by i the Minimus Channels OPERABLE requirement, effluent releases via ' this pathway may continue provided that within 8 hours samples are continuously collected with auxiliary sampling equipment as required in Table 4.11.2.1.2-1. ACTION 123 - With the number of channels OPERABLE less then required by the Minimus Channels OPERABLE requirement, affluenc releases via this pathway may continue provided the flow rate is estimated at ' least once per 4 hours. Otherwise, suspend release of radioactive affluents via this pathway. ACTION 124 - Not used. ACTION 125 - With the number of channels OPERABLE less than required by the - Minimum Channels OPERABLE requirement, effluent releases via this pathway any continue provided grab samples are taken at least once per 4 hours and these samples are analyzed for gross l activity within 24 hours. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 126 - With the number of channels OPERABLE less than required by the Minlaus Channels OPERABLE requirement, releases via this pathway to the environment may continue for up to 7 days provided that:
- a. 'the offgas system is not bypassed, and -
- b. The reactor building exhaust plenum noble gas affluent (downstream) monitor is OPERABLE; -
Otherwise, be in at least HOT STANDBY within 12 hours. I i I i
)
a 0008 3.0 Revteton e Pese 3.0-14 TAOLt. 4.3.7.12-1 RADIOACTIVE SA5EOUS EFFLUENT WONITOP INS INSTRUMENTATION StJRWEILLANCE REQUIREMEff73 CHANNEL N00E3 IN WHICH CHANNEL SOURCE CHA*NEL FUNCTIONAL SURVEILLANCE INSTRUNENT CHECM CHECM_ CAL!9 RATION TEST REQUIRED 1 REACTOR DUILDING EKHAUST PLENUM
- e. Low Range Nebte Gas Activtty Monitor -
Prov9 ding storm D M R(2) Ott) '*
- b. Iodine Sempler w N.A. N.A. N.A. *
- t. Particulate Sampler W N.A. N.A. N.A. *
- d. Sempler Flow Rete montter D N.A. R 0 *
- 2. OFFGAS m0NITORING SYSTEM (At the 2.2 minuto deley piping) l
- e. Nebte See Activity Monitor D N R(2) O(1) ***
- 3. STAN09V GAS TREATWENT MONITORING SYSTEN
- e. Lou Renee Notte Gee Activity Montter D M R(2) Ott) a
- b. Iodine Sesoter W N.A. N.A. M.A. a
- c. Particulate Sempler W N.A. N.A. N.A. #
l
- d. Sempter Fles Rate montter D N.A. R 0 #
l I J
. . - - - . , , .-. - , - . . . ~ . ,, , , . - - - --- . - . . - . - . - - - - - . . _- - _ - - - - - - - - - - _ . - - .
~ '
00CW-3.G 4 RevtSten 8. Poes 3.0-t5-TASLE 4.3.T.12-1 (Continued) RADIOACTIVE._ GASEOUS EFFLtENT WONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL N00ES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECM CHECK CALIBRATION TEST REQUIRED 4 TUR8tNE BLDC. VENTILATION WONITORING SYSTEM
- e. Lee Range Nebte Gas Activity Montter D M R(2) O(4) *
- b. Iodine Songter W N.A. N.A. N.A. *
- c. Particutete Soupler W M.A. M.A. N.A. *
- d. Sengler Flow mate mentter O N.A. R O *
- 5. SERVICE SUILDING VENTILATION MONITORING SYSTEN
- e. Lee Re> cts Nebte Gee Activity monttor D M R(2) O(4) *
- b. Iodine S e er W M.A. N.A. N.A. *
- c. Particulate Senotor W M.A. N.A. M.A. *
.. S e Fte. Rete on9ter o N.A. R o
- l ,
l l t w - w h+ -- - -- T e =5l- -a *'_=m-
-m
- goem.3,o . -
- ^
Reviaten e Pess 3.C t l . taste 4.3.T.12-1 (Cont 9nved) l RADIOACTIVE GASEOUS EFFLUENT WONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i CHANNEL MODES IN WHICH. CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE ( i INSTRUutNT CHECK CHECM. CALIBRATION TEST REOUIRED l
- 6. RADWASTE BUILDING VENTILATION MONITORING SYSTEM
- s. Lew Range Noble Gas Activity Montter D M R(2) O(4)
W M.A. N.A. *
- b. Iodine Sempler N.A.
- c. Particulate Sampler W N.A. N.A. N.A.
R 0 *
- d. Sampler Flow Rete monitor D N.A.
l T. ONSITE STORAGE SUILDING l W*.NTILATION EXHAUST RADIATION WON! TOR
- e. Lew Range Noble Gas Activity Monitor D M R(2) Ott) b, todine Sempler W N.A. N.A. N.A.
- W N.A. M.A. *
- c. Particutete Sampler N.A.
h 0 *
- d. Sampler Flow Rete monttor D N.A.
l
Nuclear Production - F;rsi' 2. ODCH-3 0 ' Offcite Dose Cali:ulation Manual Revicion 8 Pag) 3 0-17 d -TABLE 4.3.7.12-1 (Continued) TABLE NOTATIONS
#' ' At all times.
- es Not used.
*** During operation of the main condenser air ejector.
#- During operation of the standby gas treatment system. ,
(1) The CHAl0lEL FUNCTIONAL TEST shall also demonstrate that control room alara annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
- 2. Circuit failure.
3 Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode (alarm or. type).
(2) The initial CHANNEL CALIBRATION shall be performed using National-Institute of Standards and Technology traceable sources. These standards-( shall permit calibrating the.systes over the range of energy and measurement espected during normal operation and anticipated operational occurrences. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration or are National Institute of Standards and Technology traceable shall be used. (3) Not used. ., l (4) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation occurs on high level and that control room alars annunciation i occurs if any of the following conditions exists: l
- 1. Instrument indicates measured levels above the alara setpoints. .!
-i
- 2. Circuit failure. )
l 3 Instrument indicates a downscale failure.
- 4. Instrument controls not set in the operate mode (alars or type).
9 1
Buclear Productirn - C rai 2: ODCN-3 0 Offsite Dose Calculation Manual Revicien 8-Pago 3 0-18 3/4.11 RADI0 ACTIVE DTLUENTS 3/4.11.1 LIQUID DTLUENTS QQNCENTRATION w ci.s - 311.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 3 0-1) shall be limited to ten times the concentration' values specified in 10 CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2 for radionuclides other than dissolved' or entrained noble gases. For dissolved concentrationshallbelialtedto2I10grentrainednoblegases,the microcuries/a1 total activity. APPLICABILITY: At all times. ACTION: With the concentration of radioactive material released in liquid effluents to IBIRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits. I SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11.1.1.1-1. 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCN to assure that the concentrations at the point of release are maintained within the limits of : Control 3 11.1.1. l l t I i I as l
Nuclear Production - F;rai 2 ODCM-3 0 Offcite Dose Calculati:n Manual Revicion 8 Paga 3 0-19 i 1. TABLE 4.11.1.1.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)* Type Freauency Freauency Analysis (uC1/al) A. Batch P P Releaseb : Each Batch Each Batch Principal Gamma 5 1 10-7 Waste Emittersc , Sample Tanks (3) or 1-131 1 1 10-6 Condensate Storage P M Dissolved and 1 I 10-5 Tank One Batch /M Entrained Gases
, (flan =a Esitters)
P M H-3 1 I 10-5 Each Batch Composite d Gross Alpha 1 1 10-7 i P ~Q Sr-89, Sr-90 5 I 10-0 Ea?h k;ch Composite d Fe-55 1 I 10-6 B Continuous Releases
- M PrincipagGamma 5 I 10-7 Circulating NA Emitters Water Composited System (if I-131 1 1 10-contaminated)__.
W M Dissolved and 1 I 10-5 Grab Sample Entrained Gases (Gamma Esitters) M H-3 1 1 10-5 ! NA l Composite d Gross Alpha 1 I 10-7 Q Sr-89, Sr-90 5 I 10-8 NA Composite d Fe-55 1 I 10-6 i s.
Nuclear Production - FCrai 2 ODCN-3 0 Offcite Dose Calculation Manual Bevision 8 Page 3 0-20 E TABLE 4.11.1.1.1-1 (Continued) i TABLE NOTATION aThe ILD is defined, for purposes of these controls, as t,he smallest aaaamatration of radioactive material in a sample that will yield a not count, . .. above systen background, that will be detected with 955 probability with only 55 probability of falsely concluding that a blank observation represents a a real" signal. For a particular measurement system, which any include radiochemical separation , 4.66 sb LLD E V
- 2.22 x 106 . y . exp (- A t) r Where:
LLD is the "a prioria lower limit of detection as defined above, as ' microcuries per unit mass or volume, sb is the standard deviation of the background counting rate or of the ( counting rate of a blank sample as appropriate, as counts per minute,
,E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochoalcal yield, when applicable, Aistheradioactivedecayconstantfortheparticularradionuclide,and t for plant effluents is the elapk d time between the midpoint of sample collection and time of counting.
Typical values of E, V, Y, and t should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an 3, : posteriori (after the fact) limit for a particular measurement. bA batch release is the discharge of 11guld wastes of a discrete volume. ' Prior to sampling for analysee, each batch shall be isolated, and then thoroughly sized by a method described in the ODCM to assure representative , sampling. Batch liquid discharge may be made from only one tank at a time. lt/'Ty I ( l
~ Nuclear Production - Fcrci 2 ODCM-3 0 Offcite Dose Calculation Manual Revicion 8 Pag 3 3 0-21 TABLE 4.11.1.1.1-1 (Continued)
TABLE NOTATION
*The principal samma esitters for which the 11D specification applies '
esclusively are: Mn-54, Fe-59, C0-58, co-60, 2n-65, Mo-99, Cs-134 Cs-137, Co-141, and Co-144. This does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analysed and reported in the Annual Radioactive Effluent Release Report pursuant to Control 5 9 1.8. d A composite sample is one in which the quantity of liquid samples is proportional to the quantity of liquid unste discharged and in which the method of sampling employed results in a specimen that is representative of the-liquids released. This any be accomplished through composites of grab samples obtained prior to discharge after the tanks have been recirculated.
'A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of a system that has an input flow during the continuous release.
1 k W
- . . +
L ,15uclear Production - F;rai 2 ODCN-3 0 l Offaite Dose Calculation Manual Revision 8 Pag) 3 0-22 [ hAD10 ACTIVE EFFLUENTS B0sE
-Emia0LS 3 11.1.2 The dose or dose commitment to a NENBER OF THE PUBLIC from i radioactive materials in liquid effluents released, from each reactor unit, to IBDIESTRICTED AREA 5 (see Figure 3 0-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1 5 areas to the total body and to less than or equal to 5 areas to any organ, and 1
- b. During any calendar year to less than or equal to 3 areas to the i total body and to less than or equal to 10 areas to any organ. '
APPLICABILITY: At all times.
- ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical ,
Specification 6.9 2, a special Report that identifies the cause(s) l ( for exceeding the limit (s) and defines the corrective actions that i have been'taken to reduce the releases and the proposed corrective. ' actions to be taken to assure that subsequent releases will be in i' 2 compliance with the above limits. 'this Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.e
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable. )
l SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar y nr shall be determined in accordance with the methodology and param(ers in the ODCN at least once per 31 days. - l l
' Applicable only if drinking water supply is taken from the receiving water l body within 3 miles of the plant discharge. I l
i
~ ' ~ ~ ~ ~~ ~ ~ ~
~ NuclearProduction-'Fdrai'2; ~ ,
ODcM-3 0 "^l
, J0ffcite Dose Cal:ulation Manual Revicion 8.. I' Pago 3 0-23 t
RADI0 ACTIVE EFFLUDITS LIQUID WASTE TREATMDfT GUnindS , 3 11.1 3 The liquid radwaste treatment system shall be OPERABLE and , appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see > Figure 3 0-1) would exceed 0.06 ares to the total body or~0.2 area to any organ ; in any 31-day period. ; APPLICABILITY: At all times. - ACTION: I
- a. With radioactive liquid waste being discharged and in excess of.the above limits and any portion of the liquid radwaste treatment system not in operation, prepare and submit to the Commission within 30 days pursuant to Technical Specification 6.9.2 a special Report that '
includes the following information: ( 1. Explanation of why liquid radwaste was being discharged ! without complete treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.
- 2. Action (s) taken to restore the inoperable equipment to ,
OPERABLE status, and 3 sussary description of action (s) taken to prevent a ,, recurrence. ,
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l 4.11.1 3 1 Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCH. - l '4.11.1 3 2 The installed 11guld radwaste treatment systen shall be L . demonstrated OPERABLE by meeting Controls 311.1.1 and 311.1.2. l l . . . . . . . . . . .: . . i _ ._
r ~ -
- - ^ - ~ ~ ~~ ~ ^~' ~~
}
o Woolear Production'- Fcrai 2
' 00CN-3 0' :
Offcite Dose Osloulation Manuil Revicien 8 l Pass 3 0-24 l ( RADICACTIVE' EFFLUENTS 3/4.11.2 GASEQUS EFFLUENTS 1 DOSE RATE I u minU63 3 11.2.1 The dose rate due to radioactive materials released in gaseous affluents from the site to areas at and beyond the SITE B0UNDARY (see Figure ' 3 0-1) shall be limited to the following
- a. For noble gases: Less than or equal to 500 aress/yr to the total
- body and less than or equal to 3000 aress/yr to the skin, and ,
b.- For iodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or > equal to 1500 aress/yr to any organ. APPLICABILITY: At all times. ACTION: With the dose rate (s) exooeding the above limits,'immediately restore the
-I release rate to within the above' limit (s).
SURVEILLANCE REQUIREMENTS . 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be ' determined to be within the above limits in accordance with the methodology and parameters in the ODCH. i 4.11.2.1.2 The dose rate due to iodine-131, lodine-133, tritium, and all other , radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2.1.2-1. l p 4 8 b b r f J l I. _ _ _
~ '"
ODCW 3.@ ~ Rewtaten 8 Pese 3.0-25
.4 TABLE 4.11.2.1.2-1 RAOf0 ACTIVE GASEOUS WASTE SAMPLING AND ANAL.YSIS PROGRAN Minimum Leeer Lielt of Sempting Analysts Type of Detectten (LLD)e .
Ges r r Rolesee Twee Frr_ _ a , Fr.__ m , Activity Anetyste (uCt/el) Pi, S3 Pi, $3 A. Centelnment PURGE Each PURGE Each P(URGE Principal Gemme Emittera b ln 10j 1 a 10 (Pro Treatment) Grab Semple H-3
- 8. Reactor Dutiding Mc.d.e ye Principe1 Gemme Emitters b g , gg-4 Enheuet Plenum Grob Sample MC H-3 1 a 10-6 Standby Gas Tgest-ment System C. Redoeste Dutiding M N Principet Genne Emittera b 1 a 10'8 Turbine Building Grab Sample m H-3 1 a 10-6 Servtco-sutIdtne On-Site Sterose Fact 19tv r
D. All Release Types Continuous' w9 I-131 1 m-10-12 me listed in 9 Adsorbent I-133 1a 10-10 and C obove. Sonet e Continuous' w9 Principal Gemme Emittera b 1a 10*II Particulate (I-131 others) Semple Continuous f m Gross Alphe 1 a 10' " Conces 1 t e f, Particulate Semple Continuous' O Sr-99 SP-90 1 a 10'II compos 1te PertIcutete semple Continuous' Noble Gee Noble Gas 1 a 10-6 monttor Gross Bote or Gemme E. Weste Ott P P Principet Gemme Emitters b 5a 10'I Incineretten by Each Batch Each Batch I-131 1 a 10'O Injection into Weste Ott Austltery Sotter Ltquid Sample Fuel Stroom
, 2 o -
o .,,s , m, a ~ a- ..-,. f I LNuclear Production - Fgrei 2 ODCN-3 0 l
!Offcite Dose Calculation Manual Revicion 8 Pag 3 3 0-26
. TABLE 4.11.2.1.2-1 (Continued)
TABLE NOTATION ahe LLD is defined, for purposes of these controls,'as the smallest conoontration of radioactive material in a sample that will yield a not count, , above system background, that will be detected with 955 probability with only l 55 probability of falsely concluding that a blank observation represents a , ;
- "real" signal. ;
For a particular measurement system, which any include radiochemical i separation: !
-4.66 sb ILD * .
E
- V
- 2.22 x 100*Y*exp(-At) 3 Where:
ILD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, sb is the standard deviation of the background counting rate or of the l counting rate of a blank sample as appropriate, as counts per minute, s E is the counting efficiency, as counta per disintegration, T is the sample size in units of mass or volume,, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radlochemical yield, when applicable, Aistheradioactivedecayconstantfortheparticularradionuclide,and , t for plant effluents is the elapsed time between the midpoint of sample collection and time of counting. . Typical values of E, Y, Y, and t should be used in the calculation. It should be recognized that the ILD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. b The principal samma emitters for which the LLD specification applies exclusively are the following radionuclides 'Er 87 Kr 88, Ie-133, Ie-133m, 4 Ie-135, and re-138 in noble sas releases and Mn-54, Fe-59, co-58, co-60, 2n-65, i Mo-99, I-131, cs-134, cs-137, co-141, and co-144 in iodine and particulate i releases. h is list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analysed and reported in the Annual l Radioactive Effluent Release Report pursuant to control 5.9.1.8. e
^^ ~ ^ ~ Nuclear Production - Fcrmi 2 ODCN-3 0 Offcite Dose Calculation Manual Revision 8 P g3 3 0-27
, TABLE 4.11.2.1.2-1 (Continued),
TABLE NOTATION
" Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3 d
Tritius grab samples shall be taken at least once per 24 hours when either the reactor well or the dryer-separator storage pool is flooded.
'Tritius grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.
I The ratio of the sample flow rate to the sampled streas flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls 3 11.2.1, 3 11.2.2, and 3 11.2 3 8Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least 7 days g following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas sonitor shows that effluent activity has not increased more than a factor of 3 h Required when the SCTS is in operation. I In OPERATIONAL CONDITIONS 1, 2, 3, and 4, the applicable portion or primary containment shall be sampled and analyzed within 8 hours prior to the start of any PURGING. dIn OPERATIONAL CONDITIONS 1, 2, 3, and 4, when the primary containment staosphere radiation monitoring system is declared INOPERABLE or is in alara condition, the applicable portion of primary containment shall be sampled and analyzed within 8 hours prior to the start of any VENTING or PURGING and at least once per 12 hours during VENTING or PURGING through other than SGTS.
, g
- a ==
.+.
7 '
, _ . . -- .- . . - - - -- _- .__. ~ -
- Euclear Production - F;;rmi 2 ODCN-3 0 .
~
0.ffaite Dose Calculation. Manual Revicion 8' Paga 3 0-28 , RADI0 ACTIVE EFFLUENTS DOSE - 50BLE GASES wninUi.S 3 11.2.2 The air dose due to noble sases released in saseous affluents, from i each reactor unit, to areas at and beyond the SITE B0UNDARY (see Figure 3 0-1) , shall be limit.ed to the following:
- a. During any calendar quarter: Less than or equal to 5 stads for samma radiation and less than or equal to 10 mrads for beta radiation and,
- b. During any calendar year: 14ss than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times. ' ACTION: i l
- a. With the calculated air dose from radioactive noble gases in gaseous ;
effluents exceeding eny of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for esoeeding ( the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable. ;
i SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current omiendar year for noble gases shall be determined in scoordance with
- the methodology and parameters in t,he ODCN at least once per 31 days.
I
- v- - - - - - ---,_m_ ,-
, Euclear Production - Fcrai 2 ODCM-3 0 Off0ite Dose Calculation Manual Revicien 8 Pag > 3 0-29 t
RADI0 ACTIVE EFFLUENTS GASEOUS DTLUENTS DOSE - 10 DINE-131. 10 DINE-133. TRITIUN. AND RADIONUCLIDES IN PARTICULATE FORM wnanOLS 3 11.2 3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and ; beyond the SITE BOUNDARY (see Figure 3 0-1) shall be limited to the following:
- a. During any calendar quarter: IAss than or aqual to 7.5 areas to any organ and, -
- b. During any calendar year: Less than or equal to 15 areas to any organ.
APPLIC' ABILITY: At all times. ACTION:
- a. With the calculated dose from the release of iodine-131, iodine-133.
I tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. .
- b. The provisions of Controla 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2 3 Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-131, todine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCN at least once per 31 days.
..,. - - - -~ ,
Nuclear Productien'- Fcrai 2 ODCM-3 0 4 Offd te Dose Calculction Manual Revicion 8 I Pas) 3 0-30
! RADI0 ACTIVE EFFLUDfTS
.OFF-GAS TREATMENT SYSTEM.
l M amJL5 1 311.2.4 . The CFF-GA5 TREATMDff SYSTEM shall be OPERABLE and shall'be in l operation. l APPLICABILITY: Whenever the main condenser stema jet air ejectors are in i operation. ! ACTION: 1
- a. With the OFF-GAS TREATMENT SYSTEM inoperable for more than 7 days, '
prepare and submit to the commission within 30 days, pursuant to ' Technical Specification 6.9.2, a special Report that includes the following information:
- 1. Identification of the inoperable equipment or subsystems and l the reason for the inoperability, i 1
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and , ;
1 3 summary description of action (s) taken to prevent a l recurrence. l,
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable. ,
- c. The provisions of Control 4.0.4 are not applicable.
1 SURVEILLANCE REQUIREMENTS 4.11.2.4 The OFF-GAS TREATMENT SYSTEN shall be demonstrated OPERABLE by meeting Controls 3 11.2.1, 3 11.2.2, and 3 11.2 3 1 I = ~ , .. . . . . . . ,-.... .- . -e -~ ~
~
. Nuclear Production - F;rsi 2 ODCM-3 0 Offsite Does Calculation Manual- ~Revicion 8 Pag 3 3 0 I RADI0 ACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM 3 11.2.5 The VENTILATION EKHAUST TREATNENT SYSTEN as described in the 0D01 shall be OPERABLE and appropriate portions of the systen shall be used to reduce radioactive materials in gaseous unste prior to their discharge when the projected doses due to gaseous affluent releases from the site to UNRESTRICTED- 1 AREAS (see Figure 3 0-1) would esoeed 0 3 ares to any orsan in any 31-day !
period. ! 1 APPLICABILITY: At all times.
-{
1 ACTION: l
- a. With radioactive gaseous waste being discharged in excess of the above limits and any portion of the VENTILATION EIHAUST TREATNENT SYSTEN not in operation, prepare and submit'to the Consission within 30 days pursuant to Technical Specification 6.9.2 a Special Report that includes tho' following information: )
i
- 1. Identification of any inoperable equipment or subsystems, and i I the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3 Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from the site shall be projected at l least once per 31 days in accordance with the methodology and parameters in the l ODQl, when any portion of the VENTILATION EIHAUST TREA1 MENT SYSTEN is not in use. 4.11.2.5.2 The YENTILATION EXHAUST TREATMENT SYSTEN shall be demonstrated OPERABLE by meeting Controls 3 11.2.1, 3 11.2.2, and 3.11.2 3 ( l l
' Nuclear Production - F:r::12 - ODCN-3 0 1
> Offcite Dose C lculation Manual Revicisn 8 Pag 3 3 0-32 .
'RADICACTIVE EPPLUENTS
];
. TENTING OR PURGING ,
1 N l 3 11.2.8 TENTING or PURGING of the primary containment shall be through the standby gas treatment system or the reactor building ventilation system. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,.3, and 4 ACTION:
- a. With the requirements of the above control not satisfied, suspend all TENTING or PURGING of the primary containment.
- b. The provision of Controls 3 0 3 and 3 0.4 are not applicable. ,
SURVEILLANCE REQUIREMENTS 4.11.2.8.1 The applicable portion of primary containment shall be sampled and analysed per Table 4.11.2.1.2-1 of Control 311.2.1 within 8 hours prior to the start of any PURGING.
~
( 4.11.2.8.2 If the primary containment radiation monitoring system is
-INOPERABLE or is in alara condition, the applicable portion of primary containment shall be sampled and analyzed per Table 4.11.2.1.2-1 of Control 311.2.1 within 8 hours prior to the start of and at least once por 12
. hours during VENTING or PURGING of primary containment through other than the standby gas treatment system.
4.11.2.8 3 The primary containment shall be determined to be aligned for VENTING or PURGING through the standby gas treatment system or the reactor building ventilation system within 4 hours prior to start of and et least once per 12 hours during VENTING or PURGING of the containment. 4.11.2.8.4 Prior to use of the vent / purge system through the standby gas ' treatment system assure that: . a. Both standby gas treatment system trains are OPERABLE whenever the l vent / purge system is in use, and I .
- b. Whenever the vent / purge system is in use during OPERATIONAL CONDITION l 1 or 2 or 3, only one of the standby gas treatment system trains any
, be used. ! l t l 4.11.2.8.5 Prior to TENTING or PURGING, assure that at least one of the following monitors is OPERABLE: the prianry containment atmosphere radiation monit.or, the reactor building ventilation exhaust radiation monitor (at least
, one division), or the SPING monitor corresponding to the release path (the ,
reactor building exhaust plenus radiation monitor or the standby gas treatment , system radiation monitor, Division 1 or 2). , l
- Nuclear Production - Fcr::12 ' ODCH-3 0 :
Offcite Dose Calculation Manual Reviaicn 8 Page 3 0-33
-( RADIDACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3 11.4 The annual (calendar year) dose or dose commitment to any member.of the public (as defined in 40 CFR Part 190) due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 areas to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 areas.
APPLICABILITY: At all times. , 1 ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous affluents exceeding twice the limits of Controls i 3.11.1.2a., 3 11.1.2b., 3 11.2.2a., 3 11.2.2b., 3 11.2 3a., or l 3 11.2 3b., calculations should be made including direct radiation !
contributions from the reactor units and from outside storage tanks to determine whether the above limits of Control 3 11.4 have been exooeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical specification 6.9.2, a special Report that defines the corrective action to be taken to reduce 1 I i subsequent releases to prevent recurrence of exceeding the above lialts and includes the schedule for achieving conformance with the above limits. 1his Special Report, as defined in 10 CFR 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a member of the public from uranius fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been oorrected, the special Report shall include a request for a varianoe in accordance with the provisions of 40 CFR Part 190. submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. .
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.11.1.2, 4.11.2.2, and 4.11.2 3, and in accordance with the methodology and parameters in the ODCH. I 4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from outside storage tanks shall be determined in accordance with the methodology and parameters in the ODCN. This requirement is applicable only under conditions set forth in Control 3 11.4, ACTION a. i
-s -- '"*
, - - - - . - . . . - .- .-- -= - - - .- - ~-
B . ' Nuclear Production - F;rci'2 ODCM-3 0
'Offsite Dose Calculaticn Manual- Revicien 8 Pag) 3 0-34 y ,
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING j 3/4.12.1 MONITORING PROGRAM i CONTROLS r
~3 12.1 the radiological environmental monitoring program shall be conducted as specified in Table 3 12.1-1.
APPLICABILITY: At all times.
. ACTION:
- a. With the radiological environmental monitoring progran not being conducted as specified in Table 3 12.1-1, prepare and subsit to the Commission, in the Annual Radiological Environmental Operating Report required by Control 5 9.1.7, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3 12.1-2 when averaged over any calendar quarter, prepare and submit to the Consission within 30 days, pursuant to Technical Specification 6.9.2, a special Report >
( that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose
- to A MEMBER OF THE PUBLIC is~1ess than the calendar year limits of Controls 3 11.1.2, 3 11.2.2, and 3 11.2 3 When more than one of the radionuclides in Table 3 12.1-2 are detected in the sampling medium, this report shall be submitted '
if: concentration (1'l + concentration (2'l + ...>
~
1.0 reporting leve:, (1) reporting leve:, (2) Uhen radionuclides other than those in Table 3 12.1-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose # to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Controls 3 11.1.2, 3 11.2.2, and 3 11.2 3 This report is not required if the sensured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be' reported and described in the Annual Radiological Environmental Operating Report pursuant to Control 5 9 1.7. I
- c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3 12.1-1, identify ,
specific locations for obtaining replacement samples and add them to I the radiological environmental monitoring program within 30 days.
, The specific locations from which samples were unavailable may then
'The methodology used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
l i'
-Nuclear Production - F rsi 2- ODCN-3 0 Offsite Dose Calculation Manual
- Revisitn 8
, Paga 3 0-35 , RADIOLOGICAL ENVIRONMENTAL MONITORING - CONTROLS (Continued) be deleted from the sonitoring program. Pursuant to Control 5 9 1.8, identify the cause of the unavailability of ansples and identify the new location (s) for obtaining replacement samples in the next Annual
'Radioact've Effluent Release Report pursuant to Control 5.9 1.8 and also include in the report a revised figure (s) and tablo for the ODCN reflecting the new location (s).
9
- d. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIRDtENTS 4.12.1 The radiological environmental sonitoring samples shall be collected pursuant to Table 3 12.1-1 from the specific locations given in the table and : figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3 12.1-1 and the detection capabilities required by Table 4.12.1-1. ( l 4 o I
00Cm 3.0 - Ceviaton S Pese 3.0-34 . TABLE 3.12.1-1 RAOf0 LOGICAL ENVIRONMENTAL 504170R19eG PitOGRAN Mumber of Representattve Empesure Pethway semples and sempting and Type and Frequency end/or Semote 5 emote Locettons, Cottoction Freovency of Analysts 1 DIRECT RADIATIOND 37 routine monitoring stations. Duarterly Gemme dose cuarterly. with two or more dostmeters pieced as fntlows: 1) en inner ring of stations in the generet eres of the SITE BOUNDARY and additionet rings et approximatesy 2. 5. and 10 miles. ett>. e station in et tesh: =<ery other meteorological sector for each ring with the exception of those sectors over Lake Erte. The balance of the stations. 8 should be pieced in special interest areas such as population centers. neseby residences. schools. and in 2 or 3 areas to serve as control stations.
- 2. AIR 90RNE Radictedine end Semples from 5 tecettons. Continuous sempler Redtotodine Cennisters Particulates operetton with semple I-831 anstysts weekly.
- e. 3 semples from close to collection weekly. or the 3 SITE SOUNDARY loca- more frequently if tions, in different sectors, reoutred by dust Particulate Sempler of the highest calculated Iceding. Gross beto redtoactivity annual everage ground-level D/O.
enelysts folleging filter changes
- b. 1 semple from the victntty of Gemme isotopte enetysts' e community having the highest of compostte (by esiculated annual everage tocetton) quarterly.
groundlevei D/Q. 1
w ~ ODCE-3.0 - Cevtalen 9 Pope 3.@-37 TAet.E 3.12.1-1 (Continued) RADIOLOGICAL ENvlRONMENTAL IE000! Tost!NG PROGRAIB Mumber of Representettwe Esposure Pathway Semples and SemolIng and Type and Frequency and/or Somete_ 5 emote Locettons' Collection Freauency of Analysts c.1 semple from a control locetton, es for esemple 15-30 km distent and in the least prevalent wind directionC.
- 3. WATElt90RME'
- o. Surface e. I sogle upstream. Composite semple over Gamme isoteole snelysts'
- b. 1 semple downstream. 1-month period 9 monthly. Composite for trittum analysts quarterly
- b. Ground Semples from 1 or 2 sources Overterly Gemme footopic' and only tf Itkely to be tritium analysts
( effected h. Quarterly.
- c. Drinking e. 1 semple of each of 1 to Compostte semple 1-131 enelyste on each 3 of the nearest water uver 2-week period # compostte when the dose supplies that could be when 1-131 enetysts calculated for the effected by its discharge. ts performed. monthly consumption of the water compostte otherwise is greate 1 meem l
- b. 1 semple from e per year.y then Compostte f #
control locetton for gross bete end gewune 1 f isotopic enelyses' monthly. Composite for trittum analysts quarterly
- d. Sediment 1 semple from downstroom area Semtennually Gemme tootopic enelysts' from with entottog or potential somtennuelty.
shoreline recreettonal value. l l i
-- w- wr # _.____mr _- -___---__A -_ - _._u
8
~
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.povtaten 8 Pege.3.5-38 ,
TABLE 3.12.1-1 (Continued) l RA0f 01.0GTCAt. _ E88VIR08eIE88 tat. MONITOftf MG PetOGRAtl l Number et Representative Exposure pathway Semples and Sempting and Type end Frequency _ end/or Semple - Semple Locations
- Collection Frecuency __of Analvets l 4 INGESTION i s. Milk e. Semotes from milking entse r Semimonthly when Gemme isotopic' and in 3 locations within 5 km entmals are on I-131 enelysts sentmenthly distence having the highest pasture, monthly et when ontpels are on dose potential. If there other times pasturet monthly at are none. then. I semple other times.
from milking entmels in each of 3 areas between 5 to 8 km distant where doses are calculated to be g{ ester then 1 mrom per yr .
- b. 1 semple from milking l entmals et a control locetton 15-30 km distent and in the feest prevalent wind direction
- b. Fish and e. 1 semple of each commercially Semple in season. or Gemme tootopte snelysts'
. Invertebrates and recreettonelly importent semiennually if they on edible porttone.
species 13 vicinity of plant are not seasonal discharge area.
- b. I semple of some species in erees not influenced by plant discharge.
- c. Food e. 1 semple of each principal At time of hervest3 Gemme teetopte snelyses' Products class of food products from on edible portions.
any area that is trrigated by water in which flould plant westes have been discharged. d w - -s w e- . , e - .
A - 00 Css-3.8 . newtoten s. Pese 3.C+39-TR%E 3.12.1-1 (Centinued) alAD10 LOGICAL E9 WIN 0seIENTAL_ MONITOstileG Pet 0GstAN Number of Representettwo Esposure Pathway Semotes and Semp1tng and Type end Frequency __end/or Semele Semote Locettons' Cottection Frecuency of Anelvets
- c. Food b. Semples of 3 different kinds IPonthly when Gamme teetopic' and Products of brood leaf vegetetton grown ovellebte- I-131 enelysts.
(cent *d) neerest each of two different offette locettons of highest predteted annual everage level D/O if alth sempting is not performed.
- c. I semple of each of the steller Monthly when Genues testepic' and -
broad test vegetetton grown evettable I-131 enelysts. 15-30 km distent in the least prevalent wind direction if mtik sempting is not performed. 6 I
Nuclear Production - Farci 2 ODCH-3 0 Offcite Dose Calculaticn Manual Revicien 8-Page 3 0-40
! TABLE 3.12.1-1 (Continued)
TABLE NOTATIONS ASpecific parameters of distance and direction sectot from the centerline of one reactor, and additional description where partin6nt, shall be provided for each and every sample location in Table 3 12.1-1 in a table and figure (s) in the ODQl. Refer to.NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological a==aamt Branch Technical Position, Revision 1, November 1979. ' Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Contol 5.9.1.7. It is recognized that, at times, it any not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. Pursuant to Control 5 9.1.8, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next
~
Annual Radioactive Effluent Release Report and also include in the report a { revised figure (s) and table for the ODCM reflecting the new location (s). b one or more instruments, such as a pressurised ion chamber, for seasuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific
. system used and should be selected to obtain optinua dose information with minimal fading.
othe purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that have valid background data may be substituted. d Airborne particulate sample filters shall be analyzed for gross beta ' radioactivity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gaana isotopic analysis shall be performed on the individual samples.
' Gamma isotopic analysis means the identification and quantification of i
gamma-emitting radionuclides that may be attributable to the effluents from the facility.
. _ . . . . . . . ~ . . _ . . . . . - - - --
- Buclear Production - Forci 2 ODCH-3 0 Offcite Dose Calculation Manual Revicion 8 L Pag 3 3 0-41 TABLE 3.12.1-1 (Continued)'
TABLE NOTATION 1he "upstrema sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area ' beyond but near the airing sone. " Upstream" samples in an estuary must be taken far enough upstrema to be beyond the plant influence. gComposite samples should be collected with equipment (or equivalent) uhich is capable of collecting an aliquot at time intervals that are very short (e.g., bourly) relative to the compositing period (e.g., monthly). ; hounduatersamplesshallbetakenwhenthissourceistappedfordrinking ; or irrigation purposes in areas where the hydraulic gredient or recharge properties are suitable for contamination. A The dose shall be calculated for the easinua organ and age group, using the methodology and parameter.3 in the ODCN. d If harvest occurs more than once a year, sampling shall be performed during I each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products. i i i l 1
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ODCm 3.0 Revitten 8
- Pese 3.0-43 TABLE 4.12.1-1 DETECiteet CAPAstLTTIES Feet Ewv!RONMENTAL SAMPLE ANALYSIS
- lower LIMIT OF DETECTION (LLO)b.C Anotyeis water (pct /1)
Airborne PartiCutgte or Gases (pct /m ) F1sh MtIh (pC9/ho. wet) (oct/1) Food Producta Seetwent (DC9/kg. dry) (DC1/kS. met) groes bete 4 0.01 H-3 2000 ten-54 15 130 Fe-59 30 260 Co-58.60 15 130 In-65 30 260 Zr-ND-95 15 I-131 id 0.07 1 60 Co-134 15 0.05 130 15 80 150 Co-137 18 0.06 150 18 80 100 De-La-140 15 15 . E' O
Euclear Production - F;rai 2 ODCM-3 0 I Offcite Dose Calculation Manual Revicion 8 PagJ 3 0-44 I' TABLE 4.12.1-1 (Continued) TABLE NOTATIONS J aThis list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental i Operating Report pursuant to control 5.9.1 7. ' bRequired detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13.. - oThe LLD is defined, for purposes of these Controls, as the smallest concentration of radioactive material in a sample that will yield a not count, j above system background, that will be detected with 955 probability with only 55 probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which any include radiochemical separation:- 4.66 sb U E*V 2.22
- Y
- exp (- A t)
I Ilhare: ' LLD is the "a priori" lower limit of detection as defined above, as picoeuries per unit mass or volume, sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute. E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per pioocurie, Y is the fractional radiochemical yield, when applicable, {
.Aistheradioactivedecayconstantfortheparticularradionuclide,and i
t for environmental samples is the elapsed time between sample 1 collection, or end of the sample collection period, and time of counting
, Typical values of E, V, Y, and t should be used in the calculation.
i l
Nuclear Producticn - Fcrai 2 - ODCM-3 0 Offcite Dose Calculation Manual Revicien 8 Pag'a 3 0-45 TABLE 4.12.1-1 (Continued) TABLE NOTATIONS It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDe will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Control 5.9 1.7. , d LLD for drinking water samples. I t
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Nuclear Production - F rsi 2 ODCM-3 0 1
-Offaite Does Calculation Manual Revicion 8 Page 3 0-46 g -[ RADICIAGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAUD USE CENSUS s
deindS - . 3 12.2 A land use census shall be conducted and shall identify within'a distance of 8 km-(5 miles) the location in each of the 16 meteorological sectors#'of garden of the nearest greater than allk 50 mgnimal, (500 ft tp)producing nearest broad residence and the nearest leaf vegetation. >
' APPLICABILITY: At all times.
ACTION:
- a. With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Control 4.11.2 3, identify the new location (s) in ,
the next Annual Radioactive Effluent Release Report, pursuant to Control 5.9.1.8.
- b. With a land use census identifying a location (s) that yields a ,
calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being t, obtained in accordance with Control 3 12.1, add the new location (s) i to the radiological environmental monitoring program within 30 days. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Control 5.9 1.8, identify the new location (s) in the next Annual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting ~ the new location (s).
- c. The provisions of Control 3 0 3 and 3 0.4 are not applicable.
- SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best -
results, such as by a door-to-door survey, visual' survey, aerial survey, or by . consulting-local agriculture authorities. The results of the land use census I shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 5 9 1.7.
' Broad leaf vegetation sampling of at least three different kinds of vegetation p any be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Controls for broad leaf vegetation sampling in Table 312.1-1, Part 4.0, shall be followed, including analysis of control samples.
-~_
i Ihnolear Production . Fermi 2
~
ODCN-3 0 Off;ite Dose Calculation Manual, Revicion 8 . Paga 3 0-47 -! i -] . RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM f com0u 3 12 3 Analyses shall be performed on radioactiva materials supplied as part ! of an Interlaboratory Comparison Progras that has been approved by the - comminaion. ' APPLICABILYTY: At all times. 4 i
. ACTION:
- s. With analyses not being performed as required above, report the l corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to t Control 5 9.1.7.
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURTEILLANCE REQUIREMENTS i 4.12 3 The Interlaboratory Comparison Program shall be described in-the ODCN. A summary of the results obtained as part of the above required Interlaboratory i Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 5 9.1.7 i e
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._ ' Nuclear Production - Fermi 2 ODCM-4.0 Offsite Dose Calculation Manual Revicion 6 Page 4.0-1 k .
1 I l i l i 1 i SECTION 4.0
. 1 I.
l L iMf0RMAT10N OM ARMS - nurussiATION SERVICES I Date approved: 94/4'3 Release authorised by: [ h / Change numbers incorporated: 93-117-0DN '/ # DSN C4 /N - f. d Rev 6 Date MnV nasans 1 File 1715.02 , DTC TMPLAN Recipienf #1-7(I.IS l 'I i l
.- l
. Nuclear Production - F rai 2 ODCM-4.0 Offaite Dose Calculation Manual Revicicn 6 I
Pass 4.0-2 ( IESTRUMENTATION ,
.ASES l
3/4.3.7.11 RADI0 ACTIVE LIQUID EFFLUENT IONITORINGJISTRUMENTATION The radioactive liquid. effluent monitoring instrumentation is provided to aonitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alara/ trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCN to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4.3.7.12 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING ? .*RUMENTATION he radioactive gaseous affluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. ; he alara/ trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCN utilizing the systen design flow rates as specified in the ODCN. his conservative method is used because the Fermi 2 design does not include flow rate measurement g devices. his will ensure the alara/ trip will occur prior to exceeding the l limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is i consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION his control is provided to ensure that the concentration of radioactive l anterials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR Part 20.1001-20.2401, Appendix B. Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section ,
. II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC I l and (2) the limits of 10 CFR Part 20.1301 to a MEMBER OF THE PUBLIC. De concentration limit for dissolved or entrained noble gases is based upon the assumption that Ie-135 is the controlling radioisotope and its MPC in air (submersion) Mas converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP)
Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in teras of the lower limits of detection (LLDs).
~
( Detailed discussion of the LLD, and other detection limits can be found in HASL Procedure Manual, HASL-300 (revised annually), Currie, L. A., ' Limits for Qualitative Detection and Quantitative Determination - Application to l Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. E., " Detection , Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARN-C 215 (June 1975). i
. Nuclear Production - Fcrai 2 . ODCN-4.0 Offcite Dose Calculation Manual Revicicn 6 Page 4.0-3
( RADI0 ACTIVE DTLUENTS
.ASES 3/4.11.1.2 DOSE This control is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix 1, 10 T R Part 50. The control implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides ;
set forth in Section IV.A of Appendix I to assure that the releases of l radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking ; water supplies that can be potentially affected by plant operations, there is l reasonable assurance that the operation of the facility will not result in radicouclide concentrations in the finished drinking water that are in excess i of the requirement.s of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCN implement the requirements in Section III.A of Appendix 1 that conformance with the ' guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equatiora specified in the ODCN for calculating the Coses due to the actual release rates of radioactive materials in liquid affluents 1 are consistent with the methodology provided in Regulatory Guide 1.109, .
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents I for the Purpose of Evaluating Compliance with 10 GR Part 50, Appendix I,"
Revision 1, October 1977 and Regulatory Guide 1.113, "Estinating Aquatic I Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. 3/4.11.1.3 LIQUID RADWASTE TREA1 MENT SYSTEN i The OPERABILITY of the liquid radwaste treatment, system ensures that this system will be available for use whenever liquid effluents require treatment prior to their release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective g von in Section II.D of Appendix I to 10 WR Part 50. The specified limits governing the use of appropriate portions of the liquid
". radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 T R Part 50, for liquid effluents.
3/4.11.2 GA$EOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This control is provided to ensure that the dose to individual MEMBERS OF
- 4 1ME PUBLIC from gaseous effluents from all units on the sit.e will be within the I
limits of 10 W R Part 20.1301. l l
- _ . . \
..; Nuclear Production - Feral 2 . ODCH-4.0 Offcite Dose Calculation Manual Revision 6 Paga 4.0-4 RADI0 ACTIVE EFFLUENTS
.AsES 1/4.11.2.1 DOSE RATE (Continued) although this control applies to' the SITE BOUNDARY, the occupancy and exposure
, pathways applicable to a IEMBER OF THE PUBLIC who may at times' be within the SITE EDUNDARY vill usually.be such that such an individual will not1 receive' signifloantly greater dose due to gaseous effluents than a IEMBER OF THE PUBLIC uho remains outside the SITE BOUNDARY. Examples of calculations for such .
SEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given-in the ODCN. The specified dose rate limits restrict, at all times, the dose rates above background to a IEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 areas / year to the total body or to less than or , equal to.3000 aress/ year to the skin. These dose rate limits also restrict, at all times, the thyroid dose rates above background to a child via the inhalation pathway to less than or equal to 1500 aress/ year. i The required detection capabilities for radioactive materials in gaseous maste samples are tabulated in teras of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in NASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for ! Qualitative Detection and Quantitative Determination - Application to ( Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. E., " Detection Limits for Radionnalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.2.2 DOSE - N0BLE GASES This control is provided to implement the requirements of Sections II.B. III.A. and IV.A of Appendix I, 10 CFR Part 50. The control implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time laplement, the guides set forth in Section IV.A of Appendix I to assure that the releases of . radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a NENBER OF THE PUBLIC through , appropriate pathways is unlikely to be substantially underestimated. The dose i calculation methodology and parameters established in the ODCN for calculating I the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor i Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, 1 Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estinating Atmospherie Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCN : equations provided frr determining the air doses at and beyond the SITE ( BOUNDARY are based upon the historloal average atmospheric conditions. 1 l . _m __ .. . .__ ..
- Nuclear Production - Fcrai 2 '
ODCN-4.0 Offcite'Does Calculation Manual Revicion 6 Page 4.0-5 (' RADICACTIVE EFFLUENTS iAsEs 3/4.11.2.3 DOSE - 10 DINE-131. 10 DINE-133. TRITIUN. AND RADIONUCLIDES IN . PARTICULATE FORM This control is provided to implement the requirements of Sections II.C. III.A. and IV.A of Appendix I, 10 CPR Part 50. The controls are the guides set forth-
. in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the anse time laplement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive materials !
in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is ' reasonably achievable." The ODCN calculational methods specified in the ! Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCN calculational methodology and parameters for calculating the doses due to the actual release rates of the sub,)ect materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Complianos with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory guide 1.111, " Methods for Estinating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine I Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for iodine-131, iodine-133, trittua, and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were' examined in the development of these calculations were: (1) - individual inhalation of airborne radionuclides, (2) deposition of ! radionuclides onto green leafy vegetation with subsequent consumption by man, ; (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the i ground with subsequent exposure of man. t 3/4.11.2.4 0FF-GAS TREA1 MENT SYSTEM The OPERABILITY of the OFF-GAS TREATMENT SYSTEN ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. - The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the ; releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of General . Design Criteria 60 of Appendix A to 10 CPR Part 50, and the design ob,)ectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design ob,)ectives set forth in Sections II.B and t II.C of Appendix I,10 CFR Part 50, for gaseous effluents. 9
Nuclear Production - Fcrai 2 ODCH-4.0 Offaite Dose Calculatien Manual Revicien 6 Paga 4.0-6 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.5 VENTILATION EIHAUST TREATMENT SYSTEN The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous affluents will be kept "as low as is reasonably achievable." The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives met forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. , 3/4.11.2.8 VENTING OR PURGING This control provides reasonable assurance that releases from primary , containment purging operations will not exceed the annual dose limits of 10 TR Part 20 for UNRESTRICTED AREAS. 3/4.11.4 TOTAL DOSE This control is provided to meet the dose limitations of 40 W R Part 190 i ( that have been incorporated into 10 W R Part 20 by 46 FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive affluents and direct radiation exceed 25 area to the total body or any organ, except the thyroid,
'which shall be limited to less than or equal to 75 areas. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 WR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that abould result in the limitation of the annual dose to a member of the public to within the 40 W R Part 190 limits. For the purpose of the Special Report, it any be assumed that the dose commitment to the seaber of the public from other than uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the anse site or
- within a radius of 8 ka must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR Part 190, the special Report with a request for a variance (provided the release conditions resulting in violation of 40 TR Part 190 have not already been corrected), in accordance with the provisions of 40 TR Part 190.11 and 10 CFR Part 20.2203, is -
considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 WR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 TR Part 20, as addressed in Controls 3 11.1.1 and 3 11.2.1. An individual is not considered a seaber of the public during any period in which he/she is engaged in carrying out any operation that I is part of the nuclear fuel cycle. l
, Nuclear Production - Fcr:1 2 ,
ODCM-4.0 Offaite Dose Calculatien Manual Revicion 6
. Page 4.0-7
( RADI0th0ICAL ENVIRONMENTAL MONITORING sASES 3/4.12.1 NONITORING PROGRAM The radiological environmental monitoring progran required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplement.s the radiological effluent monitoring program by verifying that the measureable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent asamurements and the modeling of the environmental exposure pathways. Guidance for this monit.oring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in teras of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements i in industrial laboratories. It should be recognised that t,he 11D is defined as an a_ priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Detailed discussion of the LLD, and ot.her detection limits, can be found in RASL Procedure Manual, RASL-300 (revised annually), Currie, L. A., " Limits
. for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. E., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Banford Company Report ARH-SA-2_1} (June 1975).
3/4.12.2 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at ant beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from serial survey, from visual survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendig I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of ' leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: ( (1) 20% of the garden was used for growing broad leaf vegetatgon (i.e., slailar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m t
' ... l Nuclear Production - Feral 2 . ODCN-4.0 Offcite Dose Calculation Manual- Revicion 6 Pag 3 4.0-8
.( RADIOLOGICAL ENVIR019erfAL MONITORING
.Ases 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM
- The requirement for participation in an approved Interlaboratory
. Comparison Program is provided to ensure that independent checks on the- ! precision and securacy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance i program for environmental monitoring in order to demonstrate that the results 1 are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. END OF SECTION 4.0 l e
+
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+
6 l
Nuclear Production - Forsi 2 ODCN-5.0 Offsite Dose Calculation Manual Revicion 5 ,
,a.> 5.0-,
4 l SECTION 5.0 4 ADMINISTRATIVE CONTROLS t l 3 i l I INFORMATION ONI.Y i Aml5 - nurussi&TICE SERVICES Date approved: 9-af 43 Release authorised by: [ m Change numbers incorporated: DSN 04 dN -45d 93-117-ODN Rev 5 Date
/ ums g g DTC TMPLAN File 1715.02 Recipient -# 3 C p t
U..- 'Reclear Production - F:rai 2 . 0DCH-5.0 l Offcite Does calculation Manual Revicion 5 ; Pago 5.0-2 l ADMINISTRATIVE wmu-Lt AmrDAL RADIOLDGICAL ENVIRONMENTAL OPERATING REPORT i '5 9 1.7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar ' year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to , Nay 1 of the year following initial crittoality. !
.The annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison as appropriate, with preoperational studies, with operational controls, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses <
required by Control 3 12.2. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in'the Table and Figures in the ODCM, as well as summarised and tabulated results of these analyses and measurments in .
. the format of the table in the Radiological Ammaan= ant Branch Technical Position, Revision 1, November 1979. In the event that some individual results i are not available for inclusion with the report, the report shall be submitt,ed noting and explaining the reasons for the alssing results. If possible, the i I missing data shall be submitted as soon as possible in a supplementary report.. j The reports shall also include the following: a summary description of the radiological environmental monitoring progras; at least two legible maps
- l oovering all sampling locations keyed to a table giving distances and ,
l directions from the conter11ne of one reactor; the results of licensee ' participation in the Interlaboratory Comparison Program, required by Control l 312 3; discussion of all deviations from the sampling schedule of Table ; 3 12.1-1; and discussion of all analyses in which the LLD required by Table ; 4.12.1-1 uns not achievable. ! ANNUAL RADI0 ACTIVE EFFLUENT RPrFatR REPORTee f 5 9 1.8 Routine Annual Radioactive Effluent Release Reports covering the operatics of the unit during the previous year of operation shall be submitted within 90 days after January 1 of each year. The period of the first report shall begin with the date of initial criticality. i
. 80ne map shall cover stations near the SITE ROUNDARY; a second shall include the more distant stations. .
j
##A single submittal may be made for a mutiple unit station. The submittal
( should combine those sections that are common to all units at the station; however, for units with separate radunste systems, the submittal shall specify the releases of radioactive material from each unit. , 2 - - - .. . . . - . - - . - . .
, Nuclear Production - Fcrai 2 ODCH-5.0 Offcite Dose Calculation Manual .
.Revicien 5 Pass 5.0-3 ADMINISTRATIVE CONTROLS ANNUAL RADI0 ACTIVE EFFLUENT RELEASE RE?0RT (Continued)
The Annual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid unste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Mastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,' , Revision 1, June 1974, with data summarised on a quarterly basis following the format of Appendix B thereof. The Annual Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary any be either in the form of an, hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the fors of joint frequency distributions of wind speed, wind direction, and atmospheric stability.*** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their
~
activities inside the SITE BOUNDARY (Figure 3 0-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The assessment l of radiation doses shall be performed in accordance with t,he methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCN). The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed NEMBER OF THE PUBLIC from reactor releases and other nearby uranius fuel cycle sources, including doses from i primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection - Standards for Nuclear Power Operation. The assessment of radiation doses shall be performed in accordance with methodology and. parameters in the ODCH. The Annual Radioactive Efluent Release Reports shall include the following information for each class of sclid unste (as defined by 10 CFR Part 61) ahipped offsite during the report periods
- a. Container volume,
- b. Total ourle quantity (specify whether determined by measurement or estimate),
**'in lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required
, meteorological data on site in a file that shall ,be provided to the NRC upon request. ( l
. - - - . - - = _ . - . . . - . . - . - . -. - . .
Nuclear Production - F;ral 2 CDW-5.0 Offcite Does Calculation Manual . Revicion 5: Page 5.0-4 b ADMINISTRATIVE CONTROLS AinRIAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) l 'o. Principal radionuclides (specify whether determined by measurement or estimate), , 9
- d. Source of unste and processing employed (e.g., dewatered spent resin, compacted dry unste, evaporator bottoms), ;
- e. Type of container (e.g., LSA, Type A, Type 3, large Quantity), and '
, f. Solidification agent or absorbent (e.g., cement, urea formaldehyde). The Annual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site t,o UNRES*fRICTED AREAS of
- radioactive materials in gaseous and liquid effluents made during the reporting period.
The Annual Radioactive Effluent Release Report's shall include any changes made during the reporting period to the OFFSITE DOSE CALCULATION MANUAL (ODCM) as described in Technical Specification 6.14.2.c, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control 3 12.2. I The Annus1' Radioactive Effluent Release Reports shall also include the followings an explanation as to why the inoperability of liquid or gaseous affluent monitoring instrumentation was not corrected within the time specified in Control 3 3 7.11 or 3 3 7.12, respectively; and description of the event.s leading to liquid holdup tanks exceeding the limits of Technical Specifloation 3 11.1.4. 5.15 NAJOR CHANGES TO RADI0 ACTIVE LIQUID GASEOUS. AND SOLID WASTE TREATNENT Sh. z,ri5i 5.15.1 Licensee-initiated major changes to the radioactive unste systems J (liquid, gaseous, and solid):
- a. Shall be reported to the Commission in t,'he Annual Radioactive Effluent Release Report for the period in which t,he evaluation uns '
reviewed by the OSRO., The discussion of each change shall contains
- 1. A sunnary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.
- 2. Sufficient detailed information to totally support the reason ,
for the change without benefit of additional or supplemental information; I
' Licensees may choose to submit the information called for in this Control as part of the UFSAR revision in accordance with 10 CFR 50.71(e).
4
,.v -
- . . . . . - . . . . . , ., ..., , . , .n ,,.,
~ ~ ~
Nuclear Production F rsi 2 ODCN-5.0
. Off;1te Dose Calculation Manual Revicion 5 Pag 3 5.0-5
.(. ADMINISTRATIVE CONTROLS r
3 A detailed description of the equipment, components, and processes involved and the interfaces with other plant
, systems; ,
- 4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid umste that differ from those previously predicted in the license application and amendments thereto; 5 An evaluation of the change, which shows the expected maximum exposures to a IEMBER OF THE PUBLIC in the UNREST?.alED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;
- 6. A compariz i of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid unste, to the actual releases for the period prior to wher. the changes are.to be ande;
- 7. An estimate of the exposure to plant operating personnel as a g
result of the change; and
- 8. Documentation of the fact that the change was reviewed and found acceptable by the 05R0.
- b. Shall become effective upon review and acceptance by the OSRO.
9 END OF SECTION 5.0 t e e s
- 4 4
e l l i
% l
I- , Nuotoar Production - Fermi 2 ODCM-6.0 l . Offsite Dose Calculation M:nual P g] 6.0-1 l l- !
- 1. -
4 PART 11 CALCULATIONAL METHODS t NDP&lATION ONLY: ARMS-INFORMATION SYSTEMS OTC lbbY DSN Ob ~ N*O PAGE REV b PIS JAN 261994 I PIS RECIPIENT lC APPROVAL REO'D YES NO h i
ODCM-6.0 Revisi::n 6 Page 6.0-2 UQUID EFFLUENTS 4.0 LIQUID EFFLUENTS Thld section summarizes information on the liquid effluent radiation monitoring instrumentation and controls. More detailed information is provided in the Fermi 2 UFSAR : and Fermi 2 design drawings from which this summary was derived. This section also describes the sampling and analysis required by the Offsite Dose Calculation Manual. Methods for calculating alarm setpoints for the liquid effluent monitors are presented. Also, methods for evaluating doses from liquid effluents are provided. 6.1 Radiation Monitoring instrumentation and Controls This section summabzes the instrumentation and controls monitoring liquid effluents. This discussion focuses on the role of this equipment in assuring compliance with the Offsite Dose Calculation Manual. 6.1.1 Offsite Dose Calculation Manual (ODCM) 3.3.7.11 Requirement Fermi 2 ODCM 3.3.7.11 prescribes the monitoring required during liquid releases and the backup sampling required when monitors are inoperable. The liquid effluent monitoring instrumentation for controlling and monitoring radioactive effluents in accordance with the Fermi 2 ODCM 3.3.7.11 is summarized below:
- 1. Radiation Alarm - Automatic Release Termination
- a. Uquid Radweste Effluent une - The D11-N007 Radiation Monitor on the liquid redweste affluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 Maximum Permissible Concentration (MPC) at the discharge to Lake Erie (ten times 10 CFR 20, Appendix B, Table 2,
,/j g Column 2 values) required by ODCM 3.3.7.11. The monitor is located upstream of the isolation Valve (011-F733) on the liquid radweste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir (CWR) decent flow.
- 2. Radiation Alarm (only)
- a. Circulating Water Reservoir (CWR) Decant une - The CWR Decant i une Radiation Monitor (D11-N402) provides indication of the concentration of radioactive material in the diluted radioactive iiquid releases just before discharge to Lake Erie. As required by ODCM 3.3.7.11, the alarm setpoint is established to alarm (only) prior to exceeding one MPC.
ARMS - INFORMATION SERVICES Date approved: Release authorized by: Change numbers incorporated: g3-040-ODM DSN Rev 6 Date DTC TMPLAN File 1715.02 Recipient
- - ---~ _ _.
- = - - --
. ODCM-6.0 R;visi:n 6 Pig 3 6.0-3 Condensate Storage Tank (CST) Discharge Path - The CST s b. :
Discharge Monitor provides indication of the concentration of radioactive material in the CST Discharge Path prior to dilution by i qt[ , the circulating water decant flow. As required by ODCM 3.3.7.11, the alarm setpoint is established to alarm (only) prior to exceeding one MPC at the discharge to Lake Erie.
- 3. Plow Rate Measuring Devices s
- a. Liquid Radweste Effluent Line .In accordance with ODCM 3.3.7.11, the rolesse rate of liquid redweste discharges is monitored by
.' 011-R703. This flow rate instrumentation is located on the ,
radweste discharge line prior to the junction with the CWR decent line.
- b. Circulating Water Reservoir Decant Line - in &ccordance with ODCM 3.3.7.11, the flow rate of the CWR ' decent line is monitored -
,by N71-R802. The flow rate instrumentation is located on the decent line downstream of the junction with the liquid radweste
. effluent line. This instrumentation measures the total discharge flow rate from Fermi 2 to Lake Erie.
- c. Condensate Storage Tank Discharge Path - in accordance with -
ODCM 3.3.7.11, the flow rate of this line is monitored. g g[- , 6.1.2 Non-ODCM Required Monitor An additional monitor not required by Fermi 2 ODCM is provided by : Detroit Edison to reduce the likelihood of an unmonitored release of radioactive liquids.
- 1. General Service Water - The General Service Water (OSW) Radiation Monitor (D11-N008) provides additional control of potential radioactive cffluents. D11-N008 monitors the OSW System prior to discharge into !
the Main Condenser circuisting water discharge line to the Circulating Water Reservoir. Although not an ODCM required monitor, D11-N008 monitors a primary liquid stream in the plant that also discharges to , the environment (Lake Erie via the Circulating Water Reservoir). , Indication of radioactive material contamination in the OSW System would also indicate potential CWR contamination and the need to ; control all discharges from the CWR as radioactive , effluents. 6.2 Sampling and Analysis of Liquid Effluents The program for sarnpling and analysis of liquid waste is prescribed in the Fermi 2 Offsite Dose Calculation Manual Table 4.11.1.1.1-1. This table distinguishes two types of liquid releases': a) BATCH releases, defined as discrete volumes, from the Waste Sample Tanks (normally after processing through the radweste system) or from the ' Condensate Storage Tank, and b) CONTINUOUS releases, from the Circulating Water Reservoir (CWR) System, if it becomes contaminated Continuous releases from the CWR System are via the CWR decent line to Lake Erie. The CWR System is not expected to become contaminated. Therefore, continuous L radioactive material releases are not expected. However, the General Service Water i (OSW) and the CWR systems interface with radioactive systems In the plant. Also, L the OSW Intake is within a few hundred feet of the CWR decent line disch6rge to Lake Erie. For these reasons, it is prudent to consider the OSW and the CWR a potential source of radioactive effluents and to sample them regularly. N"7"T J::.*- . **~~'"'~"".~
R;visitn 6
. Page 6.0-4.
. 6.2.1' BATCH Releases:
Fermi 2 ODCM Table 4.11.1.1.1-1 requires that a sample representative of the tank contents be obtained before it is released. The table specifies the . following program: f
- Prior to sampling, the tank is isolated. The tank level is determined and this value is converted to tank volume. A pump with a known' .
recirculation flow rate is then activated to recirculate tank contents. The pump is allowed to run for at least the time required to i recirculate the tank volume twice.
- Prior to each batch release, analysis for principal gamma emitters (including all peaks identified by gamma spectroscopy) j
- Once per month, analysis of one batch sample for dissolved and entrained gases (gamma emitters). (See note in Section 6.2.2 below.) ;
- Once per month, analysis of a composite sample of all releases that month for tritium (H-3) and gross alpha activity. (The composite sample is required to be representative of the liquids released and sample quantities of the composite are to be proportional to the quantities of liquid discharged).
- Once per quarter, analysis of a composite sample of all releases that quarter for Strontium (Sr)-89, Sr-g0, and iron (Fe)-55.
6.2.2 CONTINUOUS Releases Fermi 2 ODCM Table 4.11.1.1.1-1 requires that composite samples be collected from the CWR System,if contaminated. The table specifies the
- following sample analysis- !
- Once per month, analysis of a composita sample for principal gamma
- l emitters and for 1-131.
- Once per month, analysis of a composite sample for H-3 and grou !
alpha.
- Once per month, analysis of weekly grab samples (composited) for dissolved and entrained gases (gamma emitters). (See note below.)
L . -
- Once per quarter, analysis for Sr-89,-go and Fe-55.
NOTE: Identification of noble gases that are principal gamma emitting radionuclides are included in the gamma spectral analysis performed on all i liquid redwesto effluents. Therefore, the ODCM Table 4.11.1.1.1-1 sampling
- ~
and analysis for noble gases in batch releases (one batch per month) and ' continuous releases (monthly analysis of weekly grab samples) need not be performed as a separate program. The gamma spectral analysis on each batch release and on the CWR monthly composite meets the intent of this ODCM requirement.
\
l
._l '"' ~~ **"~~ ~
- =. -. - - .. - . . - . -. ._ .
ODCM-6.0 i R:visi:n 6 1 Page 6.0-5 -l
. l 8.3 . Liquid Effluent Monitor Setpoints l Offsite Dose Calculation Manual 3.11.1.1 requires that the concentration of liquid '
radioactive effluents not exceed the unrestricted area MPC at the discharge point to Lake Erie. Dissolved or entrained noble gases in liquid effluents are limited to a concentration of 2 E-04 uCl/ml, total noble gas activity.' ODCM 3.3.7.11 requires that radiation monitor setpoints be established to alarm prior to exceeding the limits of l V'l Y , ODCM 3.11.1.1. To meet this specification, the alarm setpoints for liquid effluent monitors are determined in accordance with the following equation: ; l SP.< '! nn , (6-1) where: , l SP = the setpoint, in uCl/ml, of the monitor measuring the radioactivity ; concentration in the effluent line prior to dilution. The setpoint represents a value which, if exceeded, would result in concentrations exceeding the MPC in the unrestricted area i l CL = the effluent concentration limit (ODCM 3.11.1.1) corresponding to ten times the limits of 10 CFR Part 20.1302.b.2.1 at the discharge point in uCl/ml, defined in Equation (6-4) RR = the liquid effluent release rate as measured at the radiation monitor location, in volume per unit time, but in the same units as DF, below , DF = the dilution water flow as measured prior to the release point (Lake I Erie) in volume per unit time l At Fermi 2 the'available Dilution Water Flow (DF) is constant for a given release, and the waste tank Release Rate (RR) and monitor Setpoint (SP) are set to meet the condition of Equation (6-1) for a given effluent Concentration Umit, CL NOTE: If no dilution is provided, SP < CL Also, when DF is large compared to RR, then (DF + RR), ,DF, and DF may be used instead of (DF + RR) as a 1 simplification, as in Equation (6-5). 6.3.1 Tank Effluent Line Monitors j I/9Y i ! The Uquid Radweste E diuent une Monitor D11-N007 provides alarm and automatic termination of releases prior to excesding MPC. . The Condensate Storage Tank Distcharge Monitor provides alarm only prior to exceeding N l MPC. As required by ODCM Table 4.11.1.1.1-1 and as discussed in I
~
ODCM Section 6.2.1, a sample of the liquid radweste to be dischargow is .l
' collected and analyzed by gamma spectroscopy to identify principal gamma '
emitting radionuclides. From the measured individual radionuclide concentrations, the allowable rolesse rate is determined. The allowable release rate is inversely proportional to the ratio of the radionuclide concentrations to the MPC values. The ratio of the measured > concentration to MPC values is referred to as the *MPC fraction" and is calculated by the equation:
- . . - . . . +e.- **~~r-
ODCM-6.0
- Revisian 6 _ ,
L Page 6.0-6 MPCF = (6-2)
. where: .
' c MPCF = fraction of the unrestrii:ted area MPC for a mixture of gamma emitting radionuclides Ci = concentration of each gamma emitting radionuclide i measured In each tank prior to release (uCl/ml)
MPCg = unrestricted area most restrictive MPC for each radionuclide 1: ten times the value from 10 CFR Part 20, Appendix 8. Table 2 Column 2. For dissolved and entrained noble gases an MPC value of 2E-04 uCl/ml may be used, but noble gases need not be included in this calculation. including noble gases in Equation (6-2) eliminates the need for a separate evaluation of compliance with the noble gas concentration limit of ODCM 3.11.1.1. Based on the MPCF, the maximum allowable release rate can be calculated by the following equation: ; I I I (MPCF x (1+BF ) + HSMPCF I (6-3) where: MAX RR = maximum acceptable waste tank discharge rate (gal / min) (Monitor #G11-R703) DF = dilution flow rate from the CWR as observed from the Control Room readout (gal / min) (Monitor #N71-R802)
~
SF = administrative safety factor to account for variations in monitor response and flow rates. A SF value of 0.5 is suggested because it provides for 100% variation caused by statistical fluctuation and/or errors in measurements. . BF = conservative estimate of the ratio of the MPC fraction of pure beta emitters other than tritium to the gamma MPC fraction (MPCF)(The value 0.10 may be used for BF.) MPCF = As previously defined by equation (6-2) H3MPCF = conservative estimate of MPC fraction due to tritium (The value 0.13 m.ay be used for H3MPC.F.) NOTE: Equation (6-3) is valid only for MPCF >1: if the MPCF $1, the waste tank i concentration meets the limits of 10 CFR Part 20 without dilution, and the i I tank may be discharged at the maximum rate. i
. i.
e , ODCM-6.0 ' R:visi n 6 Page 6.0-7 ff MAX RR as calculated above is greater than the maximum discharge-- i y pump capacity lthe pump capacity should be used in establishing the actual
\1 Release Rate RR for the radweste discharge. For a Waste Sample Tanic the maximum discharge rate is 50 gallons per minute. This Release Rate RR is l fpy '
monitored in the Radweste Control Room by 011-R703. The Concentration Limit (CL) of a liquid radweste discharge is the asmo as the effective MPC for the radionuclide mixture of the discharge.' Simply, the CL (or effective MPC) represents the equivalent MPC value for a mixture of - , radionuclides evaluated collectively. The equation for determining CL is: CL a (6-4) I i Based on the Release Rate RR and Dilution Flow DF and by substituting -! Equation (6-4) for CL in Equation (6-1) and introducing sensitivity factorb l and factors to account for the presence of pure beta emitters, the alarm ' - l setpoint is calculated by the equation: C,
- SEN )
- W
- HW
- SF SP < '
1
, gg9 MPCF e (1+BF)
- RR - ;
(6-5) ! where:
- SP = setpoint of the radiation monitor counts per second (cps) or- //9y.
counts por minute (cpm) > l Ci - = concentration of radionuclide i as measured by gamma spectroscopy (uCl/ml) ' SENJ = monitor sensitivity for radionuclide i based on calibration i curve (cps /(uCl/ml) or cpm /(uCi/ml)) or single conservative igy ' value for all radionuclides (see below) RR = actual release rate of the liquid redweste discharge (gal / min) BF = pure beta factor as defined for Equation (6-3) ' MPCF = MPC fraction as determined by Equation (6-2) H3F = correction factor to account for estimated tritium concentration ' i et the discharge point (The value 0.gg may be used.)- Skg = background reading of monitor (cps) , DF = dilution flow rate of Circulating Water Decent Line as observed from Control Room readout (gal / min) monitor iPN71-R802. See ' note preceding Section 6.3.1. r SF = 1.0 when a single conservative sensitivity value is used; 0.5 - when individual nuclide sensitivity factors are used i f The sensitivity of Cr-51 determined from the primary calibration sensitivity l curves may be used as a single conservative value for SENg above. The Cr-51 sensitivity has been determined to be conservative based on the : nuclide mixes which have been seen in actual liquid discharges from ! a . - . , . , , - , +e-r -
./ . R: vision 6 Page 6.0-8 j a
Fermi 2. For the D11-N007 ' monitor, a monitor sensitivity value of 1.0 E6 -) cps /(uCl/ml) may be used as the single conservative value of SENg, and 'i i
.I. for the CST Discharge monitor, a single value of 1.6 E7 cpm /(uCi/ml) may ;
be used. if no radionuclides are measured by gamma spectroscopy, the alarm '
. setpoint can be established at one half the setpoint of the most recent discharge for which radionuclides were detected by gamma spectroscopy. ;
Prior to conducting any batch liquid radwaste release, Equation (6-3) is used ; to determine the allowable release rate in acco;1ance with ODCM 3.11.1.1. . Equation (6-5) is used to determine the alarm sepoint in accordance with ODCM 3.3.7.11. 6.3.2 Circulating Water Reservoir Decent Line Radiation Monitor (011-N402) ODCM 3.3.7.11 requires that the setpoint for the CWR Decant Line Radiation -l Monitor D11-N402 he established to ensure the radioactive material concentration in the decant line prior to discharge to Lake Erie does not - exceed MPC. unrestricted area (ten times 10 CFR 20, Appendix B, Table 2, i Column ? lues). The approach for determining the alarm setpoint for the CWR Decent Line Radiation Monitor is the same as presented in ' Section 6.3.1. However, the CWR Decent Line Radiation Monitor setpoint need not be changed prior to each release. Equation (6-1) remains valid, except that, for the CWR Decant Line Monitor, the dilution flow previously assumed for diluting the BATCH liquid radweste affluents is now the islease rate. There is no additional dilution prior to discharge to Lake Erie. Thus, ,
-( Equation (6-1) simplifies to:
SP _< CL (6-6) Substituting Equation (6-4) for CL and introducing a safety factor, sensitivity : factors, and monitor background, the D11-N402 alarm setpoint can be calculated by the equation: ; SP <- MPCF
+ Bkg IN ;
(6-7) ! where: . SP = setpoint in counts per minute (cpm) ; Cg = concentration of each radionuclide i in the CWR decent line l effluent uCi/ml) , SENi = monitor sensitivity for nuclide i based on calibration curve (cpm /(uCl/rni)) ; MPCF = MPC fraction as determined by Equation (6-2) with' Cg defined as for Equation (6-7) t/8)Y I SF = 0.5, administrative safety factor
- l. !
! Bkg = background reading of monitor (cpm)
~
l l
. ,, . L -
- ___ .._ _ _ _ ._. - _ ~ .
CDCM-6.0 Revisiin 6
' Page 6.0-g l7 . ,
Normally, only during periods of batch liquid redweste discharges will there L exist any plant-related radioactive material in the CWR decant line. 633 Generic, Conservative Alarm Setpoint for D11-N402 The D11-N402 setpoint could be adjusted for each BATCH release 'as is alone for.the. liquid redweste effluent line monitor. 8ased on the measured levels of radioactive material in a BATCH liquid release, the slann 3etpoint - for D11-N402 could be calculated using Equation (6-7). However, during these planned reioses, the concentrations will almost always be so low .
- (due to dilution) that the D11-N402 Monitor will not indicate measurable levels. The CWR decant line design flow is 10,000 ppm; and the maximum '
liquid redweste release rate is 50 spm, providing a 200:1 dilution. The
~
radioactive mate the range of 10-gal to 10- uCl/ml. conegntration With a nominal of BATCH liquid. releases is typicall 200:1 dilution (actual dilution has been greater since in actual releases the decent line , flow rate has been about 18,000 gpm), the ggecant linegonitor would monitor diluted activity in the range of 5 x 10 to 5 x 10 uCi/mL D11-N402 Monitor response at these levels would be 0.1 to 100 cpm, ; depending on the particular radionuclide mixture and corresponding l instrument response. These response levels are less than the monitor r background levels. In lieu of routinely adjusting the D11-N402 setpoints, generic; conservative ; I setpoints have been established based on an analysis of nuclides seen in actual liquid discharges and on the primary calibration sensitivity curve. ( 63.4 Alarm Setpoint for GSW and RHR System Radiation Monitors i Levels of radioactive material detectable above background at Radiation l Monitor D11-N008 would be one of the first indicators of contamination of ! the General Service Water (GSW) System and the CWR. Like. wise, for the Residual Heat Removal (RHR) System, the D11-N401 A and B Monitors would be one of the first indicators of contamination and subsequent , contamination of the CWR. Therefore, to provide early indication and assure l prompt attention, the alarm setpoints for these monitors should be established as close to background as possible without incurring a spurious , alarm due to background fluctuations. This level is typically around three times background. , if the OSW System or RHR System becomes contaminated, it may'become i necessary to raise the radiation monitor setpoints. The alarm setpoints should be re-evaluated to provide the CR operator a timely indication of . further increasing activity levels in the OSW or RHR System without ! spurious alarms. The method for this re-evaluation is the same as , ; described above - the alarm setpoint established at three times its current i reading. No regulatory limits apply for establishing a maximum value for l
. these alarm setpoints since these monitors are located on plant systems !
and do not monitor final release points to the environment. However, as a i practical matter, upper limits on the alarm setpoints can be evaluated using l the methods of ODCM Section 63.1 based on the actual system flows, ! i dilution and release paths in effect at the time. j
+ - _ _ - _ _ _ - _ - _ - - - - _ _ _ _ _ _ _ _ - _v - mm-- --- -
r- * * - 7 M
ODCM-6.0
- R;visi:n 6 -
Page 6.0-10 i I 6.3.5 Alarm Response - Evaluating Actual Release Conditions , i Normally, liquid release rates are controlled and alarm setpoints are established to ensure that the release does not exceed the concentration ; limits of ODCM 3.11.1.1 at the discharge to Lake Erie. However, if either ' Monitor D11-N007 or D11-N402 or the CST Discharge Monitor alarms lYW during a liquid release, it becomes necessary to re-evaluate the release conditions to determine compliance with ODCM 3.11.1.1. Following an alarm, the actual release conditions should be determined. Radioactive ' material concentrations should be evaluated by sampling the effluent stream ! or resampling the waste tank. Discharge flow and dilution water flow : should be redetermined. To perform this evaluation, the following equation may be used for all i nuclides, or dissolved and entrained noble gases may be evaluated ; separately from other nuclides using this equation: , C. . RR . (1+BF) <g CMPCg/) DF + RR H3F (6-8) where: Cg a measured concentration of radionuclide iin the effluent stream (uCl/ml) MPCg = the MPC value for radionuclide 1: ten times the 10 CFR 20, Appendix B, Table 2, Column 2 value (uCl/ml); 2 E-04 uCl/mi for dissolved or entrained noble gases RR = actual release rate of the liquid effluent at the time of the alarm, gpm DF = actual dilution circulating water flow at the time of the release *
, alarm, gpm A
H3F,BF = ! as previously defined NOTE: For alarm on D11-N402 (CWR decent line), the Release Rate RR is the : Dilution Water Flow DF and the DF term drops out of the equation. l f/'7V 6.3.6 Liquid Radwaste Monitor Setpoint Determination with Contaminated Circulating Water Reservoir J In the event the CWR is determined to contain radioactive material, the effective dilution capacity of the CWR is reduced as a function of the
- MPCF. To determine the available dilution flow capacity the MPCF for the CWR is determined using equation (6-2). The MPCF of the CWR is used to determine the available dilution flow as follows:
CWR Dilution Flow = CWR Decant Flow Rate (OPM) * (1-CWR MPCF) (6-9) l
ODCM-6.'0 ,
* - - R;visinn 6 ' Pago 6.0-11 I
The resulting' dilution flow rate is substituted in equation (6-3) to determine
- f. the maximum allowable release rate for discharges from the radweste
( system. Substituting the available CWR dilution flow from equation (6-9), the Liquid Radweste Monitor maximum release rate can be determined using i equation (6-3). 9 Once th's available dilution' flow and maximum allowable release rate have been determined the redweste monitor setpoint can be determined using , equation (6-5). 6.4 Contaminated GSW or RHR System - Quantifying and Controlling Releases , The GSW Radiation Monit'or (011-N008) provides an indication of contamination of
- this system. The Monitors D11-N401 A and B perform this function for the RHR System. Also, the CWR Decant Line Radiation Monitor monitors all liquid ,
releases from the plant and would record any release to Lats Erie from either of these systems if contaminated. . As discussed in ODCM Section 6.2.2, sampling and analysis of the CWR System is required only if this system is contaminated, as would. 6 be indicated by D11-N402 or D11-N008. Nonetheless, periodic samples are collected l from the CWR System to verify absence of contaminat. ion. Although not required by the ODCM, periodic sampling and analysis of the RHR System is also performed since it also is a potential source of contamination of the CWR and subsequent releases to Lake Erie. If contamination is found, further releases from the applicable system (GSW or RHR) via the CWR decant line must be evaluated and controlled to ensure that releases are maintained ALARA.- The following actions will be considered for- i controlling releases. I
- Sampling frequency of the applicable source (GSW or RHR System) and the !
CWR will be increased until the source of the contamination is found and
- controlled. This frequency may be relaxed after the source of contamination l j has been identified and isolated. ;
Gamma spectral analysis will be performed on each sample.
- The measured radionuclide concentrations from the gamma spectral analysis will be compared with MPC (Equation 6-2) to ensure releases are within the limits of ODCM 3.11.1.1.
- Based on the measured concentrations, the setpoint for the CWR Decant I Line Radiation Monitor (D11-N402) will be determined as specified in l Section 6.3.2. If the calculated setpoint based on the measured distribution is greater than the current setpoint (see ODCM Section 6.3.3) no adjustment L to the setpoint is required.
1
- Samples will be composited in accordance with ODCM Table 4.11.1.1.1-1 for I monthly analysis for H-3 and gross alpha and for quarterly analysis for Sr-89, 90 and Fe-55.
Each sample will be considered representative of the releases that have occurred since the previous sample. For each sample (and corresponding release period), the volume of liquid relassed to the lake will be determined based on the measured CWR decent line cumulative flow. f
- )
*uw-' -
w-++mt s m , _____ a +m- __._, _ _ __ _ , _ _ _ _ , _ _ , _ _ , _ _ _ _ _ _ _ _ , _ - _ _ _ , . ,_ _ , _ _ _ _ _ _ _ ____ _ _ _ . _ __ , _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . _ . _ . _ _ _ _ _ _ _ _ _
R:visi::n 6 ' Pags 6.0-12 , f
-- From the sample analysis and the calculated volume released, the total t radioactive material released will be determined and considered -
representative of the release period. Cumulative doses will be determined . In accordance with ODCM Section 6.5. ; 6.5 Liquid Effluent Dose Calculation - 10 CFR 50 I The parameters of the liquid release (or estimated parameters, for a pre-release calculation) may be used to calculate the potential dose to the public from the , release (or planned release). The dose calculation provides a conservative method for estimating the impact of radioactive affluents released by Fermi 2 and for comparing that impact against limits set by the NRC in the Fermi 2 ODCM, . The limits in the ; Fermi 2 ODCM are specified as quarterly and calender year limits. This assures that the average over the year is kept as low as reasonably achievable. l 6.5.1 MEMBER OF THE PUBLIC Dose - Liquid Effluents ODCM 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from Fermi 2 to: during any calender quarter;
$ 1.5 mrom to total body 5 5.0 mrom to any organ during any calendar year;
$ 3.0 mrom to total body
( $ 10.0 mrom to any organ ODCM 4.11.1.2 requires that quarterly and annual cumulative dose due to liquid effluents be determined at least once per 31 days. The calculation of the potential doses to MEMBERS OF THE PUBLIC is a function of the
- radioactive material releases to the lake, the subsequent transport and dilution in the exposure pathways, and the resultant individual uptake. At ,
Fermi 2, pre-operational evaluation of radiation exposure pathways indicated that doses from consumption of fish from Lake Erie provided the most I conservative estimate of doses from releases of radioactive liquids. l However, with the proximity of the water intake for the City of Monroe, it i must be assumed that individuals will consume drinking water as well as fish that might contain radioactivity from discharges into Lake Erie. Study of the currents in Lake Erie indicates that the current in the Lagoona Beach embayment carries liquid effluents from Fermi 2 north along the coast part of the time and south along the coast part of the time. When the current flows north, liquid effluents are carried away from the Monroe Water intake, so only the fish consumption exposure pathway must be considered. When the current flows south, toward the Monroe Water intake, both fish consumption and drinking water consumption exposure pathways must be considered. To ensure conservatism in the dose modeling, the combined fish and drinking water pathway is used for- I evaluating the maximum hypothetical dose to a MEMBER OF THE PUBLIC from liquid radioactive effluents. The following calculational methods may be used for determining the dose or dose commitment due to the liquid I' - radioactive effluents from Fermi 2:
. s. -,"m-y ,- ., ,, _
^
VeWRJ9 R;visi:n 6 Page 6.0-13 1 D, = 1.67 E-02
- VOL . .[(C, . Ag)
'4 DF
- 2 (6-10) 1 where: 1 Do
= dose or dose commitment to organ o or total body (mrom) due to release of a single tank Aoi
= site-specific ingestion dose commitment factor to the total body or any organ o'for radionuclide I (mrom/hr per uCi/mi)-
C; -= concentration of radionuclide i in undiluted liquid effluent - representative of the volume VOL (uCi/ml) VOL = total volume of liquid effluent released (gal) DF -= average dilution water flow (CWR decant line) during tank release (gal / min) Z = 5, near field dilution factor (Derived from Regulatory Guide 1.10g, Rev 0) 1.67 E-02 = 1 hr/60 min i The site-specific ingestion dose / dose commitment factors (Ajo)' represents a composite dose factor for the fish and drinking water pathway. The site-specific dose factor is based on the NRC's generic maximum individual consumption rates. Values of A i o are presented in i Table 6-1. They were derived in accordance with guidance of NUREG-0133
'from the following equation:
Ago = 1.14 E + 05 [(Uw / Dw) + (Up
- BFg)] DFj (6-11)
~
where: . Up = 21 kg/yr adult fish consumption 1 Uw = 730 liters /yr adult water consumption Dw = 15.4. additional dilution from the near field to the water intake for the City of Monroe (Net dilution factor of 77 from discharge point to drinking water intake, + Fermi 2 UFSAR, Chapter 11, Table 11.2-11) BFl = Bioaccumulation factor for radionuclide i in fish from Table 6-2 (pCl/kg per pCl/ liter) DFg = dose conversion factor for nuclide i for adults in organ o g from Table E-11 of Regulatory Guide 1.10g (mrom/pCI) 1.14 E + 05 = 106 (pCi/uCI) .103 (ml/kg) 8760 (hr/yr) r e -
4 y ; ODCM-6.0
~~'
R:visi:n 6
- P ge 6.0-14
< The radionuclides included in the periodic dose assessment required by.
- (. : ODCM 3.11.1.2 are those identified by gamma spectral analysis of the liquid ;
waste samples collected and analyzed per the requirements of OCCM Table 4.11.1.1.1-1. In keeping with the NUREG-0133 guidance, the .; adult age group represents the maximum exposed individual age group. ! Evaluation of doses for other age groups is not re' quired for demonstrating-compliance with the dose criteria of ODCM 3.11.1.2. The dose analysis for radionuclides requiring radiochemical analysis will be performed after receipt of results of the analysis of the composite samples. in keeping with~
, the required analytical frequencies of ODCM Table 4.11.1.1.1-1, tritium dose analyses will be performed at least monthly; Sr-89, Sr-go and Fe-55 dose ;
analyses will be performed at least quarterly. 6.5.2 Simplified Liquid Effluent Dose C,'culation In lieu of the individual radionuclide dose assessment presented in . ' Section 6.5.1, the following simplified d'ose calculation may be used for - d6monstrating compliance with the dose limits of ODCM 3.11.1.2. (Refer to . Appendix A for the derivation of this simplified method.) Equations (6-12) and (6-13) are to be applied tn the release of a single tank. l Total Body , j Dg = 9.69 E + 03 . VOL , [C, ' DFeZ i (6-12) Maximum Organ D,,, = 1.18 E + 04 . VOL , [C y DF e 2 (6-13) , where:
= concentration of radionuclide i in undiluted liquid CJ
- effluent representative of the volume VOL (uCl/ml)
VOL = volume of undiluted liquid effluent released (gal) - DF = average dilution water flow (CWR decant line) during - tank release (gal / min) l- Z = 5 near field dilution factor (derived from Regulatory Guide 1.109. Rev 0) Dtb
= conservatively evaluated total body dose (mrom)
Dmax
= conservatively evaluated maximum organ dose (mrom) l .
9.69 E + 03 = 0.0167 (hr/ min)
- 5.80 E + 05 (mrom/hr per uCl/ml, Cs-134 total body dose factor from Table 6.0-1)
I' . i l- 1.18 E + 04 = 0.0167 (hr/ min)
- 7.09 E + 05 (mrom/hr per uCi/mi, l
Cs-134 liver dose factor from Table 6.0-1) , l 4 l l
m = ODCM-6.0 l R;vis12n 6 I Page 6.0-15 6.5.3 Contaminated CWR System - Dose Calculation
- (E If the CWR System becomes contaminated, releases via the CWR System to Lake Erie must be included in the evaluation of the cumulative dose to a MEMBER OF THE PUBLIC as required by ODCM 3.11.1.2. ODCM Section 6.4 described the methods for quantifying and controlling releases from the CWR System.
For calculating the dose to a MEMBER OF THE PUBLIC, Equation (6-10) remains applicable for releases from the OSW System with the following assumptions:
* - DF, Dilution Flow,is set equal to the average CWR decent line flow rate over the release period.
Cg, Radionuclide Concentration,is determined as specified in ODCM l Section 6.4.
- VOL, Volume Released, is set equal to the total volume of the discharges to Lake Erie via the CWR decent line as specified in Section 6.4. ;
6.6 Liquid Effluent Dose Projections ; F 10 CFR 50.36a requires licensees to maintain and operate the Radwaste System to ensure releases are maintained ALARA. This requirement is implemented through i ODCM 3.11.1.3.' This section requires that the Liquid Radioactive Waste Processing ! System be used to reduce the radioactive material levels in the liquid waste prior to release when the projected dose in any 31 day period would exceed:
- 0.06 mrom to the total body, or_
- 0.2 mrom to any organ When the projected doses exceed either of the above limits, the waste must be
. processed by the Liquid Radwaste System prior to release. This dose criteria for processing is established at one forty eighth of the design objective rate (3 mrom/yr, total body or 10 mrom/yr any organ) in any 31 day period.
The applicable Liquid Weste Processing System for maintaining radioactive material releases ALARA is the Mixed Bed Domineralizers as delineated in Figure 6-1. Alternately, the Waste Evaporator (presented in the Fermi 2 UFSAR, Section 11.2) can be used to meet the NRC ALARA design requirements. it may be used in conjunction with or in lieu of the Mixed Bed Domineralizers to meet the waste processing , requirements of ODCM 3.11.1.3. Each BATCH release of liquid redweste is evaluated to' ensure that cumulative doses are maintained ALARA. In keeping with the requirements of ODCM 3.11.1.3, dose projections are made at least once per 31 days to evaluate the need for additional radweste processing to ensure future releases are maintained ALARA. The following equations may be used for the dose projection calculation: I b o i 6
' ' 'ODCM-6.0 R;visi:n 6 Page 6.0-16 Othp = Otb (31/ d)
(5-14)
- Dmaxp = Dmax (31/ d)
(6 *: 5) ' ; where: ! Dtbp " the total body dose projection for the next 31 day period (mrom) ! NOTE: The reference calendar quarter is normally the current calendar quartere - if there have been liquid releases in the previous quarter but not in the 4 current quarter, the previous quarter should be used as the reference calendar quarter.
=
Dtb the cumulative total body dose for all releases to date in the reference calendar quarter (normally the current quarter) as determined by equation (6-10) or (6-12) (mram)
=
Dmaxp the maximum organ dose projection for the next 31 day period (mrom) Dmax = the cumulative maximum organ dose for all releases to date in the l reference calendar quarter as determined by Equation (6-10) or
, (6-13) (mrom) e d =
the number of days from the beginning of the reference calendar ; quarter to the date of the dose projection evaluation. i 31- = the number of days in projection ; i in the case of Condensate Storage Tank releases only, it may be possible to make an ' accurate dose projection without relying entirely on releases during the reference calendar f quarter, if the above equations are not used for the portion of the dose projection pertaining to CST releases, the method used shall be in accordance with the dose projection ' procedure. ! END OF SECTION 6.0 l I e
,,,g , en em e e64 6 9
-- v -r--- - - -
ODCM-6.0
. R3vislan 6 Pa9e 6.0-17 TABLE 6.0-1
. Fermi 2 Site Specific Liquid in9estion Dose Commitment Factors :
. Ago (mrom/hr per uCi/ml) i
' Nuotide Bone Liver T Body Thyroid Kidney Lun9- Ol-LLI {
H-3 - 7.94E-1 7.94E-1 7.94E-1 7.94E-1 7.94E-1 7.94E-1 , C 14 3.13E+4 ' 6.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 Na-24 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 P-32 1.39E+6 8.62E+4 5.36E+4 - - - 1.E6E+5 Cr-51 - .- 1.29E+0 7.70E-1 2.84E-1 1.71E+0 3.24E+2 Mn-54 - 4.40:~+ 3 540E+2 - 1.31E+3 - 1.35E+4 ; Mn-56 - 1.11 E+2 1.tSE+1 - 1.41E+2 - 3.53E+3 ; Fo-55 . 6.73E+2 4.65E+2 1.03E+2 , 2.59E+2 2.67E+2 , FO-59 1.06E+3 2.50E+3 9.57E+2 - - 6.98E+2 8.32E+3 , C2-57 - 2.19E+1 3.64E+1 - - - 5.55E+2 Cc-58 - 9.32E+1 2.09E+2 - - - 1.89E+3 Co-60 - 2.68E+2 5.90E+2 - - - 5.03E+3 Ni-63 3.18E+4 2.21E+3 1.07E+3 - - - 4.60E+2 Ni-65 1.29E+2 1.68E+ 1 7.66E+0 - - - 4.26E+2 Cu-64 - 1.04E+1 4.89E+0 - 2.63E+1 - 8.aaE+2 1-t5 2.32E+4 7.38E+4 3.34E+4 - 4.94E+4 - 4.65E+4 (
.n-69 4.94E+ 1 9.44E+1 6.57E+0 -
6.14E+1 - 1.42E+1 . Br-82 - - 2.28E+3 - - - 2.62E+3 Br-83 - - 4.06E+1 - - - 5.85E+1 . Br-84 - - 5.27E+1 - - - 4.13E Br-85 - - 2.16E+0 - - - 1.01 E-15 , Rb-86 - 1.01E+5 4.71E+4 - - - 1.99E+4 Rb-88 - 2.90E+2 1.54E+2 - - - 4.01 E-9 l Rb-89 - 1.92E+2 1.35E+2 - - - 1.12E-11 Sr-89 2.38E+4 - 6.83E+2 - - - 3.81E+3 - Sr-90 5.85E+5 - 1.44E+5 - - - 1.69E+4 l Sr-91 4.38E+2 - 1.77E+1 - - - 2.09E+3 l Sr-92 1.66E+2 - 7.18E+0 - - - 3.29E+3 : Y-90 6.28E-1 - 1.68E-2 - .- - 6.66E+3 1
- Y-91m 5.93E-3 -
2.30E-4 - - - 1.74E-2 ; Y-91 9.20E+0 - 2.46E-1 - - - 5.06E+3 Y-92 5.51E-2 - 1.61 E-3 - - - 9.66E+2 Y-93 1.75E-1 - 4.83E-3 - - - 5.55E+3 Zr-95 4.04E-1 1.30E-1 8.78E-2 - 2.04E-1 - 4.11 E+2 Zr-97 ' 2.24E-2 4.51E-3 2.06E-3 - 6.81E-3 - 1.40E+3 Nb-95 4.47E+2 2.49E+2 1.34E+2 - 2.46E+2 - 1.51E+6 Nb-97 3.75E+0 9.48E-1 3.46E-1 - 1.11E+0 - 3.50E+3 t ' A3-99 - 1.26E+2 2.41E+1 - 2.86E+2 - 2.93E+2 Tc-99m 1.02E-2 2.88E-2 3.67E-1 - 4.38E-1 1.41E-2 1.71E+1 Tc-101 1.05E-2 1.51F-2 1.48E-1 - 2.72E-1 7.73E-3 4.54E-14
.F
CDCM-6.0 R;visi:n 6 ; Pag 3 6.0-18 i TABLE 6.0-1 [ Fermi 2 Site Specific Liquid Ingestion Dose Commitment Factors ; Ago (mrom/hr per uCl/ml) Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI Ru-103 5.43E+0 - 234E+0 - 2.07E+1 - 6.34E+2 Ru-105 4.52E-1 - 1.78E-1 - 5.84E+0 - 2.76E+2 Ru-106 8.07E+1 - 1.02E+1 - 1.56E+2 - 5.22E+3 Rh-103m - - - - - - - Rh-106 - - - - - - - A9-110m 1.75E+0 1.61 E+0 9.59E-1 - 3.17E+0 - 6.59E+2 Sb-124 2.18E+1 4.13E 8.66E+0 5.29E-2 - 1.70E+1 6.20E+2
'Sb-125 1.40E+1 1.56E-1 3.32E+0 1.42E-2 -
1.08E+1 1.54E+2 Tc-125m 2.58E+3 9.35E+2 3.46E+2 7.76E+2 1.05E+4 - 1.03E+4 Tc-127m 6.52 E+3 233E+3 7.94E+2 1.67E+3 2.65E+4 - 2.19E+4 Tc-127 1.06E+2 3.80E+1 2.29E+ 1 7.85E+ 1 431E+2 - 8.36E+3 Tc-129m 1.11 E+4 4.13E+3 1.75E+3 3.80E+3 4.62E+4 - 5.58E+4 TO-129 3.02E+ 1 1.14E+ 1 7.37E+0 232E+1 1.27E+2 - 2.28E+ 1 .i Te-131m 1.67E+3 8.15E+2 6.79E+2 1.29E+3 8.25E+3 - 8.09E+4 Te-131 1.90E+ 1 7.93E+0 5.99E+0 1.56E+1 8.31 E+1 - 2.69E+0
'c-132 2.43E+3 1.57E+3 1.47E+3 1.73E+3 1.51 E+4 -
7.42E+4 I 130 3.12E+ 1 9.21E+ 1 3.64E+ 1 7.81 E+3 1.44E+2 - 7.93E+1 i 1-131 1.72E+2 2.46E+2 1.41 E+2 8.06E+4 4.21E+2 - 6.49E+ 1 i 1-132 8.39E+0 2.24E+ 1 7.85E+ 0 7.85E+2 3.57E+ 1 - 4.21E+0 1-133 5.87E+1 1.02E+2 3.11 E+1 1.50E+4 1.78E+2 - 9.17E+1 1-134 4.38E+0 1.19E+ 1 4.26E+0 2.06E+2 1.89E+1 - 1.04E-2 1-135 1.83E+1 4.79E+ 1 1.77E+1 3.16E+3 7.68E+1 - 5.41 E+1 Cs-134 2.98E+ 5 7.09E+ 5 5.80E+ 5 - 230E+5 7.62E+4 1.24 E+4 l Cs-136 3.12E+4 1.23E+5 8.87E+4 - 6.85E+4 9.40E+3 1.40E+4 Cs-137 3.82 E+5 5.22E+5 3.42E+5 - 1.77E* 5 5.90E+4 1.01 E+4 Cs-138 2.65E+2 5.22E+2 2.59E+2 - 3.84E+2 3.79E+1 2.23E-3 4.25E-2 9.68E-4 5.87E-4 2.58E+0 l Ba-139 1.45E+0 1.04E-3 - 1 Ba-140 3.04E+2 3.82E-1 1.99E+1 - 130E-1 2.19E-1 6.26E+2 i Ba-141 7.06E-1 5.33E-4 2.38E-2 - 4.96E-4 3.03E-4 3.33E-10 ; Ba-142 3.19E-1 3.28E-4 2.01E-2 - 2.77E-4 1.86E-4 4.49E-19 j La-140 1.63E-1 8.22E-2 2.17E-2 - - - 6.04E+3 La-142 835E-3 3.80E-3 9.46E-4 - - - 2.77E+1 Ce-141 7.30E-2 4.94E-2 5.60E-3 - 2.29E-2 - 1.89E+2 i Co-143 1.29E-2 9.51E+0 1.05E-3 - 4.19E-3 - 3.56E+2 i CO-144 3.81E+0 1.59E+0 2.04E-1 .- 9.44E-1 - 1.29E+3 Pr-143 6.00E-1 2.41 E-1 2.98E-2 - 1.39E-1 - 2.63E+3 Pr-144 1.96E-3 8.16E-4 9.98E-5 - 4.60E-4 - 2.83E-10 t Jd-147 4.10E-1 4.74E-1 2.84E-2 - 2.77E-1 - 2.28E+3 W-187 2.96E+2 2.48E+2 8.66E+1 - - - 8.12E+4 No-239 3.49E-2 3.43E-3 1.89E-3 - 1.07E-2 - 7.04E+2 s.
ODC2J-6.W Revisi n 6 Page 6.0-19 TABLE 6.0-2
\ Bioaccumulation Factors (BFi)
(pCl/kg per pCi/ liter)* Element Freshwater Fish H 9.0E-01 ~ C 4.6E+03 Na -1.0E+ 02 P 3.0E+03 Cr 2.0E+02 Mn 4.0E+02 Fe 1.0E+02 Co 5.0E+01
~
Ni 1.0E+02 Cu 5.0E+01 Zn 2.0E+03 Br 4.2E+02 Rb 2.0E+03 Sr 3.0E+01 Y 2.5E+01 Zr 3.3E+00 Nb 3.0E+04 Mo 1.0E+01
, Tc 1.5E+01 Ru 1.0E+01 Rh 1.0E+01 Ag 2.3E+00 Sb 1.0E+00 Te 4.0E+02 1 1.5E+01 Cs 2.0E+03 Ba 4.0E+00 La 2.5E+01 ,
Ce 1.0E+00 Pr 2.5E+01 Nd 2.5E+01 W 1.2E+03 Np 1.0E+01
- Values in this table are taken from Regulatory Guide 1.109 except for phosphorus, which is adapted from NUREG/CR-1336, and silver and antimony, which are taken from UCRL 50564, Rev 1, October 1972.
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Nuclear Production - F:rmi 2 ODCM-7.0 Offsite Dose Calculation Manual Rovi:lin 6 P g3 7.0-1 GASEOUS EFFLUENTS 7.0 GASEOUS EFFLUENTS 1 7.1 Radiation Monitoring instrumentation and Controls : 7.1.1 ' Effluent Monitoring - Ventilation System Releases The gaseous effluent monitoring instrumentation required et Fermi 2 for ! controlling and monitoring radioactive effluents are specified in ODCM 3.3.7.12. The monitoring of each identified gaseous offluent release point must include the following: Noble Oas Activity Monitor lodine Sampler (sample cartridge containing charcoal or aliver zeolite) ; Particulate Sampler (filter paper) Sampler Flow Rate Monitor Meeting these requirements, a total of seven Eberline SPING Monitoring Systems are installed on the six gaseous release points (Onsite Storage Facility, Service Building, Radweste Building, Turbine Building, Reactor Building Exhaust Pienum, and Standby Gas Treatment System Division 1 and Division 2). The SPING Monitor outputs are recorded
. electronically in the CT-2B Control Terminal in the Main Control Room.
In general, a reading exceeding the High alarm setpoint of the SPING Monitors causes an alarm in the Control Room. Fermi 2 ODCM Table 3.3.7.12-1 identifies these alarm functions. 7.1.2 Main Condenser Offgas Monitoring ODCM Table 3.3.7.12-1 and Technical Specification Table 3.3.7.12-1 specify monitoring requirements for the Offgas System at the 2.2 minute delay line. The following monitors are required: Hydrogen Monitor - used to ensure the hydrogen concentration in the Offgas Treatment System is maintained less than 4% by volume as required by Technical Specification 3.11.2.6. Noble Oas Activity Monitor - used to ensure the gross activity release rate is maintained within 340 millicuries per second after 30 minute' decay as required by Technical Specification 3.11.2.7. ARMS - INFORMATION SERVICES Date approved: f - 4L / - $5 Release authorized by: [O dd/d _ Change numbers incorporated: g3-117-ODM [ DSN 04 d K - 7, d Rev 6 Date NOV 041QQ1 . DTC TMPLAN File 1715.02 Recipient /d4M l
"i.*
- ODCM-7.0 '. I Revisi:n 6 i Page 7.0-2
. These two monitors perform safety functions. The Hydrogen Monitor g; monitors the potential explosive mixtures in the Offges System. The Noble i Gas Monitor monitors the release rate from the main condenser ensuring' i doses at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment bypassing the Offges Treatment System.
7.1.3 Reacter Building Ventitation Monitors (Gulf Atomic) The Gulf Atomic Monitors (011-N408 and 410) on the Reactor Building : Ventilation System provide on high radiation levels (above alarm setpoint) !
-initiation of SGTS, isolation of drywell vent / purge, isolation of the R8 and.
Control Center Ventilation Systems and initiation of Control Center , recirculation mode ventilation. These monitors and functions are not required by Fermi 2 ODCM but are important in controlling containment i venting / purging. 7.2 Sampling and Analysis of Gaseous Effluents The program for sampling and analysis of gaseous waste is prescribed in Fermi 2 ODCM Table 4.11.2.1.2-1. This table distinguishes two types of gaseous releases: (1) containment PURGE, treated as BATCH releases, and (2) discharges from the Reactor Building Exhaust Plenum (including Standby Gas Treatment System (SGTS) when operating), and other building ventilation exhausts, treated as CONTINUOUS releases. i 7.2.1 Containment PURGE
'I ODCM Table 4.11.2.1.2-1 requires that samples be collected and analyzed before each primary containment PURGE. Sampling and analysis is required
, within eight hours before starting a PURGE. ODCM Table 4.11.2.1.2-1 - Footnote j and ODCM 4.11.2.8.2 also require that if the purging or vnnting is through the Reactor Building ventilation, rather than through SGTS, and if the primary containment radiation monitoring system is INOPERABLE or in s' arm condition, sampling and analysis is required within 8 hours prior to ' and at least_ once per 12 hours during venting or purging of the primary containment. ' The required analyses must include princial gamma emitters and, if a pre-vent or pre-purge sample, tritium. ; For a planned containment PURGE, the results of the samples and analyses may be used to establish the acceptable release rate and radiation monitor alarm setpoint in accordance with ODCM Section 73. This evaluation may be necessary to ensure compliance with the dose rate limits of ODCM 3.11.2.1. In practice, release flow rates are fairly constant and these calculations are necessary only if a threshold value of nuclide concentration in the primary containment atmosphere is reached. The alarm setpoints of the primary containment atmosphere monitor, the Reactor Building
. ventilation exhaust monitors, and the Reactor Building and SGTS SPING monitors are set to ensure that release routes are continuously monitored i and controlled in accordance with 10 CFR 20 or limit: 3pecified in the l ODCM. '
i 7.2.2 Ventilation System Releases ODCM Table 4.11.2.1.2-1 requires continuous samples of releases from the RB Exhaust Plenum, Standby Gas Treatment System Radweste Building. ! Turbine Building, Service Building, and Onsite Storage Facillity. The table specif6*- 'he following program: i
-- . - . . . - . . - . ~. . ._ .
e ' I ODCM-7.0 ' Y - R:visi:n 6 P:ge 7.0-3
~
~
'Once ~per we Jk,' analysis of an adsorbent sample of I-131 and 1-133, -)
plus analysis of a particulate sample for principal gamma emitters. Once per month, analysis of a composite particulate sample of all l 1 releases (by release point) that month for gross alpha activity. l
. Once por quarter, analysis of a composite particulate sample of all )
releases that quarter for Sr-89 and Sr-90. Once per month, analysis of a grab sample for principal gamma omitters (noble gases and tritium)f j ODCM Table 4.11.2.1.2-1 also requires continuous monitoring for noble I gases.' This requirement is met by the SPING Monitors on each of the plant ! gaseous release points. The ODCM requires more frequent sampling and analysis following reactor = ) startup, shutdown, or change in thermal power exceeding 15% within one hour. The ODCM allows exceptions to this increased sampling schedule provided that neither one of the following conditions exist: Primary coolant dose equivalent I-131 has increased more than a ' factor of three. . 1 Reacter Building SPING noble gas monitor has increased more than a l factor of three. Grab samples of the Fuel Pool Ventilation Exhaust are required tritium (-' analysis once per seven days whenever spent fuel is in the Spent Fuel Pool. Also, grab samples for tritium are required when either the reactor well or the dryer separator pool is filled. These samples are taken at the Reactor Building Exhaust Pionum and Standby Gas Treatment System (SGTS) when operating. 7.3 Gaseous Effluent Monitor Sotpoint Determination 73.1 Ventilation System Monitors-Per the requirements of ODCM 3.3.7.12, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of ODCM 3.11.2.1. 4 This section limits releases to a dose rate at the SITE BOUNDARY of i 500 mrom/ year to the total body or 3000 mrom/ year to the skin. From a j grab sample analysis of the applicable release (i.e., grab sample of the ! primary containment or Ventilation System release), the radiation monitoring alarm setpoints may be established by the following calculational method. The measured radionuclide concentrations and release rate are used to calculate the fraction of the allowable release rate, limited by ODCM 3.11.2.1, by the equation: FRAC = 1.87 E + 01
- X/O
- VF * [%
- K ) ' '
500 (7-1) PRAC = 1.67 E + Ot
- X/O
- VF * [(C, * [L, + 1.1 MJ) 3000 '
(7-2)
l?' *
'R2visi:n 6 ,
!' Pago 7.0-4 1
;.' Where:
F
!gO FRAC = fraction of the allowable release rate based on the identified radionuclide concentrations and the release
' flow rate e X/O = annual average meterological dispersion to'the -
controlling site boydary location from Table 7.0-3 (sec/m ) or plant procedures l VF = Ventilation System flow rate for the applicable release l point and monitor (liters / minute) { i CJ = concentration of noble gas radionuclide i at release , point as determined by gamma spectral analysis of grab sample (uCl/cc) (if a noble gas is not detected at the reactor building release point,its concentration at this
. release point may be calculated by applying a dilution factor to its concentration in an Offges Vent Pipe sample.) '
J Kg i
= total body dose radionuclide conversion I (mrom/yr factorf,or per uCl/m noble7.0-2) from Table gas l L3 = beta skin dose conversion factor fgr noble gas radionuclide I (mrom/yr per uCl/m , from Table 7.0-2)
( Mg
= gamma a'r dose conversion factog for noble gas .
radionuclide I (mrad /yr per uCi/m , from Table 7.0-2) , 1.1 = mrom skin dose per mrad gamma air dose (mrom/ mrad) 500 = total body dose rate limit (mrom/yr) ; 3000 = skin dose rate limit (mrom/yr) 1.67 E + 01 = 1 E + 03 (cc/ilter) * (1/60) (min /sec) Based on the more limiting (i.e., higher) value of FRAC as determined above, the alarm setpoints for the applicable monitors may be calculated by the equation: SP g (AF * [ CJ + Skg FRAC (7-3) Where: 4 SP = alarm setpoint corresponding to the maximum allowable release rate (uCi/cc) Bkg = bac'kground of the monitor (uCi/cc)
^
AF = administrative allocation factor (Table 7.0-1) for the apacific monitor and type release, which corresponds to the fraction of the total allowable release rate that is administratively allocated to the individual release points. l i en s
ODCM-7.0
?: R visi:n 6 Page 7.0-5 .
L; Cg = concentratiori of Noble Gas Radiohuclide i as determined k by gamma spectral analysis of grab sample (uCl/cc) (if a - noble gas is not detected at the reactor building release point,its concentration at this release point may be calculated by applying a dilution factor to its concentration in an Offges Vent Pipe sample.) Note: . If the monitor channel in question was showing a response to the effluent at the time of the grab sample, this response minus background may be used in ileu of the summed grab sample concentrations. The Allocation Factor (AF) is an administrative control imposed to ensure ; that combined releases from all release points at Fermi 2 will not exceed : the regulatory limits on release rate from the site (i.e., the release rate limits of ODCM 3.11.2.1). From the Formi 2 design evaluation of gaseous affluents presented in the UFSAR Section 11.3, representative values have been determined for AF. These values are presented in Table 7.0-1. These . values may be . changed in the future as warranted by operational experience, provided the site releases comply with ODCM 3.11.2.1. In addition to the allocation factor, safety factors which have the effect of ! lowering the calculated setpoints may be applied. When determining the Noble Gas Monitor calibration constant, the monitor sensitivity for Xe-133 ' may be used in lieu of the sensitivity values for the individual l radionuclides. Because of its lower gamma energy and corresponding ; monitor response, the Xe-133 sensitivity provides a conservative value for ' alarm setpoint determination. Alternatively, if the monitor channel in ( question frequently shows a response to a mix of isotopes whose , concentrations can be determined, the calibration constant may be determined from this type of data without reference to primary calibration i data. : 7.3.2 Setpoint Determination with No Nuclides Detected When noble gas concentrations for a release point cannot be determined from grab samples, there are two options for setpoint determination. First,
- the setpoint may be set slightly above monitor background (e.g. 2 to 3 times background). This approach may be used when releases are not expected from a particular release point. Second, the equations of Section 7.3.1 may be used with noble gas concentration values based either on UFSAR tables or on values from a release point for which concentrations have been determined (e.g. reactor building exhaust plenum). When this method is used, a safety factor should be used in the setpoint calculation.
73.3 Gaseous Effluent Alarm Response - Evaluating Actual Release Conditions The monitor alarm solpoint is used as the primary method for ensuring and demonstrating compliance with the release rate limits of ODCM 3.11.2.1. Not exceeding alarm setpoints constitutes a demonstration that release i rates have been maintained within the ODCM limits. When an effluent , Noble Gas Monitor exceeds the alarm setpoint, an evaluation of compliance
. with the release rate limits must be performed using actual release
- t. conditions. This evaluation requires collecting a sample of the offluent to establish actual radionuclide concentrations and permit evaluating the monitor response. The following equations may be used for evaluating
. compliance with the release rate limit of ODCM 3.11.2.1a:
W.
ODCM-7.0 Revisi::n 6 P;g3 7.0 : p Otb = 1.67 E + 01 e X/O e VF e [(Kg e C)
.i..[({
6
- (7-4) :
Da ".187. E + 01 e X/O e VF e[([Lg + 1.1 Mg] e C6) Where: Dtb = total body dose rate (mrom/yr) , Os = skin dose rate (mrom/yr) , X/O
= atmospheric SITE BOUNDARY dispersion to the c9) trolling location (sec/m l VF = Ventilation System rolesse rate (liters / min)
C1 = concentration of radionuclide I as measured in the grab sample or as correlated from the SPING Noble Oss < Monitor reading (uCl/cc) Kg
= total body dose radionuclide conversion I (mrom/yr per uCi/mfactorf,or fromnoble Table gas 7.0-2)
. l Li = beta skin dose conversion factor fgr noble gas radionuclide I (mrom/yr por uCi/m , from Table 7.0-2)
I Mi = gamma air dose conversion factogfor noble gas radionuclide I (mrad /yr por uCi/m , from Table 7.0-2) i 1.1 = mrom skin dose per mrad gamma air dose (mrom/ mrad) - 1.67 E + 01 = 1 E + 03 (cc/ liter) * (1/60) (min /sec) The above equations may also be used to verify compliance with ODCM
- 3.11.2.1.s when noble gases are detected in periodic (e.g. monthly) effluent noble gas samples.
7A Primary Containment VENTING end PURGING 7A.1 Release Rate Evaluation For primary containment VENTING or PURGING, an evaluation of acceptable release rate may be performed prior to the release. Based on the measured noble gas concentration in the grab sample coliscted per the requirements of ODCM Table 4.11.2.1.2-1, the allowable release rate from primary
. containment can be calculated by the following equation:
RRg= 600. AF 1.67 a + 01 e x/o e pug . Cs) (7-6) er RR, = 8000. AF 1.67 E + 01 e X/O e i 1.1 M il e C 6) {GL + l (7-7)
- - - 's -
R:visl:n 6 Pag 3 7.0-7 Where:
- RRtb = allowable release rate so as not to exceed a dose rate 7 of 600 mrom/yr, total body (liters / minute)
RRs = allowable release rate so as not to exceed a dose rate
, of 3000 mrom/yr, sidn (liters / minute)
, ~AF = allocation factor for the applicable release point from Table 7.0-1 (default value is 0.5 for Reactor Building Exhaust Plenum)
P 500 = total body dose rate limit (mrom/yr) '
. 3000 = skin dose rate limit (mrom/yr)
The lesser value (RRtb or RRs) a~s calculated above may be used for > establishing the allowable release rate for primary containment PURGING or VENTING, taking into account the fraction of the allocated release limit already accounted for by continuous releases from the proposed release point. As discussed in section 7.2.1, this evaluation is rarely necessary. 7.4.2 Alarm Setpoint Evaluation For a primary containment VENTING or PURGING, a re-evaluation of the alarm setpoint may be needed to ensure compliance with the requirements of ODCM 3.3.7.12. For the identified release path (RB Exhaust Plenum or g SGTS) and associated effluent Radiation Monitor, the alarm setpoint should - be calculated using Equations (7-1), (7-2) and (7-3). In Equations (7-1) and (7-2), the value of the Ventilation Flow VF should be established at the total release flow rate, including the contribution from the PURGE or VENT. If the calculated alarm setpoint is greater than the current setpoint, no l adjustments are 'necessary. As discussed in section 7.2.1, this setpoint ' evaluation is rarely necessary. 7.5 Quantifying Releases - Noble Gases The determination of doses in the environment from releases is dependent on the l mixture of the radioactive material. Also, NRC Regulatory Guide 1.21 requires ' reporting of individual radionuclides released in gaseous effluents. Therefore, 1 Detroit Edison must determine the quantities of the individual radionuclides released.
]
For noble gases, these quantities must be based on actual noble gas grab samples. l 7.5.1 Sampling Protocol l l l As required by ODCM 3.11.2.1, a gas sample is collected at least monthl from each of the six gaseous release points (Reactor Building Exhaust' y Pienum, Standby Gas Treatment System, Radweste Building, Turbine Building, Onsite Storage Facility, and Service Building). As discussed in ODCM Section 7.2.2, this gas sample is analyzed by gamma spectroscopy to identify individual radionuclides (noble gases). To date (May 1992) noble gases have been detected only in the reactor building effluent. I
~
-. =- . . . . .. -. - - - . , -
ODCM-7.0 - l n R:visi::n 6 i Page 7.0 d In addition to these monthly samples from each release point, noble gas Y grab samples from the Offges Vent Pipe may be collected using the sample lines of the abandoned Offgas Vent Pipe Monitor (011-N105 and D11-N106). Since noble gases are more concentrated at this point than at 1 the Reactor Building Exhaust Plenum, a greater number of noble gases are detected at this point. Sampling should be performed monthly at the Offgas Vent Pipe unless the reactor is shut down or noble gas concentrations increase sufficiently to allow detection of all significant noble gas nuclides at the Reactor Building Exhaust Plenum. For Containment PURGENENT, samples are collected prior to the initiation of the release and periodically throughout the release (see ODCM ! Section 7.2.1). These samples are evaluated using Equations (7-4) and (7-5) to ensure that the site boundary dose rate limits of ODCM 3.11.2.1 are not exceeded. For an extended PURGENENT period (e.g., longer than 48 hours), i drywell airbome activity levels will equilibrate. After equilibrium is reached,. ! the quantification of the PURGENENT can be adequately addressed by the periodic (typically weekly) sample and analysis of the Reactor Building . Exhaust Plenum or Standby Gas Treatment System. As required by ODCM Table 4.11.2.1.2-1, special samples are required of the , R8 Exhaust Plenum and SGTS following shutdown, startup or a THERMAL POWER change exceeding 15% within a 1 hour period. Exceptions to this special sampling are allowed as noted previously in ODCM Section 7.2.2. , [ 7.5.2 Release Concentration Determination for Reacter Building Exhaust .' Plenum in cases where both a R8 Exhaust Plenum noble gas sample and an Offgas Vent Pipe (OGVP) sample have Deen taken, the RB Exhaust Plenum noble gas concentrations are determined as follows: First, the R8 SPING channel 1-5. readings (above background) at the times the two samples were taken are , compared, and the noble gas concentrations for the sample taken at the lower R8 SPING channel 1-5 reading are normalized to the higher RB SPING channel 1-5 reading. Second, a dilution factor relating OGVP concentrations to RB Exhaust Plenum concentrations is calculated by dividing the R8 Exhaust Plenum flow rate (nominally g.43 E4 cfm) by the OGVP flow rate as indicated in the control room (N62-R808, blue pen). Third, the OGVP noble , gas concentrstions are divided by this dilution factor. Fourth, the diluted OGVP noble gas concentrations are compared to the R8 Exhaust Plenum , I noble gas concentrations, and the higher of the the two concentration l values for each nuclide is taken to be the RB Exhaust Plenum concentration for that nuclide. (For purposes of calculation, the concentrations of nuclides which are not detected are taken to be aero.) Fifth, the resulting R8 Exhaust Plenum concentrations are corrected for variations during the release period by multiplying each concentration value by the average R8 SPING channel 1-5 reading (above background) for the
\
)
l w -- .- _. _ _ - .
Revisitn 6-Pag 3 7.0-9
*. . period divided by the higher of the two R8 SPING channel 1-5 sample readings (above background) at time the samples were taken. These corrected values are then used as Ci in l( Equation (7-8) to determine the quantity of noble gases released.
7.5.3 Calculation of ActMty Released The following equation may'be used for determining the release quantities from any release point based on the grab sample analysis: Og = 1.0 E + OS
- VF
- T
- Ci (7-8)
Where: Og = total activity released of radionuclide 1 (uci) , VF = Ventilation System release rat's (liters / min) T = total time of rolesse period (min) 1.0 E + 03 = milliliters per liter Ci = concentration of radionuclide i as determined by gamma spectral analysis of grab sample (uCl/cc) corrected for variations during release period as described in Section 7.5.2 ,
' 7.6 . Site Boundary Dose Rate - Radiolodine and Particulates~
ODCM 3.11.2.1.b limits the dose rate to <1500 mrom/yr to any organ for 1-131,1-133, tritium and particulates with half lives greater than 8 days. To demonstrate compliance with this limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (nominally once por ! 7 days). The following equation may be used in the dose rate evaluation for 1-131, l-133, and particulates with half lives greater than 8 days: DR = [ (X/Or
- A -131 l
- VFr
- 16.7 '[ C ir) r I (7-9)
Where: . DR = total maximum organ dose rate for all release points (mrom/yr) X/Or
= atmospheric SITE BOUNDARY dispersion location factor (soc /mfog) from release Table 7-3point r to the controlling or plant procedures ;
RI -131 t
= l-131 uCl/m g)hild fromthyroid Table 7-4 inhalation pathway dose factor (mrom/yr per i- .
! = Average ventilation flow for release point r during release period VFr (liters / min) : 'I Cri = Concentration of radionuclide I (1-131,1-133, or particulate with half life greater than 8 days) released from release point r during the l appropriate release period (uCl/cc)-usually determined by gamma spectral analysis of effluent sample ;
- .-. . . -- . _. .- . - -- . - . . . . . -= ..
ODCM-7.0 R:visi:n 6 Page 7.0-10
, 16.7- = 1000 cc/ liter
- 0.0167 min /sec i
. b_ Release periods used in Equation (7-g) are the most recent periods evaluated for the.
different release points, and these periods may not be identical.
~
Alternatively, the site boundary dose rate may be' evaluated using the highest individual isotopic dose factors for all age groups to calculate inhalation and ground plane exposure at ,the highest dispersion factor location at or beyond the site boundary, as well as vegetation, milk, and most exposure at the garden, milk, and '! most locations with the highest deposition factors. Dose rate due to tritium is currently evaluated by this method, and when tritium has been detected in gaseous
, effluents during the most recent release period, the tritium dose rate must be added to the result from Equatica (7-g) to evaluate compliance with ODCM 3.11.2.1.b.
l t Tho' dose rate evaluation described above may have to be performed more frequently j than once per week in order to meet the requirements of ODCM Table 4.11.2.1.2-1, : footnote g: Daily sampling is required following startup, shutdown, or thermal power ! changes exceeding 15% in one hr,ur if DEI has increased by a factor of 3 or if the applicable' noble gas effluent monitor reading has increased by a factor of 3.
- 7.7 Noble Gas Effluent Dese Calculations - 10 CFR 50 l 7.7.1 UNRESTRICTED AREA Dose - Noble Gases ;
ODCM 4.11.2.2 requires that an assessment of releases of noble gases be performed at least once per 31 days to evaluate compliance with the , quarterly dose limits of 5 mrad, gamma-air and 10 mrad, beta-air and the calender year limits 10 mrad, gamma-air and 20 mrad, beta-air. The following equations may be used to calculate the gamma-air and beta-air doses. If noble gases are detected at multiple release points, these i equations must be performed for each such release point, and the ; calculated air doses must be summed. D . = 3.17 E - 08
- X/O * {(M i* Og) '
and , D M = 3.17 E - OS
- X/O * [(Ng
- Q 4)
(7-11) Where:
~
= sir dose due to 0amma emissions for noble gas D[ radionuclides (mrad)
D = sir dose t' a to beta emissions for noble gas radionuclices (mrad) l X/O = atmospheric SITE SOUNDARY dispersion location (sec/m to the cy) trolling Og = cumulative release of noble gas radionuclide i over the j period of interest (uCl) , 1
-(-
Mg = sir dose factor due to gamma orgssions from noble gas ; radionuclide I (mrad /yr per uCi/m , from Table 7.0-2)
*t. .
ODCM-7.0 5 R;visi:n 8 Pag 3 7.0-11 4 T Ni = air dose factor due to beta emisspns from noble gas ' radionuclide I (mrad /yr per uCl/m , Table 7.0-2) 3.17 E - 08 = 1/3.15 E + 07 (year /sec) 7.7.2 - Simplified Dese Calculation for Noble Ganes in lieu of the individual noble gas radionuclide dose assessment presented
. above, the following simplified dose calculational equations may be used for verifying compliance with the dose limits of ODCM 3.11.2.2. If noble gases are detected at multiple release points, these equations must be performed :
for each such release point, and the calculated air doses must be i aummed. (Refer to Appendix 8 for the derivation and justification of this , simplified method.) : o = 2.0
- s.17 E - 08
- x/o
- Me,f * [ oi
)"' (7-12) and ,
D = 2.0
- 3.17 E - 08
- x/O
- Neff * [as 8 (7-13) ;
Where: Moff
= 2.7 (mrad E /yr
+ 03, offectivg) gamma-air dose factor per uCi/m
~t Neff
= 2.3 (mrad E +/yr03, pereffectivg) uCi/m . beta-air dose factor 2.0 = conservatism factor to account for potential variability in the radionuclide distribution ,
7.8 Radiolodine and Particulate Oose Calculations - 10 CFR 50 7.8.1 UNRESTRICTED AREA Dose - Radiolodine, Particulates, and Tritium in accordance with requirements of ODCM 4.11.2.3, a periodic assessment (at least once per 31 days) is required to evaluate compliance with the quarterly dose limit of 7.5 mrom and the calender year limit of 15 mrom to any organ. The following equation may be used to evaluate the maximum l organ dose due to releases of I-131,1-133, tritium, and particulates with i half-lives greater than 8 days: ' D , = (([ (Wr
- SFp
- 3.17 E-8
- Raip,
- oir) prI (7-14)
Where:
~
Deo = dose or dose commitment to Organ o of age group a (identified in Table 7.0-3 or plant procedures) Wr = atmospheric dispersion perameter for release point r and the residence location identified in Table 7.0-3 or plant procedures. Either: l l 4
- - - --- ~ "ODCM-7.0 ' R vision 6 Page 7.0 '
a) X/O. atmospheric dispersion for inhalation pathway ' k.' and H-3(sec/m pathways and C-1g), or dose contribution via other
- b) D/0, atmospheric deposition for vegegation, milk and ground plane exposure pathways (m-)
Raiop = dose factor (mrom/yr per uCl/m3 ) or (m2 - mrom/yr
. per uCl/sec) from Table 7.0-4 for radionuclide I, age group a, pathway p, and organ o as identified in Table 7.0-3 or plant proceduras. Values for Raiop were derived in accordance with the methods described in NUREG-0133. As noted in NUREG-0133 asction 5.3.1.3, in the case that the milk animal is a goat, parameter values from Reg Guide 1.10g should be used. For 1-131, for example, use of the goat feed / forage consumption rate given in Table E-3 and the stable element transfer factor given in Table E-2 of Reg Guide 1.10g results in grass-goat-milk dose factors which are equivalent to .
the grass-cow-milk dose factors in Table 7.0-4 - multiplied by 1.2. Or i = cumulative release from release point r over the period of interest (normally one month) for radionuclide I -- 1-131,1-133, tritium or radioactive material in particulate form with half-ilfe greater than 8 days (uCl). ( SF p = annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist:
- 1) For milk and vegetation exposure pathways:
= 0.5 (derived' from Reg Guide 1.10g, Rev 1. A six ,
month fresh vegetation and grazing season (May through October) limits exposure through this pathway to half the year.
- 2) For Inhalation and ground plane exposure pathways:
= 1.0 (derived from Reg Guide 1.10g, Rev 1) 3.17 E-8 = 1/ 3.15 E7 (year /sec)
This equation should be used to evaluate organ doses for the individual with the highest potential offsite dose. This calculation is performed monthly and is added to previous results for the quarter and year. The highest quarterly and annual cumulative organ dose totals for this individual should be compared with the limits of ODCM 3.11.2.3. t
]
ODCM-7.0 R:visl:n 6' Page 7.0-13 'l I The residence, age group, and relevant exposure pathways for this individual- {. . are listed in Table 7.0-3 and in plant procedures. Plant procedures may i provide updated information which differs from Table 7.0-3. This individual is identified from data obtained in the annual Land Use Census (ODCM 3.12.2).' 7.8.2 Simplified Dose Calculation for Radiolodines and Particulates in lieu of the individual radionuclide (1-131 and particulates) dose assessment presented above, the following simplified dose calculation may . be used for verifying compliance with the dose limits of ODCM 3.11.2.3. However, the result of this calculation should not be used as the reported offsiti dose. , Dmax = 3.17 E - 08
- W
- Rg.131 * [Og ,
(7-15) ' Where: Dmax = maximum organ dose (mrom) Rs.131 = 4.76 E + 10, child thyrgld 1-131 dose factor for the vegetable pathway (m - mrom/yr per uCl/sec) W = highest D/O for residence listed in Table 7.0-3 (m-2) (' Qi = cumulative quarterly or annual release of radionuclide I
-- 1-131,1-133, H-3, and particulates with half lives greater than 8 days (uCl)
The ground plane exposure and inhalation pathways need not be considered when the above simplified calculational method is used because of the overall negligible contribution of these pathways to the total thyroid dose. It is recognized that for some particulate radionuclides (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose , contribution than either the vegetation or milk pathway. However, use of I the 1-131 thyroid dose parameter for all radionuclides will maximize the I organ dose calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway.than 1-131 for the thyroid via the vegetable or milk pathway. 7.g Gaseous Effluent Dose Projection j . As with liquid effluents, the Fermi 2 ODCM controls on gaseous effluents require l
- processing
- of gaseous effluents if the projected dose exceeds specified limits. !
These controls implement the requirements of 10 CFR 50.36a on maintaining and , using the appropriate redweste processing equipment to keep releases AI. ARA. ' ODCM 3.11.2.5 requires that the VENTILATION EXHAUST TREATMENT SYSTEM be used I to reduce radioactive material levels prior to discharge when the projected dose exceeds 0.3 mrom to any organ in any 31 day period (i.e., one-quarter of the design objective rate). Figure 7.0-1 presents the gaseous affluent release points and the VENTil.ATION EXHAUST TREATMENT SYSTEMS appilcable for reducing effluents prior to release.
- Revisisn 6 I Page 7.0-14 f
'? Dose projection is performed at least once por 31 days using the following equation: -l
.h Dmemp = Dmax * (31/ d)
Where:
,- Dmaxp - = maximum organ dose projection for the next 31 day period (mrom)_
s NOTE: The reference calender quarter is normally the current calendar quarter. If the dose projection is done in the first month of the , quarter and is to be based on dose calculated for the previous quarter, the reference calender quarter is the previous quarter. . Dmax = the cumulative maximum organ dose from the beginning of the ; reference calendar quarter (normally the current quarter) to the end of the most recently evaluatoo Misase period as determined by Equation (7-14) or (7-15) (mrom) d = number of days from the beginning of the reference calendar quarter to the end of the most recently evaluated release period. 31 = number of days in projection 7.10 Waste Oilincineration As indicated in Table 4.11.2.1.2-1, waste oil containing radioactive material may be incinerated by injection into the Auxiliary Boiler fuel stream, but this shall only be done according to approved plant procedures. Prior to incineration of such oil, a sample of the oil must be analyzed by gamma spectroscopy to determine . ; radionuclide concentration. Based on the analysis results and proposed incineration rate, the proposed release must be determined to be within site boudary dose rate limits (when combined with dose rates from other release points) prior to , incineration. After incineration, the activity released must be determined and recorded. The equations of this ODCM section (section 7.0) should be used in these calculations, and the highest dispersion factors in Table 7.0-3 may be used, unless data specific to the Auxiliary Boiler stack is available. (Since this stack is farther from most land receptors than other plant stacks, Table 7.0-3 dispersion factors should be conservative foi this release point.) END OF SECTION 7.0
~
l
- j l
. _ _ _ _ . . .__ _ _ ~_ _ . . _ . . _ . __
- l. ODCM-7.0 l
~ RIvlsl:n 6' Pags 7.0-15 TABLE 7.0-1
{ Values for Evaluating Gaseous Release Rates and Alarm Setpoints Allocation Allocated Does : Release Point Plow Rate
- Factor Rate Limit i Siter/ min) . (AF) (mrom/ year)
Re:ct:r Building 2.67E6 0.50 T Body = 250 l Exh:ust Plenum Skin = 1500 ' D11-P280 Organ = 750 i St:ndby Oss 1.07ES 0.10 T Body = 50 Treatment System. Skin = 300 Div i D11-P275 Organ = 150 Standby Oas 1.12E5 0.10 T Body = 50 Tr:stment System Skin = 300 Div 11 D11-P276 Organ = 150 Turbin) Building 8.67E6 0.20 T Body = 100 V:ntilation ' Skin = 600 D11-P27g Organ = 300 Srrvice Building g.06ES 0.01 T Body = 5
- 1. 'lation Skin = 30 D.:-P282 Organ = 15 Radweste Building 1.13E6 . 0.02 T Body = 10 V:ntilation Skin = 60 D11-P281 Organ = 30 :
Onsite Storage 3.06ES 0.02 T Body = 10 Buliding - Skin = 60-V:ntilation Organ = 30 , D11-P281 ' l R: actor Building - 2.57E6 0.50 T Body = 125 l V:ntilation** Skin = 750 { Culf Atomic
. M: nit:rs D11-N408,N410 1 l
1 V:ntilation flow rate values are subject to change due to plant modifications and changing plant
. conditions; therefore updated values in plant procedures may be used. -
D11-N408 and N410 will start the SGTS, close the Drywell PurgeNont Valves, isolate Rx Building V:ntilation System, isolate Control Center, and initiate emergency recirculation mode. I l
DDCM-7.0 R:visi:n 6 Page 7.0-16 ~ TABLE 7.0-2 Dose Factors for Noble Gases
- Total Body Skin Gamma Air Beta Air
~
Gamma Dose Beta Dose Dose Factor Dose Factor Nuclide Factor Ki factor Li Mi Ni (mrom/yg)per uCl/m (mrom/yg)per uCi/m (mrad uCi/m /yrf)or (mrad uCl/m /yrf)or Kr-83m 7.56E-02 --- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+ 03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01 E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+ 02 1.56E+02 1.11 E+03 X:-133m . 2.51 E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 X:-135 1.81 E+03 1.86E+03 1.92E+03 2.46E+03 X:-137 1.42E+03 1.22 E+04 1.51 E+03 1.27E+04 X:-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+ 03 2.69E+ 03 9.30E+03 3.28E+03 ( N2TE:
- Dose factors taken from NRC Regulatory Guide 1.109 b.
t ODCM-7.0 R visi:n 6 Page 7.0-17 TABi.E 7.0-3
.k
. Controlling Locations, Pathways, and Atmospheric Dispersion for De,se Calculations
- Atmosphoric Disporsion Factor ODCM Location Pathway (s) Controlling X/O D/O Control Age Group (soc /m )
8 (1/m2) 3.11.2.1a site boundary noble gases N/A RB: 1.25E-6 N/A
$ 07 d, NW) direct TB: 5.71E-6 exposure RW: 2.66E-6 3.11.2.1 b site boundary inhalation child RB: 1.25E-6 N/A (0.57 mi, NW) TB: 5.71E-6 RW: 2.66E-6 .
3.11.2.2 site boundary gamma-air N/A RB: 1.25E-6 N/A
. (0.57 mi, NW) beta-air TB: 5.71E-6 RW: 2.66E-6
( 11.2.3 residence vegetation child RB: 1.10E-6 1.5gE-8 (0.67 mi, WNW) inhalation, and TB: 4.02E-6 3.06E-8 ground plane RW: 1.53E-6 1.76E-8 NOTE:
- The identified controlling locations and pathways are derived from land use census data and dispersion and deposition factor data tables. The dispersion and deposition factor values listed are
' conservative values; they represent the highest annual average values seen between 1984 and 19g2. When performing dose and dose rate evaluations for plant surveillances, the dispersion and deposition factor values and location and pathway information found in plant procedures should be used. These data in plant procedures should be the same as the above data unless recent a information has shown the above data to be non-conservative or inaccurate. When performing dose evaluation for the Annual Effluent Release Report, the annual average dispersion and deposition factors for the year being evaluated should be used.
( e a"
. - - - . , , . . _ - . . - _ . . ___ - ~
_ . . - . - - ~ . aue Sonstitise pesmoor asse tennere - asus.7 ODCM-7.0 R2vislan 6 1
~ terever ser aiu 9 Pago 7.0-18 ;
.i Seelide . Bees Meer 9tyroW Reeosy Asas abl&I T.teer 5-3 .
.- 1.Mb3 9.MS*3
- t.M8* 3 ' t.368*3 1.368*3 1.36tel bu 1.838 4 3.418*3 3.418*3 3.418*3 3.etb3 3.41b 3 3.418*3
. Be*24 1.838+4 1.038 4 1.438 4 1.038*4 f.8384 1.038+4 1.43b4 P-33 t.338+6 7.718*4 . . . 8. M be 5.91t*4 tr St .
- 8.96bt 3.38tet 1.M8*4 3.32b3 1.008*3 no-54 . 3.968 4 . '9.84tel 1.48be 7.74be 6.Seb3 me M = 1.M8+e . s.Mtes . 9.ub3 3.038 4 1.8M-1 '
fe-SS 3.448 4 1.30B+4 . . 7.318 4 4.038*3 3.Mb3 fe-H 1.tebe 3.7084 . . 1.e38 4 1.eabl 1.eud
. to-57 - 6.938*3 = . 3.7tb5 3.M84 4.7tb3 i
es le = t.988 3 .~ - 9.388+l 1.06bl 3.07b3 ' es-te = 1.ttb4 * . l.978 4 3.858*$ 1.44b4 5443 4.338 $ 3.1484 * . 1.788+l 1.M84 9.45b4 , sadl 1.M be 3.908-1 . . l.Get*3 1.838 4 9.138-3 i eret = 1.448*0 . 4.63be 6.78b3 4.908 4 6.tSbt ] Se45 3.34t+4 1.03E+l
- 4.90E+4 8.M8+l 5.348 4 6.Mtd ,
as 49 3.388 3 6.518-3 - 4.32b3 9.308*3 1.6M*1 4.838-3 tr-83 . . . . . 1.068 4 1.35t+4 ar-43 = * = . . 3.33b3 3.etp3 ! tr se . . . . . - 1.kb3 3.13b3 i tratl . . . = . . 1.388 1 , ab-te . 1.358+l . . . l.MS*4 S.90E*4 , ab-80 = 3.478*3 - . = 3. Mbt 1.938 3 DD-89 . 3. NE* 3 . . . . 1.708*3 l Statt 3.NE*l . . . 1.4eb6 3.98tel 8.73b3 Gr-90 9.938*7 - * . . 9.408 4 7.32bl 6.teb6 , Statt 4.h t . . . 3.668 4 1.915+5 3.008 4 } St 93 4.MS*0 = = . 1.658+4 4.308 4 3.91bt - F9e 3.998*3 = . . l.7ept 5.Mbl S.ettet b9te 3.618-4 . . . 1.93p3 1.388 4 1.03F3 b98 4.638*$ = = . 1.70E*6 3.858+l 1.341 4 bi2 1.038+l . * = 1.l?t*4 7.358+4 3.83bt b93 9.ubt = = . 4.861*4 4.32bl 2.6tb8 > 1r-98 t.07tel 3.ube . l.48t*4 1.778+6 1.tetel 3.331 4 3r-97 9.48bt 1.M8*1 . 3.978*1 7.878 4 S.33bl 9.Mbe ( sb 95 1.418*4 7.838 3 = 7.74te3 3.elbl 1.NEel 6.318*3 ab-97 3.33bt S.63* 3
- 6.W3 3.40t*3 3.43b3 3.elb3 .
ar99 . 1.31843 . 3.91843 9.138*4 1.48bl 3.Setet Te 99e 1.038 3 3.9tF3
- 4.431 3 7.Mb3 4. ME*3 3.708-3 Te let 4.988-8 4.835 8 . 1.088-3 3.99t*3 . 1.908-4 .
Betel 1.938*3 .
. l.038*3 5.eltel 1.tekt 6.Sebt aclel 7.90S*1 . . 1.938*0 1.10t*4 4.838*4 3.115 1 to-lh 6.988+4 . . 1. h t
- 5 9. ht*6 9.138*5 8.738*3 as-tene . . . . . . . . i 36-106 = . . = . . . ;
I as-Stes 1.e88*4 1.00t+4 . l.97be 4.6384 3.03bt 8.hp3 - 5 -134 3.138+4 8.898*3 7.S$8*1 = 3.488 4 4.ettel 1.34t+4 , Sb-tal S.ME*4 S.95b3 S.4eE*1 - 8.148 4 f.etbl 1. Mad fe-tale 3.43843 1.58b3 1.05t*3 1.34E+4 3.ubt 7.068*4 4.67bt to-t37e 1.Mbe 5.778*3 3.39b3 4.9884 9.tebl 1.tebl 1.17E*3 fe-137 1.400 4 6.438 1 1.068 4 S. teed 4.818*3 $.7484 3.988-1 l Te tite 9.168*3 4.478 3 3.M8* 3 3.M8+4 1. ht+6 3.83bl 1.988*3 t to-139 4.98ba 3.39b3 3.908-3 8.87bt t.ME*3 t.578 3 1.Mb fo 13te 6.998*1 4.368+t 1.80E*1 3.99E*3 1.Mt* 4 8. ht*l 3.90s 9t 9e 131 f.9tb3 8.96b3 9.Mb3 4.37b3 9.Hb3 5. Mbt 3.Hb3 .
- So-133 3.0e8*3 3.23 1.988*3 f.44b3 3.088+l B.tetel 1.638*3 b130 4.988*3 1.M84 1. ME*6 3.008 4 . 7.69E*3 S.388*3 b181 8.538 4 3.988 4 1.198*7 8.138 4
- 6.388*3 3.058 4 >
bt33 5.968*3 3.M8* 3 1.ME*l S. tab 3
- 4.968 3 t.teb3 l l 133 4.Mb3 1.488 4 3.158 4 3.Sebe . 4.00b3 4.588 3 !
I-t h 4.MS*3 8.738*3 3.988 4 3.758*3
- 1.918 4 6.tlb3 bill 3.44848 6.98b3 4.48pl 1.118 4 . $.358*3 3.57bl es.tM 3.738*$ 8.48tel = 3.87 Eel 9.76t+4 1.Mbe 7.34bt es tM 3.908 4 1.44D S = 0.568+4 t.388 4 1.1?be 1.tetel -
en-137 4.388*5 6.31bl = 3.33bl 7.534 4 8.40b3 4.3845 ! i es-VM 3.M8*3 6.398 3 = 4.00B*3 4.86tet t.868-3 3.MS*3 Se 139 9. Mbt 6.468-4
- 4.328-4 3.768*3 8.96b3 3.MS-3 to Me 3.90t*4 4.9ept . 1.67tet 1.373 4 3.188*$ 3.878*3 >
to-141 1.Gebt 7.538 8
- 7.e98 5 1.Mb3 1.ME-7 3.368 3 as let 3.638 3 3.708-8
- 3.398-8 1.1943 . l.Mb3 i
. &e-144 3.Mb3 t.74b3 .
- 1.htel 4.Setel 4.90b t
&e ut 6.83bt 3.908 1 ^ =
- 6.3M*3 3.198*3 7.738 3 '
to-141 t.9984 9.388 4 . 4.ME*3 3.638*S 1.30tel S.llb3 es-MS t.068*3 9.30bt . 6.eebt 7.908+4 3. Mbt 1.53849 Go-Me 3.435 4 s.438 4 . 8.488*$ 7.788+6 8.Mbl 1. Mbt pt 943 9.368*3 S.758*3
- 3.tet*3 3.81tel 3.estel 4.44B+3 pr*M4 3.0tb3 1.3SF3
- 7.eSt 3 5.938*3 3.t M-e 1.838-3 N -947 5.37s*3 4.908*3 . 7. **3 3.Sitel 1.7M*5 3.888*3 tklet 8.488 4 7.008*4 .
- 3.90t+4 1.55tel 3.4ebe '
ar 339 3.3eb3 3.Mte t . . 7.00t*1 3.76t*4 1.19bl 1.Mtel
. _ _ . _ _ _ _ - . _m . , - _ _ _ . _
. . . . . . ~ . . - - . . . . - - . -- -
tabe ?c-a teentw4) ODCM-7.0 -
# sue sens Mle pan ,a.e. reu.te - tsanasta-- R:visiin 6 tareary, p e acu.8) Page 7.0-19 smute a uew rarrass sM , le 8 cs-saa v. amer a 93 . 1.3?bl 1.3?t*3 1.378*3 1.37t*3 1.3?b3 1.378*3
; 1 CM '3.488 4 4.S?8e3 4.878+3 4.87t*3 4.87t*3 4.8?b3 4.8?b3 Se-M 1. Mt+ 4 1.388+4 1.388+4 1.304*4 1.308+4 1.388*4 1.308+4 P 33 1.09b6 ~1.108*$ = = . 9.388 4 3.Mbe Cr-ll . . 7.leb t 3.9?b t 3.90t+4 3.Seb3 1.Hb3 -
, 1 Its M = 5.114 4 . 1.378+4 1.tebe 4.44b4 8.448*3 1
" ,- . no M = 1.7e84 . 1.798+4 1.538 4 $.Mid 3.63bt Fe45 3. M t+4 3.30be . . 1.348*5 6.HE* 3 5.Mb3 Fe-59' 1.598+4 3.708+4 * . 1.$384 1.?tbl 1.ube ~ ,
cel? . 4.93b3 . * ' l.06bl 3.M84. 9.3Eb3 es lo . 3.078+l . . t.Mbe 9.8384 3.78b3 Code = 1.518+4 . . 3.738 4 3.99bl S.988 4 ' 3143 S.tehl.4.M84 . . 3.078 5 1.438 4 1.988 4 3446 3.188 8 3.93bt . . = 9.368*3 3.678 4 1.3?bt ende . 3.93be . 6.498 4 1.198 4 6.Mbe 8.48bt So45 3.Mt+4 1.ME*l = 8.Mt+4 1.24E+4 4.ut4 6.M84 : Se49 4.8 M 3 9.305 3 = 4.038 3 1.38b3 3.058*3 6.Hb3 St-83 . . . . . . 1.838+4
*- st-e3 . . . . .. . 3.ut*3 St-44 . . . . . . 4.338*3 i
St-06 . . . .. . . 1.038+t abe6 . t.90sel . '. . 1.??be 4.4et+4 ape 8 . l.468*3 . . . 3.93bl ~ 3.73bt 8>09 . 3.53M3 . . . 3.388 7 3.33b3 St=89 4.pbl . ". . 3.438+6 3.718+5 1.368 4 St 90 - 1.08t+8 . . . 1.458*? 7.6Sbl 6.48be Statt O.toget . . . 4.978+4 3.598+l 3.898 4 st 93 9.538 0 . . . 3.748 4 1.19tel 4.06b t b90 3. 98b 3 . . . 3.93bl S.99bl 8.00b t b9te 3.?tb1 * * = 3.30b3 3.03bt 1.43F3 b91 6.6ttel . . . 3.96t+6 4.99 4 l 1.??34 b93 1.478*1 . . . 2.68t+4 1.658 5 4.39t*t 1 93 1.3Sb3 = = . s.33t+4 5.79bl 3.73b9 Statl 1.46bl 4.98be = 6.74E+4 3.698+4 1.49Dl 3.158+4 8t-97 1. NE*3 3.738*9 . 4.13b t 1.388 5 4.388+5 1.3M*1
. e>95 1.Mt+4 1.038 4 = 1.00t 4 7.lttel 9.688 4 5.Mb t -
i se9? 3. Mbt 7.788 3 . 9.13E 3 3.938 3 3.17E*3 3.848 3 no 99 . 1.698+3
- 4.t t b 3 t.ktel 3.69bl 3.33b t te-90s 1.NE 3 3.M8 3 . l.76b3 1.188*3 6.138*3 4.99bt .
Teatet . l.938-5 4.408 5 . 1.632 3 6.478 3 8.73b? 8.M84 De-le3 3.teb3 . . 7.43b3 7.8Mel 1.99tel 8.Mb3
- Se=tel 1.138+0 . . 1.41be 1.838+4 9.M t+4 4.ME 1 Se 106 9.Nbe . . 1.90E*6 1.618e? 9.eebl t.34t+4 3 > tele . . . . . . .
Sett6 . . . . . . . as-ttes 1.Mt+4 1.348 4 . 3.988+4 4.?lt*4 3.? Mel ?.998 3 sel34 4.M8+4 7.948*3 9.768*1 . 3.058+4 3.988*$ t.488+4 Sklal 7.Mtd 8.00b3 7.848*1 . 3.148+4 9.934+4 1.?as 4
- 9e-tale 4.888*3 3.34bt 1.488*3 e. 5.Mtel ?.tept 6.6?b3 to-187e 1.00b4 8.tepa 4.3eb3 6.Mbe t.66be 1.lttel 3.taba fe 137 3.9tt+0 9.1381 1.438*4 ?.30t+8 1.138 4 8.08b e 4.438 9 Setate 1.398 4 6.$88*3 4.34b3 S.198+4 1.988 4 4.058+5 3.3Sb3 fe-139 7.108 3 3.3Sb3 5.188 3 3.ME-1 3.ME*3 1.63b3 t.768-3 ,
Tet3te 9.M8*1 4.91E*1 ?.3Sb t 4.39t*3 3.383*5 6.3ttel 4.038 1 1 te 135 1.58b3 S.33FS t.M8 3 6.9803 3.Mp3 9.6tpt S.tob3 fe-133 3.408*3 3.90t*3 3.Mba 1.9tb3 4.498*5 4.638*$ 3.19b3 l 3-tae 6.Mb3 1.798+4 1.498+4 3.?lE4 = 9.13b3 7.178*3 3+131 3.94t*4 4.918 4 1. Hie? 8.488+4 . 6.49t*3 3.Mbe b133 1.$9t*3 4.38b3 1.lttel 4.938*3 . t.378*3 1.94b3 3 133 S.3344 3.elb4 3.938+4 3.Hb4 . l.eM4 6.32b3 3 th 8. Bet *3 3.33b3 3.958 4 3.Mb3 = 3.04t*1 8.48t*3 ' 1 I 3 139 3.70b3 9.u t*3 6.Sibi 1.49b4 . 6.Mb3 1.49b3 Co tM S.tabl 1.138 4 . 3.?ltel 1.46tel 9.?68*3 S.49bl , co-tM $.198+4 1.M8+ l = 0.10E*l 1.?88+4 1.09t*4 9.378+5 I es 137 6.Petel 8.48tel
- 3.Mbl 1.*tbl 8.48b3 3.ttbl es tM 4.448*3 0.948+3
- 6.638 3 7.87tet 3.108-1 4.448 3 to-139 1. Mt+0 9.u t-4 . S.sebe 4.48*3 4.458 3 3.908 3 to Me 4.478+4 8.Me+ t . 3.388*1 3.83t+4 3.398+5 3.838 3 as ut 1.43bt 9.868 4 = 9.Mbl 3.39E*3 F.4484 4.MS-3 as tea 1.188-3 3.108-5 . 3.Mbl t.91b3 . 3.3?b3 34*H4 4.79b 3 3.368+3 e . 3.54t+6 4.378 5 6.M8*1 g
le* M3 9.488 9 4.85bt * . 1.63B+4 1.30E+4 1.86bt Get41 3.04t+4 f.958+4
- 8.885 3 6.948*l 1. MS*5 3.178*3 Sr MS 3.Mp3 1.Mb3 . 8.Mtet 1.30t+5 3.568+l 3.16bt GrM4 4.09b6 3.eas+e . 1.31t+4 1.ME*? 8.64bl 3.63h3 Pt-M3 1.M8+4 S.31b3 . 3.09t*3 4.esbl 3.M8+$ 6.63b3 Pt-144 4.105 3 1.768-3 . l.018-3 1.758 3 3.358 4 3.10b3 34-147 ?.84t*3 4.568 3 . l.638*3 3.73tel I.838 5 5.t38+3 16 587 9.30bt 9.Mbe .
- 4.?C J I.??t+6 3.43bt ar-339 3.388 3 3.198*1 = 1.088*3 4.49t+4 1.33t+$ 1.?? Set l
l I
w , , , , - , _ Table 7A-4 (reettauen ' a.2* 3enel:ues postner seee rees.te . csu
- ODCM-7.0 I
teree/rr per scuey Revisi:n 6 8 -ua Page 7.0-20 sees ue . sees ueer 9errou sweet seu t. seer 53 e 1.138 3 1.138*3 - 1.138 4 1.138 3 1.138*3 1.13bt ,
- a. '
C te 3.l9b4 6.738*3 6.7M* 3 6.7M*3 6.73b3 6.738 3 6.73h3
~/ so-34 9.41t*4 1.618 4 1.618+4 1.618*4 1.618 4 1.6tt*4 1.618+4
{-* P43 3.60t*6 9.Mbl . * *
- 4.32b4 9.888+4 St St *
- 8.llb1 3.4He t 1.788 4 1.088*) .1.94b3 ,
. Ne*le = 4.39t*4 = 1.0et.4 t.Mt+4 3.398*4 9.918*3 Ne* M = 1.M84 = 1.678*0 1.31b4 1.338*l 3.13bt i Fe ll 4.74t*4 3.53be = = 1.1ttel 3.47s*3 7.778 3 Fe-59 3.978 4 3.ht*4 - = ~ 1.37b6 7.478+4 9.678 4 Se-57 - 9.038 3 6 *. S.07bl 1.338 4 1.07Be3 -
8 -98
- t.778*3 *
- 1.198 6 3.M8*4 3.H8*3 !
Co-60 = 1.3tb4 *
- 7.078*6 9.638 4 3.365 4 +
31 63 8.3tb l 4.638+4 =
- 3.78bl 6.3M*3 3.00b4 51-45 3.998 4 3.Mb t .= = 8.185*3 8.448+4 1.Mb t 8544 = 1.998 4 = 4.9M4 9.688*3 3.678 4 1 0784 Se-el 4.Mtd 1.tMel
- 7.148+4 9.95tel 9.6M4 7.038 4 Se49 6.708 3 9.Mb3
- l.8Sb3 1.43b3 4.ets4 8.938-3 Stata = = = * *
- 3.99b4 i tv-el - * * * * = 4.M8*3 ge-se . . = . =
- 5.488*3 tr-el . * = = = = 3.538*1 SkN
- 1.9BE*l - = = 7.99b3 t.ubl apes = 5.63b3 = =
- 1.7tb1 3.44b3 3>09 = 3.4Sb3 = = = 1.09be 3.90tet ar-89 9.998 5 * - - 3. M8 +6 1.678el 1.738+4 i ar-90 1.9tb8 * = = 1.48t*7 3.438+$ 4.M84 St-91 1.818*3 * *
- 5.33b4 1.74 Bel 4.99be St-93 1.318*1 = = = 3.4eb4 3.43bl 5.388 1 b90 4.118*3 = =
- 3.638*l 3.Mbl t.118*3 b9te S.07bt = = = 3.015*3 t.738*3 1. Mbt b91 9.14t*5 - =
- 3.6M4 1.Mbl 3.44be bt3 3.h t* 1 = = = 3.398 4 3.398*$ l.Stbt i bt3 9.848*3 = = = 7.44t*4 3.898el S.115*4 3r 95 1.90bl 4.188*4 = l.96be 3.33be 4.118*4 3.708 4 tr-97 9.Mt+3 3.73b t = 3.89pl 1.tMel 3.ltbl 1.685*1 ,
putt 3.355*4 9.18tel = 8.63b3 6.utel 3.70t*4 4.558 4 se9? 4.398-1 7.308 3 = 8.558 3 3.43b3 3.788 4 3.688 3 l ee-99 = 1.738 3 = 3.935*3 1.368*5 1.378*5 4.3441 fe-99e 1.788 3 3.488 3 = l.078 3 9.$18*3 4.8tb3 S.778-3 fe-101 8.448 5 8.Stbl = 1.451 3 S.Siba 1.H8+l f.etb3
- to-103 3.79t*3 = = 7.83b3 6.63bi 4.44b4 1.07E*3 to-tel 1.53be =
- 1.Mtd 1.59t*4 9.958*4 $.558-1 .
Se-106 1.Miel = = 1.84b5 1.438*7 4.395el 1.6984 8b-103e = = = * * *
- l ab-106 = = = = * * =
as flee 1.498+4 1.Mbe = 3.138*4 S.488*6 1.008+l 9.ut*3
- Se134 S.74be 7.40t*3 1.NE* 3 = 3.348*6 1.648*$ 3.808*4 SW135 9.848+4 7.ltt*3 9.letet = 3.338+6 4.93t+4 3.075+4 fe-t3le 6.73b3 3.3He3 1.938*3 = 4.778+l 3.38be 9.MS*3 fe-137s 3.498 4 8.llb3 6.078 3 4.ME*4 1.48t*4 7. M8*4 3.e38 3 Te-137 3.778 4 9.518-1 1.Mbe 7.0784 1.Gebe S.6384 6.118-t Te-tits 1.938 4 6.858+3 6.338*3 8.83t*4 1.76be 1.038*l 3.04b3 7e 139 9.778*3 3.308 3 7.44R 3 3.87P t 3.9M* 3 3.368+4 8.388-3 fe-tale f.MR*3 S.938et 9.77b t 4.80S*3 3.NEen 3.estel 3.eftet fe-131 3.178-3 8.Hb3 1.708 3 S.teba 3.088*3 1.338*3 4.99b3 fe-133 4.818*3 3.73b3 3.17E*3 3.778*) 3.778*$ t.388*3 3'.638*3 3-130 8.188,3 t.M8*4 5.85t+4 3.4 M+4 = l.ttE*3 8.448 3 ~
3-131 4.818+4 4.81B+4 1.638*7 7.8Sb4
- 3.045*3 8.738+4 3-133 3.138e3 4.07t*3 1.96tel 6.35t*3 = 3.3eb3 1.e03*3 3 133 S.M8*4 3.038 4 3.85t+6 3.388 4 = l.488*3 7. Pes *3, 3-tM t.1?b3 3.Mb3 5.978 4 3.30t*3
- 9.llb3 9.958*3 1 138 4.93tel 8.738*3 7.938el 1.M8*4 = 4.448*3 4.Mb3
. es 1M 6.lttel 9. Stb 6 = 3.308 4 1.31b5 3.868 3 3.388+l CoatM 8.518 4 1.718+l = 9.llb4 1.458+4 4.188 3 1.148+8 Co-137 9.87bl 8.3Spl -
- 3.83bl 1.Mbl 3.638 3 1.388+l Ce=tM 6.33b3 8.48bs
- 6.32b3 6.sttet 3.MS*3 8.S$8e3 to-139 1.MR4 9.86b4
- 8.635 4 5.7?b3 S.778+4 S.378 to-Me 7.488 4 6.48b t * .
3.11b1 1.748*6 t,93b l 4.330*3 i' to-tet 1.968-1 1.99be
- 9.47b8 3.988*3 3.7Het 4.NE-3 to-M3 l.eepa 3.ee8-l
- 3.9tbl S. M b s 3.74be 3.79s.3 to-tee 6.44b3 3.388+3 * - 1.8Mel 3.36b l 7.Steet 8a-M3 1.Sehe 4.1thi =
- 8.70t*3 7.99b4 t.398-1
, l So-Mt 3.91b4 1.96be = 8.558*3 5.44tel S.MS*4 3.988*3 ' Co-143 3.Mt* 3 1.99pt.
- 8.36tet t.ttsel 1.378el 3.878*1 ee-M4 4.778+6 3.13be = 1.178 4 1.3e8 7 3.8905 3.ettel pt 143 1.858 4 5.llb3 = 3. cob 3 4.338 5 9.73be 9. M b 3 l Pr-M4 S.ME-3 1.058 3 '
- 5.778 3 1.lfb3 1.9?bt 3.0e8-3 '
abt47 1.888 4 6.73b3
- 4.018*3 3.38tel 8.318+4 6.8tb3 46 187 t.638*1 9.Mbe = = 4.tiges p.teges c38+e er-839 4.ees*3 3.Mb t = 9.?M*1 S.ett+4 6.40s*4 3.H8*1
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+-
5-3 = 4.478*3 6.47E*3 4.4?b3 4:4?t*3 4.4343 4.47b3 C-te *3.658 4 S.31t*3 $.3itel 3.11t*3 5.3tb3 S.318*3 S.318*3 me 34 1.Mt+4 1.eM*4 t.ME*4 t.mt*4 1.Mbe 1.mt*4 1.M84
*. b33 3.elb6 1.tabl = .
- 1.618*4 f.74b4 Ct=51 = = $.?ltet 1.32b1.1.M84 3.87b3 8.958*t "b * *
- 3.llt+4 = 4.90t*3 1.eet*6 7.Mb 3 4.98t*3
= hrM = .' l.545 4 .. 1.108 4 1.358*4 7.t?b4 3.318 1 Fe-ll ' t.978+4 1.17E*4 = = 8.69t* 4 1.09E*3 3.338*3 i Fe-M 1.M8+4 3.35b4 .. . 1.038 4 3.44be 9.488*3 ,
j~ , en 57 a 4.ltb3 = .- 3.?tbl 4.Mb3 6.418*3 erle = t.33b3 .=
- 7.77tel 1.198 4 9.838+3 '
4.llt+4 3.198+4 1.188*4 co-te = 8.038*3 . .* 5143 3.39tel 3.Mtd *
- 3.09tel 3.43b3 1.448 4 si45 3.Ht4 3.842-1 = = 8.13b3 $.Stue 1.33bt ende . 1.388 4 . 3.988 0 9.308*3 1.905 4 7.74bt .
Se45 9.9M4 6.Mb4
- 3.36t+4 6.47pl S.M84 3.198+4 Be49 S.W3 9.678-3 = 4.838 3 1.478 4 1.338 4 7.188 3 St=83 = = = . =
- 1.3M+4 Dr=43 = = = = * .
- 3.etb3 st=e4 = . * * * . 4.gebt 3r.45 * = = * . = 3.e4t+t Spee = t.90tel . = = 3.k t*3 8.838+4 R>as
- 8.57b3 -- . . 3.39t*3 3.87t*3
- Seet = 3.318*3 a =
- 4.835*t 3.06ba St-09 3.908*$ ' . *
- 3.eM+4 6.4044 1.H84 I' Br-te 4.ette? = .
- 1.13te? 1.318*$ 3.998 4 St-91 9.M&et . = = $.3484 7.M t+4 3.448 4 St-93 1.068*1 = = = 3.M84 1.488+l 3.918+t '
F9e 3.39t*3 e' . . 3.69E*$ t.M8el 8.838*l F9te 4.efbt = = = 3.795 4 3.3Sb3 1.39b3 b91 5.088*$ = * = 3.4SE*4 7.eM*4 1.575 4 b93 1.Mb t * = = 3.45be 1.37b5 4.41bt bf3 t.Setet = = = 7.64t*4 1.678,5 4.eftee tv-95 1.titel 3.19b4
- 3.11t*4 1.758*4 3.97844 3.eH+4 5t=97 1.Seb3 3.Sep t . 3.59E*1 1.letel 1.4ept 1.17b t '
e>95 1.578+4 6.43D3 = 4.735 3 4.79tel 1.378 4 3.748*3 k up97 3.43bt 7.39b3 = 5.708-3 3.338*3 3.698 4 3.6M-3 me-99 = 9.65ba - 3.458*3 1.M8+l 4.878 4 3.338+l fe-90s 1.408 3 3.esb3 = 3.ttb3 8.118*3 3.08t*3 3.738-3 fe-let 4.118 5 8.33bt = 9.?98 4 5.84t*3 8.448*3 4.133 4 Actel 3.83b3 = = 4.34t*3 S.535*5 1.618 4 6.79b3 Artet *t.338*e = = 0.992-1 f.87b e 4.848 4 4.10b1 Bele6 8.Mb4 =
- 1.07bl 1.48*7 1.Mbl 1.098 4 Ab-tele * = = = = =- .*
* * * = = = =
gh 104 = 4e-114e 9.90t*3 f.33b 3
- 1.098+4 3.67t+4 S.Sebe 5.0043 a>134 3.798+4 5.Mp3 1.01b3
- 3.658 4 5.9tb4 1.308 4 S>135 S.17be 4.778*3 6.3He t = 1.p t+6 1.478+4 1.098+4 fe-135m 4.748*3 1. 99b3 1.438*3
- 4.47bl
- 395+4 6.988*3 fe-tate 1.678 4 4.9054 4.07t*3 3.7584 .1.3 tb4 s.738 4 3.078*3 fe-137 3.3M4 9.538 1 1.elbe 4.Mtee 1.038+4 3.448*4 4.09bt fe 139e 1.41E+4 4.09t*3 $.478*3 3.198 4 t.688*6 6.908 4 3.33b3 fe-139 7.888 3 3.47b3 6.?tb3 1.758-1 3.008*3 3.638*4 1.00b3 Te-tite 1.078 4 S.lebt 8.95t*1 3.698*3 1.99bl 1.198*$ 3.638*1 fe-tSt s.74>3 8.338 3 8.5e8 3 3.99b3 3.M p 3 3.33b 3 3.ee8-3 fe 138 3.73b3 3.3?p3 3.79t*3 1.e3t*3 3.4etel 4.4184 1.76bt blae 6. ME* 3 1.398*4 1.408+4 1. 5M+4
- 1.998*3 S.578*3 3-931 3.798 4 4.44t+4 1.485*7 S.548 4
- t.M8* 3 t.MS*4 3-133 1.69b3 3.94t*3 1.69tel 3.958 4
- t.90t*3 1.Mp3 3.tX t.Sthe 1.93be 3.H8+6 3.348 4
- 3.tob3 8.eeb3 3 tM 9.3tb3 1.0084 4 Al8*4 3.ME*3 = 1.398*3 6. ele *3 3-135 3.Mb3 7.40b3 6.M b l 8.47t*3 = 1.elb3 3.??b3 artM 3.Mbl 7.03bl
- 1.9etel f.9?b4 1.3M*3 7.458+4 Co-1M 4.8M*4 1.36bl
- 8.Mtd 1.188+4 1.438*3 8.398+4 se-87 5.49bl 6.13bl = t.Patel 7.938*4 S.3M 4 4.558+4 es-tae S.elb3 7.elb3 = 4.te8*3 6.848*1 4.768 3 8.9ept as tH 9.48td 9.Mbe = l.938-4 S.t$t*3 3.10t+4 4. 30>3 as-tee 5.4e8+4 S.4ebl
- 9.348*1 1.088+4 3.ME*4 3.986*3 art 41 1.87bt 1.08t4
- 6.308-5 3.978 4 4.?$3*3 4.97b3 to-M3 3.98b3 3.3Obl = 1.988 6 1.ll8e3 6.93bt 1.MS-3
- Le-tee $.Sebt 3.008*3 -
- 1.688+l 8.488 4 8.198 1 l So-M3 1.030 4 3.77bt = = 0.338*3 8.988+4 9.Mb3 erl41 3.77B*4 9.678 4 .
- 8.388*3 $.178*$ 3. M be 1.998*3 to M3 3.988+3 1.938*3 = 6.ute t 1.Meel 4.97p4 3.Staet erM4 3.198*4 1.31>4 = S.388*5 9.84t+6 1.4ebl 1.768*l pr-M3 1.408 4 5.Mb3
- 9.9?b3 4.3Mel 3.73b4 6.9ebt fr.M4 4.798 4 1.058 4 = 6.138 3 9.61t*3 4.388*3 3.4tb3 apM7 7.9484 4.13b3
- 3.tSt+3 3.33tel 3.132 4 8.setet #
ep-tet t.3ept 9.838 4 = = 3.Mbe 3.548 4 3.13Det my-339 3.?ts*3 3.33b1 . 6.63bt 8.9M.4 3.495 4 s estet
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- tesa area /yr ser ecuen) su where Pags 7.0-22 swum- Some une .Terrou Saeer see8 ebsu t. Seer . .. 5-3 . *' 3.4M*3 7.68t*3 7.63bt 7.6 2 3 7.688 3 f.4M*3
- 1 ble 3.43bl 7.M8*4 7.3484 . 7.M8+4 7.M84 7.M8+4 7.M8+4
, . es-?4 ' 3.k8+4 3.M8+4 3.Mbe 3.94be 3.M84 3.Mbe 3.M'4
- , P43 1.?t8+te 1.065+9 . . . l.93t+9 4.4e8 8 Ce41 . . .1.798 4. 4.3ebl 3.888 4 7.308 4 3.848 4 es M no-M
=
8.40t+4 4.333-3 3.90t+4 S.MS-3
=
3.l?8*7 1.408 4 l.358 1 7.958 4 .
. Fe-SS 3.C18+1 1.738 7 . . 9.478+4 9.958+4 4.M8+4 fe 59 3.9Bb7 7.00b? . . 1.9%D7 3.3M+8 3.488e?
, , co 57 . 1.38E4 . . . 3.388a7 3.13See
. 4 as M = 4.73t+4 . . . 9.578*f 1.eM*?
Co-te . 1.Nt* ? . . . 3.esb8 3.43b7 5443 4.?M+9 4.M8+8 . . . 9.73bt 3.368+8 5145 3.MS-1 4.415-3 . . . 1.338+8 3.1 5 3 ende
- 3.45t+4 . 6.888+4 . '3.968+4 1.138 4 3 4$ t.3?bt 4.Ms+9 . 3.93t+9 . 3.788*9 t 978+9 to 49 .. = . . . . .
Br-83 . . . . .. 3.73b7 3.3Sbt Br-83 . . . . . . 1.4 5 1 ' 1.038 t
. ar-e4 . . . . . . .
Sr-es . . . . . . . RPM = 3.998+9 . . . 3.118+8 1.318+9 spe8 . . . . . . .
- O>49 . . . . . . .
St-49 1.468e9 . . . . 3.338 4 4.Mbf St-te 4.M t+10 . . . . 1.SSb9 1.158+10 3r-91 3. tM*4 = . . . l.4 5 5 1.378e3 3r-93 4.8 W 1 . = . . 9.M tet 3.118-3 F9e 7.07b t - . . . 7.letel 1.tebo ' y-9te . . . . . . . 1-91 8.008 4 . . . . 4.738 4 8.30t*3 bla 5.438 5 . . . . 9.498 1 t.8884 bt3 3.3M 1 . . . . 7.398,3 4.4M4 - 3r41 9.44b 3 3.838 3 . 4.743*3 . 9.638*5 3.058*3 St 97 4.34bt 8.595 3 . 1.NE t . 3.ME4 3.9H-3 e>95 8.3M*4 4.554 . 4.Mbe . 3.798*8 3.478+4 { e>97 .' . . . . 5.478-9 . me-99 . 3.13t+7 . .$.73t*7 . S.858*7 4.00be Te 90s 3.358+8 9.19be . 1.488*3 4.30E+e 5.ubs 1.178*3 fe let . . . . . . . Sw=t03 1.e38+3 . . 3.89bl
- 9.19E+1 4.39b3 as-tal 8.578 4 . . 1.118 3 . 5.345-1 3.38E4 be-SM 3.Mbe . . 3.Mt+4 . 1.338+4 3.58b3
. = . .
me tale . . . ny1M = . . . . . As-tles 5.8Me? $.3tb7 = 1.Mt+8 . 3.30ble 3.Seb? s>134 3.578+7 4.Mt*5 4.34t+4 . 3.00t+7 f.Ht+8 1.93E*? Sel38 3.htet 3.38tel 3.088 4 - 1. Mb? 3.35t*8 4.Mbe
- To-tale 1.4Het 8.908 4 4.90t+4 6.4 m 7 . 4.leb7 7 188+4 To-Itte 4.MS*7 t.M8*7 1.178*7 1.Mt+8 . l.Mt+8 8.J884 fe st? 6.73b3 3513e3 4.Mb3 3.748*3 . 5.308 4 1.4Sb3 fe-itte 6.Mb7 3.358 7 3.08b7 3.53bt - . 3.Mb8 9.378+4 te-139 .
fe-tHe 3.6tbl 1.7?b l 3.80$*5 9.798 4 . l.70b7 1.4?bl To-ut . . . To-03 3.39t*4 1.S$be t.f tS4 t.498+f . 1.32b? 1.4the htte 4.Mbl 1.ht4 1.etb8 1.M8+4 . l.ee84 4.tebl 3 ut 3.M8+8 4.ht*8 f.3mit f.3784 . 1.13h4 3.438+4 but t. Mbt 4.378-1 1.ubt 6.978-1 . 8.338 3 1.538 1 3 u3 3.97b6 6.985 4 1.0t#9 1.3ebt . 6.388 4 3.988 4 3.tM . . . . 3 ul 1.39b4 3.438+4 3.408+4 S.8M+4 . 4.908 4 1.k84 80 tM S.468*9 1. M 8+le
- 4.SSE*9 1.Mb9 3.3S8+4 1.tetete es tM 3.4tb8 1.93bt . 5.?4E+8 ?.875*7 1.178+8 7.438+e es-137 f.38bt f.elbte = 3.4M+9 1. M8+9 1.968 4 6.618+9 en-us . . .
So-1M 4.188-4 . . . . 8.348 8 6.388-9 to-Me 3.Hbf 3.388 4 . 1.tShe 1.9384 S.Mbf 1.M84 . to-ut . . go-M3 . . . . . la-Me ' 4.495eo 3.M880 . . . 1.M8+l S.97bt
. . . . . 3.e38-4 .
G i Es-143 Co-M1 4.M84 3.3?hl . l.838*3 . 1. Stb? 3.?tb3 So-MS 4.19b1 3.498 4 . t.M8*1 = 1.Mte4 3.43be es-M4 3.908+l 1.988 5 - 8.878*4 . 1.3tb8 1.900 4 h=t43 9.888 3 6.3?8+1 . 3.088 1 . 6.Mbl 7.tebe pr.lu .
'.e >Mt 9.435 1 . 1.0eb3 . 6.3?> t . S.33bl 6.938+4 4 87 6.948*3 S.48b3 * . ., 1.008 4 t.988*3
+p3M 3.M8et 3.Sebt . 1.13be #* 7.398+4 1.98bt
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' .- P 33 3.11 tete 1.958*9 = = . 3.6Sbt t.338*9 ,'
Cr-ll . . 3.748*4 1.198*4 7.138*4 4.408 4 5.008 4 . me-le . t.4etof . 4.178*6 . 3.J7bt 3.788*4 me-M = 1.598 3 . 9.908 3 . 4.hp1 1.338 4 Fe-ll 4.elbf 3.16tet . . 3. Sept 1.378*7 f.M84 Fe-59 $.308*1 1.318*8 - = 3.83bf 3.gtb8 4.Mb7 es-87 - 3.388*4 , .- . = 4.198+f 3.M84
- co-88 = 1.968*6 = . -
- 4. tab 8 4.838+f Co-te . 3.708*7 . . = 3.63b8 4.368+f el-63 1.18telt 8.388*8 - = . 1.338*4 4.018*8 el-65 6.?tbl 4.M8 3 - = . 4.fe84 3.MS-3 es-M = 4.398 4 . 1.09bl . 3.338 4 3 838*4 -,
So-el 3.118*9 f.318*9 =- 4.448 9 -
- 3.te> 9 3.418*9 3e-69 . . . . .. . .
'st-33 . . . . . . l.Mb? '
St-33 . . . . . . 1.918-t ge-se . . . . . . . . De-06 . . - . . . - Bute
- 4.7M*9 . . . F.eeb8 3.338*9 36-88 ., . = . . . =
gue9 . . = = = = . St-89 3.478*9 . = = = 3.14be F.Mb7 St-90 6.618*10 = . . = 1.868*9 1.49 tete a St-91 5.768*4 = . . . 3.618*l 3.398*3 8t-93 8.955 1 * = = . 3.3Bbt 3.818 3 1-90 1. Nt* 3 . . - . 1.878*4 3.908*0 1 9te * = . . f . - T-91 1.lebe . . . . 6.488*6 4.MR*3 bt3 1.008 4 - = = = 3.758 4 3.905 4 T-93 4.308-1 . . = = t.318*4 1.188 3 8t 95 1.668*3 l.Mtet . 7.678*3 . 1.Seb6 3.898 3 3r-97 f.?lt 1 1.H8-1 = 3.338 1
- 4.168*4 7.06b3
~
sb-96 t.4tb l 7. set *4 . 7.578*4 - 3.Mbe 4.3084 ( e>97 . . . . = 4.M84 . me-99 - 4.Mt*7 - 1.M8*8 = 8. ME* f 8.698 6 te-9te S.M8*4 1.578*1 = 3.ht*3 8.7M4 1.038 4 3.M8+3 ft-101 = = = . . *
- So-le) 1.01t*3 . . 6.40b 3 . 1.libi f.?l8*3 Se-tel 1.lfb) . . 1.978 3 = 1.MS*0 6.8She ko 106 3.76be . . 7.338 4 = 1.088*6 4.738*3 1 ab-tone . . . . - . = l Ab=106 = = = . . *
- I es-stes 9.nb7 9.118*1 = 1.748*4 = 3.Mble S.MS*?
89134 4.598 7 8.4B+l 1.048*l . 4.918e? 9.888+8 1.798*7 Settl 3.658*7 3.99tel 3.498*4 . 3.318+f 3.048*4 8. MS*6 Ts-tale 3.008+f 1.088+f 8.39b6 - = 8.M bt 4.328*6 , fe-131. 8.448*7 3.998+f 3.018+f 3.438*8 - 3.teb8 1.88b? 1e-137 1. Mbt 4.41b3 0.598*3 l.Mbl . 9.618 4 3.04t** Te-tite 1.198*8 4.teb? 3.lfb7 4.63b8 - 4.tStes 1.788*1. fe-139 - = = t.67b9 = 3.184-9 . To-tale 6.lfbl 3.1Sbt 4.74bl 3.'39pe
. 3.83tel 3.630*$
to.131 = . = = . . . fe-133 4.388 4 3.?tt+6 3.MS*4 3.0e8*1 . e 8.98b? 3.988*4 3-13e f.498+l 3.t?b6 1.778*8 3.ht 6 . 1.678 4 8.M8+l 3 131 5.38b4 f.538*0 3.388*11 1.388*9 = t.498+4 4.ME*8 3 133 3.9pb1 1.598 1 3.96bt 1.30 tee . 3.3tbt 3.738-1 3-133 7.Mb6 l.338*7 1.73b9 3.16bf = 9.308 4 , 8.788*6 3-tM = = . . . . . bill 3.478*4 6.388 4 4.088+6 1.00tel . f.el8*4 3.388*4 Co-tM 9.818*9 3.318*10
- 7. Mbt 3.008*9 3.8784 f.9Phle -
Co tM 4.465 8 1.?l8*9
- 9.53t*8 1.90t*4 9.4 tb8 1.teb9 Se=t37 1.M8*10 1.f88* te .. 4.06b9 3.3Sb9 3.$3be 6.30t*9 ge tm . . . . . . ..
Se-t M 8.09>0 . * * = - f.?lt-? 3.93b9
- Se-tee 4.99&T 8.998 4
- 8.038 4 4.008*4 7.49pf S.13h4 se-le t . . . . . . .
ge-43 . . . . . . . Sie lee 8 0684 3.Nbe . . . 3.3?bl f.elbe { 1a-143 . . . . - 3.330-7 . Co-let 8.8?b3 5.98h3 . 3.?tt*3 = t.698+f 6.8tb3 8e 143 f.ettet $.888+4
- 3.litet . 1.Mid 6.388*4
. Co-144 6.lsbl 3.73D$ . 1.638+l . l.MS*4 3.M84 Pr=143 3.938** 1.tFb3 . 6.7Fbt . 9.ettel 1.438+t Pr-144 . . . . . . .
ad 147 1.018*3 1.978*3
- 1.168*3 . 7.918*$ t.188+l W-107 1.308*4 9.febl * * = 3.6tbe 3.43b3 ar-3M 6.99b8 4.998 1
- 3.978 4 . 1.06bs 3. Mbt i
~
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- (et a areWyt per seusee) for etters anellee asse Meet ThrteM EMesy See8 . SI4L1 .T.8ety 34 . 9.978 4 1.878 4 1.67h3 1.578+3 1.878 4 1.578*3 s 14 .9.668 4 3.39bt 3.39bl 3.395el 3.39pl . 3.39Dl 3.398*5
( se.34 9.3M*4 9.33be 9.33b6 9.33be 9.3384 9.3M+4 9.338 4 P-33 7.778*10 3.Mb9 . . - 3.tsb9 3.ee8*9 Ct-St ** . 5.ube 1.554+4 1.eM*$ 8.418+6 1.038*5 me-se . 3.e98*7 . S.878 4 . 1.768*7 5.988+4 he-54 . 1.3tb3 . 1.$48 3 = 1.tebe 3.968 3 h SS 1.138+8 5.9M*7 * . 3.llbf 1.988+7 1.448e? Fe-39 1.308+8 1.98b8 . . l.658*7 3.tM4 9.?tb7 Co 57 . 3.M84 * . . 3.M8*? 7.775 4 co-88 . 1.318*7 . . . 7.teb ? 3.738*7 Co-te . 6.33pf a = '. 3.398+4 1.378+8 N43 3. M 8 ele 1.998*9 . . 8. 1.478+4 1.018+9 M el 1.68+e S.MS-1 . . . 1.91b1 9.ttb3 es-64 - 7.6S&4
- 1.838*l = 3.948 4 4.94b4 Se45 4.1H+9 1.tek te = 6.M8*9 .. ' t.9 2 +9' 4.888+9 Se49 .. . . . . 3.148-9 .
3r.83 . . . . = = - 1.168+8 gr.83 . . . . . . 4.698-1 3g-34 . . . . . . . ge-el . . . . . . . Bete . 8.77bt . . . ,S.ube 8.39pt . BHM . . . . . . gett . . . . . . . 8s-89 4.638*9 . . . . 3.M8+8 1.098 4 St-90 1.138*11 . . . . 1.618*9 3.83b le St-91 1.418+l - = . . 3.13bl $.338*3 Br-93 3.198+e . . . . 4. H8* 1 8.48 3 T-90 3.338*3 . . . - 9.988*l 8.6tbo g-9te . . . . . . . 1 91 3.918+4 . . . . 6.318 4 1.MS*3 7-93 3.448 4 * * . . 7.108+0 7.8H4 1-93 1.M84 . . . . 1.573*4 3.908 3 8t 95 . 3.848*3 8.458*! . 1.218*3 . 8.818*l 7.538 3 Staff 1.898 4 3.738 1 . 3.918*1 . 4.1M4 1.6tbt
. . Obtl 3.18bl 1.348*$ . 1.Mbl - 3.398 4 8.M84 obt? = . . . . 1.458 4 .
Ite 99 . 8.398*7 . 1.778*8 . 6.MD7 3.el8*7 fe 99e 1.398*1 3.M8e t . 3.68b3 1.398*1 1.MI+4 4.308+3 j
. . . . .. . . i te let 3 103 4.398*3 . . 1.088+4 . 1.158*l 1.elb3 to-tal 3.838 3 . . 3.M8-3 - 3.498 4 1.79F3 Se-106 9.348+4 . . 1.3Sbl . 1.448 4 1.168 4 30-103e -
R>les * *
- AS 118e 3.098+8 1.4t8+4 . 3.638*8 . 1.688*19 1.138+8 Se134 1.09b8 1.418*8 3.4e8+l . 6.038e? 8.798*8 3.818*7 l 8etal 8.748*? 9.418*4 8 M84 . 4.858*7 3.888+8 1.888*7 Te-13Se 7.M8*7 3.908*7 3.078*7 . . 7.138*? 9.848 4 l Se-tate 3.08b8 8.80847 4.978*7 5.9M4 . 1.48bt 3.478*7 l te-137 3.068*3 8.3tb3 3.133*3 8.718*3 * . 1.3e8+l 4.Mb3 fe-139e 3.738+8 7.etb7 8.788*7 8.80bt . 3.333 4 4.33b7 fe-139 . . . 2.878-9 . 6.138-8 -
Te-tate 1.088 4 S.5Mel 1.M84 8.388 4 . 3.pb7 8.898+l Te-131
~
Teat 33 1.038*7 4.538 4 4.863*6 4.3e07 = 4.S$b7 S.448+4 3 1M t.7684 3.94be 3.908*8 S.398 4 . 1.M84 1.838 4 ) 3-131 1.308*9 1.3tb9 4.M8+tt 3.198*9 - 1.178+8 7.448 8 3*l33 6.MS t 1.Mbe 8.068*1 1.938*0 . 1.488*G S.008 1 3-133 1.Mb7 3.188e? 6.04b9 3.638*7 . 8.778 4 8.338 4
. . . .. . . . l 3-SM I 3 135 S.M84 1.eMel 9.308+4 1.6thl . 8.008+4 6.978+4 '
es-tM 3. M b le 3.?ttelt . 1.158*10 4.198*9 8.00b8 7.83b9 Co-1M t.sebt 3.748+9 = 1.47b9 3.198 8 9.708*7 1.79b9 Co-137 3.338*10 3.etble - 1.ettele 3.638+9 1.9M 4 4.888+9 es-tst . . . . i as-tM 3.14b7 . . . . . t.338-8 6.19b9 to-Me 3.17b8 1.03bl . 3.348+4 8.138+4 8.Mb7 . 6.848a6 as-tat . . go M3 . . . le-MG 1.93bt 6.748 4 . *
- t.48bl 3 37>0 le-M3 . . . . . 3.$184 .
es-Mt 3.198 4 1.098*4 . 4.788 1 . 1.Mb7 1.43843 es-M3 1.898*3 1.038*$ = 4.398el . 1.9e84 1.48bt Os-H4 1.638 4 S.99bl . 3.83bl
- 1.3M*8 S.848+4 Pr.M3
- 23h.i 3.t??*3
. 1.178*3 . 3. Sept S.9801 pr.tu . . . . . . .
abl47 4.&l8*3 3.4e8*3 . 1.9sb3 . l.7tb l 3.798*1 IP 107 3.918+4 1.738+4 . . . 3.438+6 7.738*3 SP-339 1.738*1 1.33be . 3.578 4 . 9.14be 8.tebt ! l
, an*e,= -_ m eMe.e,So.e ruler.4ar wr ODCM-7.0 es Revilian 6 terever W e arearrr for sese, se acue, s.3 eens'*
ul ser ans C.M Pags 7.0-25 1 auslide Goes &Acet Shrteld Eidser Emes' SI-ILI f.8ser
, 63 * .
3.34b t 3.388*3 3.34b3 haeba 3.38b3 3.388*3 C-14 3.H84 6.898+l 4.898+l 6.898*$ 6.89bl 4.898*8 4.898* $ - ( as-34 P43 1.618*? 1.618*7 1.618*7 1.80bts 9.438*9 = 1.618*7 1.618*7 1.818*7 f.648*7
'. . 3.178+e 4.3tbt St44 .
- 1.elbl 3.388+4. 3.el8+l 4 7184 1.618*$
80s M . 3.898 7 - 8.63b4 . 1.438*7 8.838 4 As-M = 3.318 3 . 3.fe8 3 . 3.918*G l.538 3
- j Fe-ll 1.3M*4 8.738*7 . . 4.378*7 1.198*7 3.338*7 '
re-59 3.388*8 3.93b8 - . l. Mb4 1.88b8 9.S$8*4 ! So l?
- 8.988 4 . . .
3.888*7 1.M8*7
~ co-te . 3.438*7 . . . 6.888*7 6.868*7 Co-ee . 8.018 7 . . .. 3.908*8 3.888*8 AM3 3.49h te 3.168*9 . . . 1.078+4 t.318*9 54 4 l 3.388+e 3.978 1 . . . 3.e38 5 1.Sibt 81s- 4 4 . 1.88bl - 3.178 5 = 5.8Sb6 8.698 4 -
3e-48 S.S$8*9 1.Sesete . 9.338*9 . 1.6tbte 8.70bt seat . . . . . 7 M8-9 - 3r=83 . . . . . . t.968*8 l De-03 . . . . . . 9.9M-1
= 44 . . . . .
. . i l
se-el . .- . . . .. . a>M = 3.338 94 . . . 8.69b8 f.98bte mest . . . . . . . 3e-09 . . . . . . . . St=89 1.348*14 . . . . 3.598*8 3.618+8 i St-98 1.338 11 . . . . 1.53b9 3.tetete St-91 3.96bl . . . . 3.488+l 1.868 4 St 93 4.658ee . . .. .- S.ettet 1.738*t b98 6.088*3 . .* . . 9.398*$ 1.83bt F9te . j bt1 7.333 4 . . . . 8.36be 1.958*3
- b93 $.338-4 . . . . 9.978 4 1.478-$
- F93 3.368 4 . = . . 1.768*4 4.138-3 It-95 4.83b3 1.Mb 3 = 1.79b3 . 8.388*$ 1.188*3 8t-97 3.99be 6.8581 - 6.9tbt . 4.378*4 3.185-1 skti 5.9&bl 3.448*l = 1.75bl . 3.e68*8 1.418+l skt? . . . . . 3.708 4 .
E j me 99 . 3.138*8 . 3.1?b8 = 4.988*7 4.138*7 1 te-90s 2.698*1 $. lib t . l.978*3 3.908*1 1.618 4 7.1M*3 te.181 . . . . . . . ) Se-le3 . 0.698*3 . = 1.848*4 = t.M8+l 3.918*3 i Su-tel 8.MS.3 . . l.938 3 . 3.3tbe 3.718 3 So-tM 1.98848 * - 3.3Sbl
- 1.44be 3.3884 34-183e * . . . * .
- 2h-104 . . . . . . .
As-ties 3.M8*8 3.83b8 . 4.038*8 . 1.448*10 1.868*8 l Sette 3.09b8 3.e88 4 $.948+l . 1.318+8 6.MS*8 6.498 7 ! Skt35 1.498*8 f.488*4 f.87bl . 9.388*7 1.99b8 3.87b7 Te=ttle 1.518*8 S.e48*7 $.078*1 . . 7.188*7 3.068*7 Te-tafe 4.348+4 1.408*8 1.338*8 f.MS*9 . l.708*4 8. tete? j 1 fe-137 6. Set *3 3. tSD 3 $.398*3 1.898*4
- 1.368 5 1.488+l -
To-late 5.598*8 9.938+8 3.1Sbe 1.488*9 . 3.M8*4 8.638*7 I to 139 3.00bt . 1.788 9 1.188 9 = 1.Mb7 . Te-13te 3.388+4 1.M84 3.h>4 9.388+6
- 3.39b7 1.138 4 3 134 . . . . . . .
Te=t33 3. W8*? 1.048e? 1.M8*7 4.898*7 - S.868 7 9.738*4 1 bile 3.488 4 7.938 4 8.088*8 8.M8*4 . 1.708*6 3.188*4 htSt 3.738*9 3.3tbt 1.058*l3 3.758 9 . 1.958*4 1.etb9 , btlt 1.438 4 3.09p e t.358*3 3.32be . 3.348*G t.esbe 3 135 3.73b7 8.458+7 9. Mbt 6. MS*7 . 9.M84 t.Seb? bth . . 1.etbt . . . . bill 1.3tbl 3.418*$ 3.H8*7 3.698el
- 8.M 8*4 0.008 4 Co-lM . 3.658*10 6.sehts . 1.758*10 7.188 9 t.8the 4.878+9 co th 1.948*9 6.778*9 . 3.30$*9 4.788+8 8.76bt 3.198*9 to t37 S.168+te 6.828*10 . 1.638*18 6.S$bt t.888*4 4.378+9 es.138 . . . . . . .
Se-139 4.958 0 . . * . 3.088-6 1.338-8 8e-144 3.498*8 3.418*l
- S.738 4 1.48bs S.93b7 1.M8*7 8.-141 . . . . . *
- So-M3 Le Me 4.elpt 1.988*1 . . . 1.07b l 4.898ee l Le 143 . . . . . $.3184 .
ee 141 4.338 4 3.MB4 . 8.ttb3 . 1.378 7 3.ttbl Co-143 4.0e8*3 3.6Sbl = 7.738+t . 1.ll8*6 3.etb1 Se-M4 3.H84 9.63bl . 3.8Spl . 1.3M*8 1.3ebl . Pr 143 1.498*3 B.99b3 = 3.888 3 . 7.898*5 7.4ttet pr-Me . . . . . . . se 147 8.838*3 9.M8*3 - 3.498*3 . S.hbt $.15pt w-187 4.138+4 4.Mbe . . . 3.3e84 f.4784 ar 339 3.M8*1 3.388 4
- 4.698 4 . 9.4e84 1.048 4 l l
l I
F ' ODCM40 l
- 8.e.crenede-asstsemerSm.eruta..osar~ R:visi:n 6 >
tweerre we ecuan tw 84 ses c-14 Page 7.0-26 (e8 e aremryr por s24/see) ist othere asslide amme - &ber Terrole Sneer lame 8bl&! T.8mer b 8-3 3.368*3 3.368*3 3.35b3 3.3Sb3 3.358*3 3.358*3 IL C-te 3.338*5 4. Nt*4 4.668+4 4.Mt+4 6.468+4 6.M8+4 4.Mt+4
'( , me 34 1.M8 3 1.Mb3 1.048-3 1.048 3 8.84b3 1.048-3 1.04t4 P43 4.6Sbt 3.89b4 . . . . S.33b4 1.eue+8 St.51 = ='
4.338 4 1.M8* 3 9.Mid 1.788 4 f.87b3 me-M . 9.tibe = 3.73b6 = 3.808*7 f.788 4 me.84 . . .. . . . . Fe.tl 3.98t+8 3.ett+8 . . 1.138+8 1.16t*4 4.738*7 Fe-H 3.47t*8 6.37b8 = ,. 1.758+4 3.00bt 3.448 4 Co l' . .- $.M8+4 . = . l.438+8 9.37be es-SS . 1.83b7 '. . . 3.708*8 '4.108 7 Co-te . 7.538*7 . . . 1.4tb9 1.448e4
'8443 1.89b H 1.31bt . . . 3.73b8 4.338*4 34 45 . . . . . . . ;
area = 3.918 7 . 7.458 7 . 3.138-6-1.39bt se48 3.568+8 1.188+9 . 7.878+8 . 7.13be 5.t&S 3e-49 . . . . . . .- St-83 . . . . . 1.MS*3 1.368+3 tr-83 . . . . . . .
- St-44 . . . . . . .-
pg.88 . . - . . . .. St-M . 4.8?be = = = 9.60t+7 3.3?be 86-80 = . . . . . Be-89 =. .
= = . . $
St=09 3.018+4 = . . 4.ht+7 8.668 4 St-to 1.348+10 . . . - 3.l b e 3.96t*9 St 91 . . . . . 1.38b9 . St-93 . = . . . . . i T-90 1.07b3 - * . . 1.138 4 3.Mbe b9te . . . . . . . F91 1. tab 6 . . . . 6.Mb8 3.038 4
- r-93 . . . . - . .
F93 . . . . . 3.888 7 = 8t 95 9.Mt+4 6.Mtel
- 9.44bl a t.918*9 4.09tel 6 tr-97 1.038 8 3.698-4 = $.588 4 = 1.148+0 9.698-4 sb95 3.398+6 1.30t+6 = 1.348 4 '. 7.788*9 4.Mbl I put; . . . . . . .
me-99 . 1.99bl - 3.46tel = 3.$3bl 3.97b4 Te-99m = . . . = . . te-set = . . . .. . . < 8e-103
- t.Mtet . . 4.938*8 . 1.33t+10 4.56t*7 8.-105 * * = . . . .
Se194 3.80t+9 . . l.40t*9 *
. 1.818+11 3.54b8 Sh 18h * * . . . . .
Sn-806 . . . = = . *. ' es 118e 6.69b4 4.19be . 1.32b7 . 3.638+9 3.678 4 SW134 1.90t+7 3.74t*5 4. tope . 1.lete? 8.638+8 7.85t+4 ' Setti 1.9tte? 3.138+8 4. M 8+ 4 . 1.478s7 3.188*8 4. ht+4 fe-13As 3.59t*8 1.Mb8 t.esb8 8.M8+9 . 1.43bt 4.8tb7 fe-137e 1.138*9 3.998*6 3.8148 4.llbf . 3.748+9 1.Mb8 , to-137 . . . 1.098 9 = 3.100 4 . te 139m t.14b9 4.378+8 3.98b e 4.778+9 . S.76b9 1.018+8 Te 139 . . . . . . . fe tale 4.818*3 3.3tb3 3.588*3 3.Mp3 . 3.198 4 1.Mb3 fe-131 . . . . . . . Te 133 f.4et+6 9.etbl 1.00t+6 8.7M4 . 4.998 7 8.Sthl 3-130 3.365 4 4.948 4 S.848-4 t.sebt . l.988 4 3.748-4 ht31 1.088*7 1.848*7 8.958*9 3.Mb7 . . 4.878 4 8.838 4 3-133 . . . . . . . 3-133 4.308 l 7.478 1 1.teb3 1.Sebe . 6.73bt 3.3ebt
. 3.jp . . . . . . . .
3 138 . . .. . . . . Co-th 4.578+8 1. ht*9 . l.068+8 1.48t+8 3.748+f 1.38bt Co tM t.tSte? 4.67t*7 . 3.eeb? 3.348 4 S.308 4 3.ME*7
, co 137 8.73b8 1.198+9 - 4.068+8 1.35t+8 3.318*7 7.018+8 .
co-tu . . . . . . . , to.139 . . .- . . . . ) Se 144 3.88b7 3.6tb4
- l.838 4 3.97b4 S.988+7 1.08b6 se 141 . . . . . . .
an 143 . . . . . . . 34 148 3.409-3 1.8tb3 . . = 1.33b3 4.79b) . t ie-oa . . . . . . = en-le t t.4ebe 9.488 3 . 4.4e8 3 . 3.63b7 1.es>3 )' 9e 843 3.998 3 1.llbt - 4.Seb3 . l.788+2 1.7t>3 ee-M4 8.44t+4 4.09bl . 3.418+l . 4.938+4 7.834 4 1 Pr-us 3.138+4 8.948*3 . 4.93b3 = 9.33b? 1.068*3 pr-us . . . . . . . ee-u? 7.88b3 4.188 3 . 4.7e8 3 . l.938*7 4.9e8,3 0k187 3.168-3 t.018 3 . . . 8.933 8 6.328 3 ar-339 3.ME 1 3.liba . 7.048 3 . l.168 3 t.39t 2
u_ atou,L F v 00CM-WO so, e,eee e eH eMMer toes reet . . muun
- g. R :visi::n 6 ,
(aren/rf ses sil/o8) for b3 and ble P:ge 7 0-27 l (a a arentri per act/ees) get ettees s aucIlde ,tene' 3 dest Ttfread R& deer Seas SI-113 f.teer " B3 . l.96t*3 1.ME*3 1.ME* 3 1.Mt*3 t.96t*3 t .ME* 3 (. CH he 34 3.Stbl S.438+4 S.638 4 S.632*4 5.63t*4 S.68t*4 5.638*4 1.478 3 1.4?b3 1.478 3 1.478 3 1.4?b3 1.47h3 1.478 3 P-33 3.938*9 3.448*8 . . . 3.30t*8 1.538*e Ct-St . . 3.MS* 3 1.34 b3 8.078*3 9.Mael 5.658*3 no M . 6.Mbe . 3.083+6 . 1.4M*f 1.M8*4 . no-M . . . . . . . Fe-ll 3.388*8 1.69E*4 . . 1.07E*8 f.Sete? 3.938 7 Fe-59 3.tM*8 4.90t*4 . . 1.578*4 1.let*9 1.938*8 ee-57 . 4.53be . . . 8.488*7 7.99b6 . ee 58 . l.4tte? . . . 1.M8+t 3.35t*1 So-te . l.83te? . . . 7.tob8 f.Sthe 51-43 1.838*19 1.078+9 . . .. 1.788+8 8.988+4 i 51-45 . . . . ~. . . < co-44 . 3.418-7 . 4.108 7 - 1.878 A 1.138-7 . l Se45 3.Seb8 4.698*8 . 5.Mb8 . 3.688*4 4.068+8 to-49 . . . . .- . . tr-43 . . . . . . 9.988*3 St-83 ' . . . *. . . . se-se . . . . . . . k45 . . . . . . . Sk84 . 4.Mt*8 . ,
. . 6.015*7 1.918*8 p>M . . . . . . .
3939 . . . . . . Sr-89 3.64 8 . . . .. 3.038 7 f.39t*4 St-90 8.05t*9 . . . . 3.Mt*4 1.998 9 Sr-91 . . - . . 1.let-9 . St-93 . . . . . . . f-90 8.98Eet . . . . 7.488 5 3.438*4 T-91e ., . . . . . . T-91 9.letel . . . . 3.93b a 3.56t*4 f 93 . . . . .. . . T 93 . . . . . 1.698-7 . 3r.91 1.11t*6 4.76t*$ . 6.9995 . 1.1899 3.37tel 3r-97 1.53b5 3.638 4 . 4.56 1-4 . 4.185 1 1.39pe l De9l 1.?9be 9.Mbl
- 9.448*l . 4.38bt $.478*$
36-97 . . . . . . . me-99 - 8.98t*4 . 3.ht*l . 1.618*$ t.7184 fe.99u . . . . . . . fe-101 . * . . . . . Re-103 8.60b7 . . 3.9308 . 7.10bt 3.ME*7 a.-105 . . . . . . . Re-106 3. Mbt . . 4.558 9 . 1.138*11 3.978+4 an-tene . . . . . . . St-tM = . . . . . . es 190s S.ht*4 4.79b6 . 9.H8+4 . 1.388*9 3.918+4 Sel34 1.438*7 3.90bl 3.678+4 . 1.4tb7 3.Mb8 4.3tb4 Skt35 1.848 7 1.712+5 9.49E*4 . l.37t*7 1.33t+8 3.Mbe fe-tale 3.038*8 1.99E*6 8.47bt . . 8.M3*8 4.96b7 fe-tale 9.41t*8 3.Mba 3.34b8 3.83b9 . 3.358*9 1.13t+4 To.137 . . * . . 1.758 0
- Te-13te 9.88ba 3.968*8 3.99E*8 4.etbt . 3.608 9 t.33be Te=t39 . . . . . . . ,
Te.tlle 3.768*3 1.00t*3 3.71t*3 1.088*3 . 1.454*4 1.308 3 l Te-131 . . . . . . . Te-133 1.16b6 f.Mbl 7.Mbl 6.97s*4* . 3.teb? 6.048+l 3*190 1.098-4 S.448-4 4.478 4 8.448 4 = 4.318 4 3.198 4 1 3-138 8.988+4 1.3&b7 3.Mb9 3. M8+ f . 3.48be 4.73be g.tM . . . . . . . . l 3-133 3.608 t 4.teht 4.ltbt 1.878*e = 4.418-1 1.NE
- 3-tM . . . . . . .
3 135 . . . . . . . ee tM $.33t+4 1.33bt . 3.9tb8 1.498*8 1.53te? $.798 4 es-lM 9.338+4 3.6&te? . t.97b7 3.118*6 3.938*4 8.448+f ee 137 7.M8*8 9.688+8 . 3.388+8 1.378*8 1.3?b7 3.Mbe ee.138 . . . . . . . Re-1H . . . . . . . as-Me 3.303*f 3.91>4 . 9.088 3 t.968+4 3.67>1 1.838*4 i brut . . . . . . . ; I 3.-M3 .
. 8. M +3 3.8,8-3
( ,4 , 3. 68 3 1.4M.8 . . 1 u.M3 i se-141 1.14b4 f.e6b3 . 3.988*3 .. 3.388*7 9.838*3 i go.us t.Ms 3 1.388*t . 3.74b3 . 3.058 3 1.438-3 I g,.ug 3.338+4 S.888+ 5 . 3.M8+l . 3.09b8 6.608 4 i 9e.u) g.798+4 f.ttb3
- 4.94b3 . S.988*1 8.938*3 Dr.M4 4.348*3 4.798 3 . 3.988*3 . 4.068 3 e4-84,1 g.gg t.gt8 3 1.488 3 . . . 3.668'e?
3.98b 5.178 3 l sp-339 3.338 1 3.168 3 . 6.618 3 . 3.39b3 1.178 3 - i
an* crese ene. sees pe Me, ten ruiere . casta ODCM-7.0 R vibirn 6-haserre see sesM) ser s 3 ene C*" Page 7.0-28 W s atentyt pge pC1/ees) for othere ass 184e Sean & Joes Thyroid Siemer Ames SI-8&B T.tesy 53 = 3.23 1.948*3 3.Ma*3 3.ht* 3 3.M8*3 3.ME*3 C 14 5.395e5 3.06tel 1.068*$ t.e6tel 1.M8el 1.ME* 3 1.M8* 3 (:- so se bla 3 kb3 3.h3 3.M83 3.Mg.3 3.Ms 3 3.Ma4 3.ht 3 3.415*9 3.478*8 - . . 3.9&be 3.M8*8 Cr-S1 . . 4.99g43 1.M843 8.93b3 4.67bl 4.81b3
' me-M . 7.998+6 . 3.M B+4 . 6.708*4 3.tM*4 me M . . . . . . .
Fe-SS 4.l?t+8 3.43t+8. . . 1.378+8 4.498*7 7.59847 Fe $9 3.74b8 4.138*8 . -. 1.77be 4.3?b8 3.068 8 es 87 . l.938+6 . . . 4.8M*7 t.808*7 es SS . 1.658*7 . . . 9.88b7 8.M8*7 ao 40 = 6.9M*7 . .
.. 3. M b e 3.teB+8 5143 3.918ett 1.ht . . . 1.068+8 9.918*8 mi-el . . . . . . .
Co-44 . 3.348 4 . 7.838 7 . 1.538-6 1.968 7 Se-el 3. fib 8 1.e0E*9
- 4.308*8 = 1.76b8 4.33b8 So-69 . . . . . . .
. Sr.83 . . . . . . 1.948*3 -
se-83 . . . . . . . grace . . . . . . . k4l . . . . . . . 36-M . l.768*8 . . . 3.?tb7 3.5484 36-88 . . . . . . . 3>09 . . . . . . . St=89 ,4.838*8 . . . . l.hte? 1.388*7 tr-90 1.e45*10 * . . . 1.40b8 3.M8*9 er-91 . . . . . 1.918-9 . Stata . . ._ . . . . 1-90 1.705 3 . . . . 4.848*$ 4.llE*e b9te . . . . . . ,. , T 91 1.81t+6 . . . . 3.4 tb8 4.838 4
. 1 93 . . . . . . .
1 93 . . . . . 1.568-7 . St 91 3.Mt*6 5.09t*5 . 8.4Mel - 6.148*8 8.348*$
. 8# 97 3.848 5 4.10be . l.098 4 . 4.313 1 3.438 4 l aptl 3.99E*6 1.30t+4 = 1.13t+4 . 3.33bt 8.618*$
5 g>-97 . . . . . . . No-99 . 1.3Sb t . 3.4?bl . 1.8M*$ 3.0984 fe-99e . . . . . . . Te-198 . . .. . . . . Re-103 1.668+8 . . 3.938*8 - 4.e48*9 5.98847 to tel . . . * * . . i ae-106 4.44b t . . S.99t*9 . 6.90b le 8.548+4 at-tose . . . . . . . 36 106 . . . es Item 8.408+6 S.678*4 . 1.M8*7 . 6.75t*8 4.53t+4 Set 34 3.938*7 3.00tel 4.468+4 = 1.438*? 1.888+8 1.este? Setal 3.86te? 3.19tel 3.Mt+4 . 1.59b7 6. Set *7 8.968+4 fe-tale 5.698*8 1.M be 1.408*8 . . l.49E*8 7.598*7 fe-137e 1.77B*9 4.788*8 4.34t+8 5.e68+9 . 1.44pt 3.118*8 fe-137 = = = 1.398-9 . 1.MS-8 . Te-129e f.ette9 5.Mbe 8.83b8 5.3es*9 - 3.3es*9 3.003*8 te.139 . . . . . . . Te-tale 7.00t*3 3.43b3 4.98b3 3.M84 . 9.83b3 3.98bt 3e 131 . . . . . . . To 133 3.etb6 9.378*$ l.Sted 8.40E*4 .- 9.338+4 1.taba b130 3.395 4 4.068 4 f.Mbe 1.038-5 . 3.30E-4 3.838-4 blat 9.M8*? 1.47E*7 8.83b9 3.748*7
- t.4984 9.49E*6 3 133 . . . . . . .
3 933 4.684 3 S.Mt=1 1.538*3 1.38be = 3.5M 1 3.185-1 3-164 . . . . . . .
. 3 335 . . . . . . .
es lH 9.338+8 1.61844 . 4.69pa 1.6ebe 4.958*6 3.998*4 es-tM t.59t*7 4.3?b? . 3.3M*? 3.4?be 1.ME*6 3.esb? Os=137 9.3M*9 1.388*9
- 4.M84 1.98b8 7.99be 1.88b8 es-138 . . . . . . .
3 139 . . . . . . . 9e-144 4.39t*7 3.S$he . 1.Sthe 3.39b4 3.338*7 3.96be as 441 = . * * = = . as-M3 . . . . . . . i . Es-M8 S.4t>3 1.89b3 . * . $.37Be3 6.388-3 i se.M3 . . . . . . Co 141 3.338 4 1.ttb4 = 4.Mb3 . l.M847 1.M8*1 co-143 3.308 3 1.79bt . 7.ltb3 . 3.638 3 3.898 3 l Se 144 3.335 4 7.36 Bel . 4.838+l - 1.89t+8 1.M8el pr M3 3.39b4 1.03be . 8.9tb3 . 3.M847 1.848 4 pg.94 . . . . . . . es M7 1.178 4 9.48b3 . . S.Seb3 . t.geb? 7.M3*3 W 187 3.36b3 1.998 3 * . . , 3.79be 4.tBb3 sp 339 4. set-t 3.425 3 = 8.7M4 - '* 2.3M4 3.tabt
' Au. sesotatsse potamer,8ese Festete
- 485LT Revisi:n 6 t
!' (aren/yr got sti/ms) fee 8 3 est 8-14 Page 7.0-29 (ese aren/Fr per sta/ses) fe, einere
< Smallee asse Meer - Thyroid. 814eef Esos SI-BM T. Beer 5-3
- 3.34b3 3. nee 3 3.368 3 3.M8 3 3.M84 3.M8 3
$f ble 8.978 5 1.79b 8 1.79tel 1.79tel 1.79tel 1.79bl 1.798el
( se 34 P43 3.7Het 3.76pl 3.765 5 3.M8 5 3.768 5 3.Mbl 3.768el t.404 9 8.7 Met * = . 1.988*4 S.438 7 ' tr St = . 3.7M 4 1.e38+4 6.198+4 1.178 7 4.M8+4 ste M = 3.11t+4 '= 9.373 7
- 9.Mb8 S.M8e7 8te M
- 1.6thi =
'3.NE*1
*- S.138 3 3.868+8 Fe-Il 3.09t*8 1.468+4 - = 0.068 7 8.398*7 3.3?b7 fe-H 1.378+8 3.99be * = 8.368 7 9.968+8 t. Mbt ee 57 - 1.17Ee7 =. = = 3.97be 1.98b7 es le = 3.998 7 * = = 6.M88 6.988 7 es-te .
- t.6?b8 * *
- 3.Mb9 3.etE*4 7- 5643 1.Mble 7.3tb8 = *
- 1.988 8 3.49be
. 3145 4.158 1 7.9the = =
- 3.0M 3 3.658ee "
es-44 = 9.378 4 = 3.M84 ' 3.9ebl 4.388*3 Se45 3.178+8 1.018*9
- 6.798*8 =- 6.M8+8 4. M8+8 Se49 8.788 4 1.678-5 = 1.09bl
- 3.518 4 9.M84 area = * * *
- 1.7M*4 1.51h4 8t=83 = * = =
- 4.6M4 3.318 4 ge.84 = . . . . . .
k45 * * = * * * * .
. a>Se . 3.198 4 * =
- 4.33b7 1.03be
' a>84 . = . . . . .
e>Se . = = = . = = St-89 9. Mbt = = = = 1.Geht 3.Mbe St-90 4.068 11 * * *
- t.798*18 1.488 11
' St-91 3.tebl = = *
- 1.838+4 1.398 4 -
St-93 4.378 3 - = = = 8.468+3 1.8Sb t F90 1.3M 4 * * *
- 1.418+8 3.56b3 F9ta S.83b9 * = *
- 1.788-4 =
F91 S.138+4 - = = = 3.83bt 1.378el F98 9.418 1 * = = = 1.948+4 3.638 3 T 93 1.74b3 = * * . l.538+4 4.Setee St 98 1.198+6 3.ettel = l.97tel - 1.3 t b9 3.588 5 8t-97 3.335 3 4.738 1
- 1.038 3 = 3.88b7 3.088+1 gett 1.438 5 7.91t+4 = 7.8the . 4.80S*8 4.388+4 f upt? 3.9004 7.M8 7
- 8.M8 7 = 3.?tb3 3.448 7 me-M = 4.358*4 = 1.418*7 = 1.458*7 1.198 4 -
fe=99e 3.Mbe 8.M84 = 1.32b3 4.348 4 8.13b3 1.105 3 fe tel = = = = = *
- 8e-103 4.80t*4 =
- 9.83b7 . l.618+8 3.078+6
- 8e-135 S.Mtel =
- 4.ME* 3 - 3.308 4 3.138 1 l Se-tM 1.93pt = = 3.73be = t.388+le 3.44b7 l abtele = = = = = *
- 86546 * * * * = *
- As-ties 1.M8*7 9.M8+4 = 1.938 7 = 3.98pt 8.808 4 Set 34 1.M8+8 1.M8+4 3.535el - 8.988*7 3.95bt 4.118*7 l Setal 1.Mbe 1.538 4 1.H8el
- 1.058 4 1.lebt 3.358 7 I fe itle 9. ME* 7 3.90t*7 3.90s.7 3.M8+8 . 3.M be 1.398 7 Te-137s 3.498+8 1.88b8 8.938*7 1.438+9 '- 1.t?b9 4.M8 7 4 fe-137 S.768 4 3.07b3 4.378*3 3.358 4 = 4.Mbl 1.388 3 Te-t?9e 3.168*4 9.98b7 8.758 7 1.86bt
- 1.888+9 4.e38*7 Te*189 6.668 4 3.905-4 S.198 4 3.79b3 - 8.038 4 1.638 4 Te-tale 9.13bl 4.468el 7.468el 4.53E+4 = 4.43bf 3.738+l fe-tSt = = = * = * ='
fo-933 4.39t*4 3.??>4 3.e684 3.678*7 . = 1.318+8 3.408 4 1 1M 3.M8el 1.17t+6 9.988*? 1.83be = 1.etb4 4.418 5 btSt 8.99b7 1. ut+8 3.79b le 1.98t+8 - 3.068e7 6.638+7 bt33 S.748 1 1.Mba S.Mb3 3.658 3 - 3.09bt. $.388 1 bt33 3.138 4 3.698 4 S.43be 6.448+6
- 3.318+4 1.18~+4 3-tM 1.e684 3.Sede 5.0084 4.5984
- 3.$1b7 1.eM4 3-135 4.048+4 1.etbl 7.M84 1.718+l - 1.$tbl 3.M84
. es-tM 4.M8 91.11ble
- 3.99b9 1.198*9 1.Mb8 9.078+9 !
es-tM 4.8ete? 1.ht+8 = 9.M8e7 1.378 7 1.898 7 1.19b3 1 Se=137 8.29 8.feb9 = 3.96bt 9. Stb 4 1.84b8 S.788+9 l l es-tte * * * = = =
- as-tM 3.958 3 3.908-8
- 9.968 5 1.198-9 5.333 3 8.Hb4 8e=144 1.398+8 1.638el
- 8.49b4 9.3t?4 3.45pt 4.438*4 go.141 . . . . . . . ,
wm .
= . * . . . l Le-144 1.978 3 9.938 3 - * = 7.388*7 3.688+3
! se 143 f.688-4 6.llbl * * = 4.pht 1.588 8 co-let 1.9sbl 1.3M+5 = 4.17b4 = $.e88+4 1.818*4 Co 143 1.005 3 7.43bl - 3. Mbt = 3.778*7 8.3tpt to-M4 3.398 7 f.38b7
- 8.484 - 1.11ble 1.??84 pr-M3 6.M84 3.Mbe . 1.67be = 3.788 4 3.Mb3
. p,.144 . * . . . . . .
e6947 3.M84 3.Mbe
- 3.Hb4 = 1.058 8 3.3tb3 0-147 3.838+4 3.19b4 = * = 1.058 7 1.138 4 sy-n1 1.438 3 1.4eb3 = 4.37b3 = 3.87t*? 7.738 1
J e '- Totee 7.0-4 henueve8) ODCM-7.0 s e* 9eseuues pauser me.e 7esure . T8ssasen
- R3visirn 6 (area /yr ser sci /e83 for 8-3 and Cate Page 7.0-30 (ese aree/yr per sS4/ses) #et o nete ses11de emes Meet Terrold Sideer Sees -
81 443 T.tedy
. 53 . 3.598 3. 3.598*3 3.59s*3 3.998 4 3.99s.3 3.698*3
>M 1.4l84 3.91545 3.9tbl 3.9ttel 3.9tbl 3.9tbl 3.9tbl
,l.<
.- so 34 3.45tel .$.458*$ 3.45tel 3.46bl 3.451 5 3.45bl 3.4Mel
.\ P43 1.615*9 9.96b7 = . = 1.3nb4 4.3M*7 Cr-St . . - 3.448 4 1.Mt+4 8.058 4 t.MS*7 6.3et+4
. Wo-M . 4.838*4 . 1.168 4 . 9.378 4 0.9?s*7 ,
no M = 1.46tel = 1.035 1 . 9.Mtet 3.888+e 1
- Fe-SS 3.365*8 3.3tb8 ' = .
- 1.448 8 9.908*7 8.3tb7 Fe 99 t.8tb8 4.338*8 - - 1.3M*8 9.98b8 1.48t*8 l es-l? - 1.798+7 . . . 3.Mb8 3.4Db7 ,
es-SS . 4.38t*7 . . . 6.Mt+8 1.018+8 co . 3.498 4 . . . 3. Mbt 5.6ebe , 5443 9.ettete 1.13b9 . . = 1.018+8 5.4M+8 ! 5446 S.?M* t 7.33t+0 . . . 3.97b3 3.Mbe ende
- 8.4043 . 3.13t+4 - 6.918+5 3.958*3 as-el 4.ht+8 8.478*9 . 9.4tt+8 . 4.33t+8 4.868 4 Em49 8.198 4 t.Sebl = 1.03bl . 3.888-5 1.998 4 '
.Br-43 . .
. . . - - 1.Hb6 St-83 - = = * * . 3.918 4 se.44 . . . . . . .
, kal . . . . . . .
Spe6 - 3.733+8 . . . 4.098 4 1.385 4 a>te . . . . . . - a>49 . - - . - . . St 89 1.5tE*14 . . . = 1.00t+9 4.318+8 Sr.90 7. stb *tt . . . = 3.ttb ie 1.058*11 Statt 3.998 5 . . . = 1.M8 4 1.198 4 st-93 3.978*3 . = = . 1.018 4 1.49E*1 T-90 1.34b4 - - - - 1.43b8 3.348*3 F9te 5.43F9, . . = . 8.Mb7
- e 1 91 7.878+6 - = = = 3.H8*9 3.118+5 bt3 0.478 1 . = = . 3.338 4 3.4Sb3 bt3 1.4 M* 3 . = = = 4.988+4 4.4?be
.St-95 1.74t*6 5.49b$ . 8.07s 5 . 1.378 9 3.78bl Er-97 3.995+3 6.118*1 - 9.34tet . l.65te? '3.8t b t I
R>95 1.93tel 1.Wtel . 1.e38*5 . 4.llbe S.M84 5>97 3.698-4 6.678 7 . 7.808 7 - 1.99bt 3.44b? he-M
- 5.745*6 = 1.3 tb7 - . 1.9M*7 1.ettee fe*99e 3.708 4 7.Mbe - 1.13b 3 4.19E*4 4.968*3 9.77bt fe let = . . . . . =
1 as-103 6.878*4 = . 3.43b? . l.74t+4 3.htd Re=tel 5.00b t . = 6.31t*3 - 4.M be 1.ME*t as-146 3.etb8 = . 5.97t+8 - 1.485 14 3.tob7 R> 1ene . . '= . . . . i B> t06 . = . . . . = As 110e 1.138*7 1.448*7 . 3.74b7 . 4.Mt+9 4.748 4 sette 1.988+8 3.858+4 3.Hbl . 1.36t+8 3.ttb9 6.enb7 i setal 3. Mt+ 0 3.ME*6 3.'w l = t.Mb8 1.ME*9 S.00b7 Te-ttle 1.488+8 5.ME* 7 4,4b? = = 4.375*4 1.90E*? Te-137e 6.518+8 1.M8*8 1. A F 4 3.Hbt . 1.379 9 6. M p f ; Te 137 S.438*3 1.988 3 3.748*3 3.30t*4 - 4.198 $ 1.178*3 l fe-189e 3.67b8 1.ME*8 1.988+8 1.548 9 . 1.38t*9 S.Sth? Te-139 6.138-4 3.33be 4.4584 3.618 3 - 3.408 3 1.518 4 i Te-13te 4.448el 4.tMel 6.99bl 4.338+4 . 3.3Sb7, 3.3sbl J Te-131 . . . . . . . i Te-133 3.90t*6 3.478+6 3.605 4 3.378+7 . 7.83b7 3.338+6 ! 3 13e 3 ktel 1.038+6 8.Hb7 1.888+6 . 3.47bl 4.00bl 3 131 7.708 7 1.08 b8 3.948*10 1.058 4 = 3.1M*7 S.79b7 !' 3 333 S.188+t 1.Mb3 4.578*3 3.hp3 . S.91bt 4.8?b t ! 3 133 1.978 4 3.M84 4.M84 S.86td . 3.lM4 1.488 4 3-tM 9.598-$ 3. N t 4 4.MS-3 4.91b4 . 3.358 4 9.ttb5 b1M 3.688+4 9.408 4 6.908 4 1.lebl . 1.tStel 3.5384 Co tM 7.99t*9 1.47b10
- 8.3eb9 I.03t*9 3.085*8 7.74k9 es tM 4.39b7 1.698+8 = 9.19h? 1.458 7 9. Mt* ? 1. t M+8 Se-137 1.81ble 1.Hble = 4.598*9 1.748*9 f.93b8 4.69k9 go.134 = * . . = = . l Seatst 3.M8-3 t.908 5 . t.Mbl 1.Mbl 3.4Pb1 8.885 4 1 to Me 1.M8+8 1.49bl . l.76be 1.MS*$ 3.13b8 8.918 4 1 a 14: . . . . . . .
wm . . . . . . . j l Be-let 1.Osb3 8.MBe3 . . . 3.488 7 3.358*3 i ( to let 1.38b4 S.698-l . = = 1.M84 1.438 5 1 Co-le t 3.838el 1.0 Bbl . 8.Mbe - 5.38b8 3.14E+4 1 co-u3 9.37E*3 6.835*l = 3.06b3 . 3.elb? 7.63bt ' co %4 5.3784 3.18b7 . 1.Sek? . 1.338*19 3.8M+4 pr-us 7.138 4 3.84be - t.455 4 . 3.M84 3.6684 prot 44 . . . . . . . spe s* . 3.H8+4 3.M84 - 3.33be . 1.438 4 3.Mb3 p.es/ 3.H84 3.tebe = . . 7.Mbe 1.038+4 sy-339 1.M8*3 1.38b3 . 4.49b3 - 3.18t*7 7.34t*1
+
1 I
a . 9 unes e.4a , cr untae .ER> e
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se8 a manere por ecu.m) sw .4mers seslM. . a user verreM aM , sent Susu t*8 8r
-*- so - 4.etS*3 4.M E*3 4.M 8*3 4.ei4*3 4.M8* 3 4.e18*3 L DM 3.let*6 7.ettel 7.018*$ 7.ettel f.el8+l f.elbl 7.018*5 1 Ap34 3.838 5 3.838el 3.838el 3.838+l 3.838*$ 3.8 Mel 3.83bl
'{
- e 33 3.37bt' 1.Mb8 . . . 9.3eb7 1.308*4 ,
i Sr*ll *
- 8.548*4 1.798*4 1.198*$ 4.3M4 8.10b 5 J 1
Ih44 . 4.61t+4 . 1.858*8 . l.858*8 1.M8*8 l nela . 1.938 1 . 3.398et . 3.788e3 4.M84 Fe-ll 8.eeb8 4.348 8 . . 3.488*8 7.M8*7 1.318*8 ) pe49 4.018*8 6.49b8 . . l.888 8 6.Mbe 3.338*8 1 ert? . 3.998 7 . . . 3.458*8 6.ME*7 I
*- er$8 . 4.478*7 . *
- 3.77b8 1.988 4 aree - . 3.788+8 - . . . 3.188*9 1.13b9 5443 3.998*10 3.198*9 . . . t.438+4 1.M8*9 M48 1.088*3 9.898*0 . . . 1.218*3 S.778+e aree . 1.ltb4 . 3.488+4 ' . 5.3e8*5 6.69b3 arel 8.138*8 3.M8*9' . 1.Ma*9 . 3.008 4 1.3M*9 se49 1.ltbl 3.188-l . 1.338-6 . 1.388 3 3.038 4 Br-83 . . . . . . 3.M84 i tr-83 . . . . . .- l.lSE*e
. ge-04 . . . . . . . F gr-85 . . . . . . .
a>Se - 4.83be . . . 3.9t3 7 3.7884 akte . . . . . . ^ . apoo . . . . . . . St 49 3.998 ele . . . . 1.39E *9 1.4M*9 St-ce t.348*13 . . . . 1.678*10 3.tlbst St-91 5.letel . . . . 1.318 4 3.08be St-93 7.388*3 . . . . 1.3She 3.93tet b90 3.3eb4 - . . . . 6.548*7 6.17b3 b96e 9.M8-9 . . . - 1.9tb5 . F91 9.878*7 . . .- - 3.49t*9 S.018*5 bt3 1.Mbe . . . . 4.51t*4 4.M84 b93 3.etb3 . . . . 4.488+4 8.35be 3r-95 3.Mb6 8.58bl . 1.3M4 . 0.95b4 7.M8* 5 3r 97 5.M8*2 8.158*l . 1.178*3 . 1.3M* 7 4.818*l se9t 4.108*l 1.59tel . 1.lebl
- 3.958*8 1.M8+l s>97 4.904 4 S. Sib 7 . 9.438 7 = 3.738-t 4.13b7 l me-99 . 7.8M+6 = 1.678*7 . 6.488 4 1.Mb6 Te-99m .4.45I 4 9.138*0 . 1.338 3 4.638 4 S.19b3 1.ltb3 te.let . . . . . . .
Benet 1.llte? . . 3.89t*7 . 3.998 8 S.ME*6 trial 9.17t*1 . . 8.N!*3
- l.98be 3.3Het Re-tM 7.45t*8 . = l.elb9 . l.M8+te 9.3eb7 a> tene . . . . . . .
- . . . . . =
- akte6 As-118e 3.338*7 3.178*7 . 4.el8*7
- 3.988*9 1.748*7 Sette 3.535 8 4.578*4 7.78tel . 1.M8+8 3.30t*9 1.33b8 Setal 4.99b8 3.8the 4.4Jtel . 3.788*8 1.198*9 1.elb8 Te-128. 3. lit *8 9.lete? 9.84te? . . 3.38t*8 4.67b7 e to-late 1.338*9 3.leb8 3.48+8 3.77bt . 1.078 9 1.578+8 !
fe-137 1.008*4 3.70t*3 4.938 3 3.858 4 = 3.9ttel 3.168 3 Teat?9m 8.Hb8 3.39b8 3.758*8 3.51>9 s 1.Mt*9 1.3M*4 fe 139 1.168 3 3.338-4 8.338 4 3.378 4 . 7.178 3 3.MS4 Te-13te 1.545 4 S.338*$ t.908 6 5.16be = 3.1M* 7 S.48bl teat 3% Teatst 4.988 4 3.098*4 4.988 4 3.87E*7 = 3.158*7 3.738*4 3 13e 4.stbl 1.N8*6 9.M8*8 1.888 4 = l.87bl 8.478*l 3-131 1.438+8 1.44b8 4.?M*10 3.ME*4 . 1.38847 8.188*7 bt33 9.30tet 1.698*3 7.848 4 3.598 3 . 1.998*3 7.77b t i bial 3.698+4 4.M8+6 8.358+8 7.408 4 . 1.79pe 1.8884 3-th 1.788 4 3.14>4 7.Mt*3 4.Mrd - 3.teb4 1.M84 i I bt35 4.H8*4 1.188*$ t.ME*? 1.818+l 8.985+4 S.578*4 - I es-tM t.808*l0 3.6 Mete
- 8. M8*9 3.938*9 1.438*8 S. Mbt es tM 8.M8e? 3.32b8
- 1.188*8 1.76t*7 7.798+6 t.43be l So-137 3.H8ett 3.39pH . 7.ubt 3.488*9 1.43b8 3.38bt !
ept38 . . ,. . . . . as-139 S.ttha 3.73bl . 3.388 5 9.418-8 3.9the 1.4eb 3 en Me 3.7784 3.438+l . 7.908*4 1.458*5 1.488*4 1.43b7 heut .
- 3. M3 . . . .
, to-Me 3.33b t 1.t38 3 . . . 3.188*7 3.818*3
' Le M3 3.338 4 7.4ebt . . . 1.47bt 8.88b8 erut 1.338*l 6. MS*4 . 3.49be . 7.M8 7 9.838 3 erM3 1.?M* 3 9.M8+l . 3.93>3 . 1.37bf 1.MS*3 arto4 1.878*8 3.008*7 . 3.3tb7 . 1.M8* 44 6.788 4 er*MS 1.48bl 4.448*4 . 3.418*4 . t.Sebe 7.37b3 pr.144 abM7 7.ME**- 8ebe . 3.188*4 . 9.tSt*7 4.40E*3
#.187 8.478*4 3.838 4 * . . l.MS*4 1.73be By-399 3.M8*3 1.83b8
- 8.30t*3 . 1.ME*7 1.39b3 i
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____j
Table 74-4 (seet!*ad) ODCM-7.0 L Rat. Canad Flame Petteer Seu Festen R:vist:n 6 Page 7.0-32 (e8 a aree/rv per sci /ees) amellee any ersee
- t. -
.3 .
i C-te . I 5e=34 1.218*7 ' P 33 . Cs 51 4.Mt+6 to 54 t.Mt*9 me-56 9.958*5 Fe-55 '
=
- Fe-59 3.75t*8 Co SS 3.838*8
' Code 3.M8*te 51 43 -
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*E * . Nucle:r Productlin - F:rmi 2 ODCM-8.0 Offsite Des 3 Calculation M:nu:I Revisian 5 P g3 S.0-1
,- SPECIAL DOSE ANALYSIS
. 8.0 SPECIAL DOSE /.NALYSES 8.1 ' Doses Due to As,+1vities inside the SITE BOUNDARY In accordance with ODCM 5.9.1.8, the Annual Radioactive Effluent Release Report
. submitted within 90 ocys after January 1 of each year shall include an assessment of radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBUC due to their activities inside the SITE BOUNDARY. .
Two locations within the Fermi 2 SITE BOUNDARY are accessible to MEMBERS OF THE
. PUBLIC for activities unrelated to Detroit Edison operational and support activities.
One is the over-water portion of the SITE BOUNDARY due east of the plant. Ice fishermen sometimes fish here during the winter. The other is the Formi 2 Visitor's Center, outside the protected area (but inside the Owner Controlled Area), approximately 470 meters SSW of the Reactor Buliding. The Visitor's Center is open to the public and is routinely visited by MEMBERS OF THE PUBUC, including school tour groups on a frequency of once per year. Also, due to a change in the definition of MEMBER OF THE PUBUC in the revised 10 CFR Part 20, some individuals working within the SITE BOUNDARY may be considered MEMBERS OF THE PUBUC. Conservative assumptionc of locations, exposure times, and exposure pathways for assessing doses from gaseous and liquid effluents due to activities inside the SITE BOUNDARY are presented in Table 8.0-1. The calculational methods presented ' in ODCM Sections 7.6 and 7.7 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC based on the above assumptions. Alternatively, the effluent concentration values of Appendix B, Table 2, of the revised 10 CFR Part 20 may be used to assess dose since these concentrations,if continuously inhaled or ingested, produce a total effective dose equivalent of 50 mrom per year. The potential dose from the fish pathway to a MEMBER OF THE PUBUC engaged in ice fishing within the SITE BOUNDARY is accounted for by the modeling presented in ODCM Section 6.5. Therefore, no additional special dose analyses are required for this exposure pathway for reporting in the Annual Radioactive Effluent Release Report. B.2 Doses to MEMBERS OF THE PUBLIC - 40 CFR 190 The Annual Radioactive Effluent Release Report shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from effluents and direct radiation from onsite sources). For the likely most exposed MEMBER OF THE PUBUC in the vicinity of the Fermi 2 site, the sources of exposure need consider only the radioactive effluents and direct exposure contribution from ARMS - INFORMATION SERVINdd meeMKTir:M ANIY ' Date approved: f .;L /43- Release authorized by: I v I " Change numbers incorporated:__ LCR 93-117-ODM DSN CJdN,-8,d Rev 5 Date NOV 041QQ1
. DTC TMPLAN File 1715.02 Recipient # 3hM i.
,.o- -
,,m.,
ODCM-8.0 Revisian 5 l
, Pag 3 8.0-2 .; No other fuel cycle facilities contribute significantly to the cumulative dose to a .
g MEMBER OF THE PUBUC in the immediate vicinity of the site. Davis-Besse is the ' ( closest fuel cycle facility located about 20 miles to the SSE. Due to environmental dispersion, any routine releases from Davis-Besse would contribute insignificantly to the potential doses in the vicinity of Fermi 2.
. As appropriate for demonstrating / evaluating compliance with the limits of ODCM 3.11.4 (40 CFR 1g0), the results of the environmental monitoring program may be used to provide data on actual measured levels of radioactive material in the actual a pathways of exposure.
8.2.1 Effluent Dose Calculations For purposes of implementing the surveillance requirements of ODCM 3.11.4 and the reporting requirements of ODCM 5.g.1.8, dose calculations for Fermi 2 may be performed using the calculational methods contained within . this ODCM; the conservative controlling pathways and locations of Table 7.0-3 or the actual pathways and locations as identified by the land use census (ODCM 3.12.2 and ODCM g.0) may be used. Liquid pathway doses may be calculated using Equation (6-10). Doses due to releases of radiolodines, tritium and particulates may be calculated based on Equation (7-14). The following equations may be used for calculating the doses to MEMBERS OF THE PUBUC from releases of noble gases. Equation (8-2) is not used for evaluating compliance with 40 CFR Part Ig0, since this regulation does not address skin dose. If noble gases are being released from more than one i point, these equations must be used to evaluate each release point separately, and then the doses must be added to obtain the total noble gas dose. Dtb = 3.17 E - 08
- X/O * [(Kg
- Og)
(8-1) and Ds = 3.17 E - 08
- X/O * {([L +i 1'1 Mg]
- Og) .
where:
= total body dose due to gamma emissions for noble gas Dtb radionuclides (mrom)
Ds = skin dose due to gamma and beta emissions for noble gas radionuclides (mrad) > X/O = atmospheric dispersion to the offsite location (sec/m3 ) Og = cumulative release of noble gas radionuclide i over the , period of interest (uCl)--may be determined according i to Equation (7-8) l Og = concentration of radionuclide i as determined by gamma I spectral analysis of media (uCi/ml)
-- _ . _ _ _ _ _ _ _ - _ . _ . , __ _,.__.m,v, .- .v.-
. -~~- - - . - . . . . . - - -. .
ODCM-8.0 R;visi:n 5 Page 8.0-3 1.67E + 01 = (IE + 03 ml/ liter) * (1 min /60 sec)
~
Ki
= total noblebody dose factor gas radionuclide due to gamma I (mrom/yr per uCl/memissig)ns (from from Table 7.0-2) ~
, Li = skin dose factor due to beta emisglons from noble gas
. radionuclide I (mrom/yr por uCl/m ) (from Table 7.0-2)
Mi
= gamma (mrad /yrair perdose uCl/mfag)or (from for noble Table 7.0-2) gas radionuclide I 1.1 = mrom skin dose per mrad gamma air dose (mrom/ mrad) 3.17 E - 08 = 1/3.15 E + 07 yr/sec Average annual meterological dispersion parameters or meterological conditions concurrent with the release period under evaluation may be used (e.g., quarterly averages or year-specific annual averages).
8.2.2 Direct Exposure Dose Determination From ovaluations performed in the Formi 2 Environmental Report, Section 5.3.4, the direct exposure to the highest offsite location from the Turbine Building N-16 skyshine dose has been calculated to be approximately 3 mrom/ year. This value may be used as a baseline for g actual direct exposure during plant operations. Direct exposure to offsite or
, onsite individuals may be evaluated based on the results of environmental measurements (e.g. area TLD and survey meter data) or by the use of a radiation transport and shielding calculational method. Only during atypical conditions will there exist any potential for significant onsite sources at Fermi 2 that would yield potentially significant offsite doses to a MEMBER ;
OF THE PUBLIC. However, should a situation exist whereby the direct ' exposure contribution is potentially significant, onsite measurements, offsite measurements and calculational techniques will be used for determination of dose for assessing 40 CFR igg compliance. The calculational techniques will be identified, reviewed, and approved at that time, and will be included
, in any report on doses due to such atypical conditions.
8.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data Normally, the assessment of potential doses to MEMBERS OF THE PUBLIC must be calculated based on the measured radioactive effluents at the l plant. The resultant levels of radioactive material in the offsite environment i are usually so r.1lnute as to be undetectable. The calculational methods i presented in this ODCM are used for modeling the transport in the
~
environment and the resultant exposure to offsite individuals. ) 1
. ODCM-8.0 l R:visisn 5 )
Page 8.0-4
~
y- The results of the radiological environmental monitoring program can
\ provide input into the overall assessment of impact of plant operations and radioactive effluents. With measured levels of plant related radioactive material in principal pathways of exposure, a quantitative assessment of potential exposures can be performed. With the monitoring program not identifying any measurable levels, the data provides a qualitative assessment - a confirmatory demonstration of the negligible impact.
I Dose modeling can be simplified into three basic parameters that can be l applied in using environmental monitoring data for dose assessment. D = C
- U
- DF (8-3) where:
l D = dose or dose commitment C = concentration in the exposure media, such as air concentration for the inhalation pathway, or fish, vegetation or milk concentration for the ingestion pathway U = Individual exposure to the pathway, such as br/yr for direct exposure, kg/yr for ingestion pathway ( DF = dose conversion factor to convert from an exposure or uptake to en individual dose or dose commitment The applicability of each of these basic modeling parameters to the use of environmental monitoring data for dose assessment is addressed below: Concentration - C , The main value of using environmental sampling data to assess potential l doses to individuals is that the data represents actual measured levels of radioactive materialin the exposure pathways. This eliminates one main l uncertainty and the modeling has been removed - the release from the i plant and the transport to the environmental exposure medium. j
, Environmental samples are collected on a routine frequency per the ODCM.
To determine the annual average concentration in the environmental medium for use in assessing cumulative dose for the year, an average concentration should be determined based on the sampling frequency and measured levels: '
% = [(C)
- t)/365
_ (8-4) I where. I
% = average concentration in the sampling medium for the year j
Cg = concentration of each radionuclide I measured in the individual sampling medium l l l l
, , . \
! ODCM-8.0 , Revisi:n 5 Page 8.0-5 j' t = period of time that the measured concentration is
\- considered representative of the sampling medium -
(typically equal to the sampling frequency; e.g.,7 days for weekly. samples,30 days for monthly samples). if the concentration in the sampling medium is below the detection , i capabilities (i.e., less than Lower Limits of Detection (LLD), a value of aero i' ! abould be used for Cl(C3 = 0). Exposure - U Default Exposure Values (U) as recommended in Regulatory Guide 1.10g are presented in Table 8.0-2. These values should be used only when specific data applicable to the environmental pathway being evaluated is unavaliable. Also, the routine radiological environmental monitoring program is designed to sample / monitor the environmental media that would' provide early indications of any measurable levels in the environment but not necessarily levels to which any individual is exposed. For example, sediment samples are collected in.the area of the liquid discharge: typically, no individuals are directly exposed. To apply the measured levels of radioactivity in samples that are not directly applicable to exposure to real individuals, the approach recommended is to correlate the location and measured levels to actual locations of exposure. , Hydrological or atmospheric dilution factors can be used to provide - ( , reasonable correlations of concentrations (and doses) at other locations. The other alternative is to conservatively assume a hypothetical individual . at the sampling location. Doses that are calculated in this manner should be presented as hypothetical and very conservatively determined - actual exposure would be much less. Samples collected from the Monroe water supply intake should be used for estimating the potential drinking water doses. Other water samples collected, such as near field dilution area, are not applicable to this pathway.
- Dese Factors - DF The dose factors are used to convert the intake of the radioactive material to an individual dose commitment. Values of the dose factors are presented in NRC Regulatory Guide 1.10g. The use of the RG 1.10g values applicable to the exposure pathway and maximum exposed individual is referenced in Table 8.0-2.
l l t
. l m
w--+ww- - -
'ODCM-8.0 R:visi:n 5
. Pag 3 8.0-6 Assessment of Direct Exposure Doses from Noble Gases Thermoluminescent Dosimeters (TLD) are routinely used to assess the direct )
exposure component of radiation doses in the environment. However, j because routine releases of radioactive material (noble gases) are so low, l the resultant direct exposure doses are also very low. A study
- performed for the NRC concluded that it was generally impractical to distinguish any plant contribution to the natural background radiation levels (direct exposure) below around 10 mrom per year. Therefore, for routine releases ;
from nuclear power plants the use of TLD is mainly confirmatory - ensuring actual exposures are within the expected natural background variation. For releases of noble gases, environmental modeling using plant measured releases and atmospheric transport models as presented in ODCM Sections 7.6 and 8.2.1 represents the best method of assessing potential
~
environmental doses. However, under unusual conditions, direct radiation from noble gas concentrations could be sufficient to cause significant increases in TLD readings; any observed variations in TLD measurements outside the norm should be evaluated. t
- NUREG/CR-0711, Evaluation of Methods for the Determination of X- and Gamma-Ray i Exposure Attributable to a Nuclear Facility Using Environmental TLD Measurements, Gall
( dePlanque, June 1979, USNRC. END OF SECTION 8.0 (: I I e
ODCM-8.0 i
.'" R:visi:n 5 "
. Page 8.0-7 TAtl.E 8.0-1 j I(' '
i Assumptions for Assessing Doses Due to 'I Activities inside SITE BOUNDARY l 8ee Fishing Visitor's Center Site Personnel .; Distance / 470 meters / E 470 meters / SSW various Direction: Estimated 240 hr/yr 4 hr/yr . 2500 hr/yr Exposure '(20 hr/ week over (4 hr/ visit,1 visit Time: 3 month period) per year) Exposure direct exposure direct eiposure direct exposure Pathways: from noble gases from noble gases from noble gases inhalation of inhalation of inhalation of tritium, lodines, tritium, lodines tritium, lodines particulates particulates - particulates ingestion of potable water l direct exposure ( from skyshine Meteorological annual average annual everage annual' average Dispersion: (as determined (as determined (as determined for year being for year being for year being evaluated) evaluated) evaluated) 6.48E-6 sec/m3
- 2.54E-6 sec/m3
- various depending on location of personnel being evaluated l ' Annual average X/O values for 19g1. These values are shown as examples of the range of values to be expected. ,
t
- 1.
i e I
ODCM-8.0 Revisi:n 5 Page 8.0-8 i TABLE 8.0-2
\.
Recommended Exposure Rates in Lieu of Site Specific Data
- Table Reference Exposure Pathway ' Maximum Exposed Exposure Rates for Dose Factor Age Group from RG 1.109 Liquid Releases Fish Adult 21 kg/y E-11 Drinking Water Adult 7301/y E-11 Bottom Sediment Teen 67 h/y E-6 ,
Atmospheric Releases inhalation Teen 8,000 m3 /y E-8 Direct Exposure All 6,100 h/y** N/A Leafy Vegetables Child 26 kg/y E-13 Fruits, Vegetables Teen 630 kg/y E-12 and Grain Milk infant 3301/y E-14 { Adapted from Regulatory Guide 1.109, Table E-5. This table is not a complete list of exposure rates; other applicable values may be found in Regulatory Guide 1.109. Not exposure of 6,100 h/y is based on the total 8760 hours per year adjusted by a 0.7 shielding factor as recommended in Regulatory Guide 1.109. END i { _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___._m
Nucle:r Production - F rmi 2 ODCM-9.0
=
Offsite Dono Calcul:ti:n M:nu:1 Revi:lta 5 Pago 9.0-1
, ASSESSMENT OF LAND USE CENSUS DATA 9.0 ASSESSMENT OF l.AND USE CENSUS DATA A Land Use Census (LUC) is conducted annually in the vicinity of the Fermi 2 site. This census fulfills two main purposes: 1) Meet requirements of ODCM 3.12.2 for identifying controlling location / pathway for dose assessment of ODCM 3.11.2.3; and 2) provide data on actual exposure pathways for assessing reaiistic doses to MEMBERS OF THE PUBLIC.
g.1 Land Use Census es Required by ODCM 3.12.2 As required by ODCM 3.12.2, a land use census shall be conducted during the growing season at least once per twelve months. The purpose of the census is to identify within a 5 mile distance the location in each of the 16 meterological sectors of all pilk producing animals, all meat producing animals, all gardens larger than 500 ft producing broadleaf vegetation, and the closest residence to the plant. The data from the LUC is used for updating the location / pathway for dose assessment and for updating the Radiological Environmental Monitoring Program. If the census identifies a location / pathway (s) yielding a higher potential dose to a MEMBER OF THE PUBUC than currently being assessed as required by ODCM 3.11.2.3 (and ODCM Section 7.7 and Table 7.0-3), this new location pathway (s) shall be used for dose assessment. Table 7.0-3 and plant procedures. shall be updated to include the currently identified controlling location / pathway (s). Also,if the census identifies a location (s) that yields a calculated potential dose (via the same exposure pathway) 20% greater than a location currently included in the Radiological Environmental ( Monitoring Program, the new location (s) shall be added to the program within 30 days, unless permission to take samples cannot be obtained from the affected landowner. The sampling location (s), excluding control locations, having the lowest calculated dose may be deleted from the program after October 31 following the current census. As required by ODCM 3.12.2 and 5.g.1.8, the new location / pathway (s) shall be identified in the next Annual Radioactive Effluent Release Report. The following guideline shall be used for assessing the results from the land use census to ensure compliance with ODCM 3.12.2. g.1.1 Data Compilation
- 1. Compile alllocations and pathways of exposure as identified by the land use census.
- 2. From this compiled data, identify any changes from the previous year's census. Identify the current controlling location / pathway (critical receptor) used in ODCM Table 7.0-3. Also, identify any location currently included in the REMP (Table 10-1).
..e m.G M TTeM b ARMS - INFORMATION SERVICEhINNd I Date approved: 94 /- f3 Release authorized by: [d M d, Change numbers incorporated: LCR g3-117-ODM I #
DSN O d4M. " 6. G@ Rev 5 Date NOV 041993 DTC TMPLAN File'1715.02 Recipient #3 [M J
,-. n-- , . - - -
- ODCM-g.0 ' I e ' R:visi:n 5 1 P:g3 g.0-2 l
,. g.1.2 Evaluation of Relative Dose Significance ,
if any identified changes are likely to change the identity of the critical receptor or to result in a potential dose via a particular pathway which is at least 20% greater than the current maximum dose for that pathway, perform the following:
- 1. Determine the historical, annual average meterological dispersion and (
' deposition parameters (X/Q, D/Q) for any location to be evaluated for i dose significance. All locations should be evaluated against the same historical meterological data set. ;
- 2. ' Perform relative dose calculations based on actual Fermi 2 gaseous . i
- . effluent release's for a recent period of reactor operation, using the - i pathway dose equations of the ODCM. In identifying the critical receptor for Table 7.0-3, all age groups and all pathways relevant to !
ODCM 3.11.2.3 that may be present at each evaluated location are i considered. The critical receptor is assumed to be a member of the age group with the highest calculated dose to the maximally exposed '! organ due to 1-131,1-133, tritium, and particulates with half lives. . greater than 8 days. Other receptors may have higher doses to other organs than the critical receptor has to those organs.
- 3. Formulate a listing of locations of high dose significance in ~
. descending order of relative dose significance. Include the relative dose significance in the listing.
l g.1.3 Program Updates j (
- 1. If any receptor is identified with a higher relative dose than the !
current critical receptor in ODCM Table 7.0-3, this receptor and its ! associated location and pathways should replace the previously ; 4 identified critical receptor information in Table 7.0-3. '
- 2. The Land Use Census data should be used to revise the REMP and '
Section 10.0 of the ODCM in accordance with ODCM 3.12.2,
~
Action item b.
- 2. Any changes in either the controlling location / pathway (s) (critical ,
receptor) for the ODCM dose calculations (Section 7.7 and Table 7.0-3) l or the REMP (ODCM Section 10.0 and Table 10-1) shall be reported to NRC in accordance with ODCM 3.12.2, Action items a. and b. and ODCM 5.g.1.8. l 1 l
. __ _ _ . .__ _ . . . _ . _ . _ , . _ ~ . , _ _ . . _.
ODCM-9.0 . Revislan 5 1 l- Pag) 9.0-3 I I i k NOTE: As permitted by footnote to ODCM 3.12.2, broadleaf vegetation sampling may be performed at the SITE BOUNDARY in two locations, in different sectors with
, highest predicted D/Os,in lieu of the garden census. Also,
- for conservatism in dose assessment for compliance with j ODCM 3.11.2.3 (see also ODCM Section 7.7 and Table 7.0-3),
~ hypothetical exposure location / pathway (s) and conservative dispersion' factors may be assumed (e.g., milk cow at 5 mile location or garden at SITE BOUNDARY in highest D/O sector).
By this approach, the ODCM is not subject to frequent ! revision as pathways and locations change from year to year. A verification that the hypothetical pathway remains conservative and valid is still required. Also, for NRC reporting, the actual pathways and doses should be reported - along with the hypothetical. The reporting of the actual l pathway and doses provides a formal documentation of the { more realistic dose impact. l 9.2 Land Use Census to Support Realistic Dese Assessment The LUC provides data needed to support the special dose analyses of the ODCM Section 8.0. Activities inside the SITE BOUNDARY.should be periodically reviewed for dose assessment as required by ODCM 5.9.1.8 (see also ODCM Section 8.1). Assessment of realistic doses to MEMBERS OF.THE PUBLIC is required by ODCM 3.11.4 for demonstrating compliance with the EPA Environmental Dose. ( Standard,40 CFR 190 (ODCM Section 8.2). To support these dose assessments, the LUC shall include use of Lake Erie water on . and near the site. The LUC shall include data on Lake Erie use obtained from local and state officials. Reasonable efforts shall be made to identify individual irrigation and potable water users, and industrial and commercial water users whose source is t Lake Erie This data is used to verify the pathways of exposure used in ODCM Section 6.5. END OF SECTION 9.0 9 f [ l t 1
Nuclear Pr:ducti n - F:rmi 2 CDCM-10.0 Offalte 00se C lculati:n Manu:1 R:visi:n 5 Pass 10.0-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM I k-10.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of ODCM 3.12.1. The sampling and analysis program described herein was developed to provide representative measurements of radiation and radioactive materials I resulting from station operation in the principal pathways of exposure of MEMBERS OF THE PUBLIC. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50
, and thereby supplements the radiological effluent control program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the affluent measurements and the modeling of the environmental exposure pathways. Guidance for the development of this monitoring
- program is provided by the NRC Radiological Assessment Branch Technical Position on ;
Environmental Monitoring, Revision 1, November 197g. l 10.1 Sampling Locations Sampling locations as required by ODCM 3.12.1 are described in Table 10.0-1 and ,, shown on the maps in Figures 10.0-1,10.0-2,10.0-3, and 10.0-4. Fermi 1 sampling locations are described in Table 10.0-2 and shown on the map in Figure 10.0-5. NOTE: For purposes of implementing ODCM 3.12.2, sampling locations will be modified as required to reflect the findings of the annual land use census as described in ODCM Section g.1 and as required by other contingencies (e.g. unavailability of milk from a listed location). Such changes will be ( documented in plant records and reflected in the next ODCM revision, the next Annual Effluent Release Report, and the next Annual Radiological Environmental Operating Report. Also, if the circumstances of such changes involve a possible change in the maximally exposed individual evaluated for ODCM Control 3.11.2.3, the identity of this individual will be reevaluated. 10.2 Reporting Levels ODCM 3.12.1, Action b, describes criteria for a Special Report to the NRC if levels of plant-related radioactive material, when averaged over a calendar quarter, exceed the prescribed levels of ODCM Table 3.12.1-2. The reporting levels are based on the design objective doses of 10 CFR 50, Appendix i (i.e., the annual limits of ODCM 3.11.1.2,3.11.2.2 and 3.11.2.3). In other words, levels of radioactive material in the respective sampling medium equal to the prescribed reporting levels are representative of potential annual doses of 3 mrom, total body or 10 mrom, maximum organ from liquid pathways; or 5 mrom, total body, or 15 mrom, maximum organ for the gaseous effluent pathway. These potential doses are modeled on the maximum individual exposure or consumption rates of NRC Regulatory Guide 1.10g. .
...e. met 10% N ARMS - INFORMATION SERVIChN Date approved: 9 .1/.'/3 Release authorized by: [h/ MM/
Change numbers incorporated: LCR g3-117-ODM [ # I DSN C4t*,N -/4 d Rev 5 Date NOV 041993 DTC TMPLAN File 1715.02 Recipient NW
ODCM-10.0 Revisi:n 5 P:ge 10.0-2 The evaluation of potential doses should be based solely on radioactive material resulting from plant operation. As stated in ODCM 3.12.1 Action b, the report shall also be submitted if radionuclides other than those in ODCM Table 3.12.1-2 are. detected (and are a result of plant effluents) and the potential dose exceeds the above annual design objectives. The method described in ODCM Section 8.2.3 may be used for assessing the potential dose and required reporting for radionucl. ides other than those in ODCM Table 3.12.1-2. 10.3 Intortaboratory Comparison Program A major objective of this program is to assist laboratories involved in environmental radiation measurements to develop and maintain both an intralaboratory and an ! interlaboratory quality control program. This is accomplished through an extensive Isboratory intercomparison study (' cross-check") program involving environmental media (milk, water, sir, food, soll, and gases) and a variety of radionuclides with activities at or near environmental levels. Simulated environmental samples, containing known amounts of one or more radionuclides, are prepared and routinely distributed to all laboratories upon request. These laboratories perform the required analyses and return their data to the Quality Assurance Branch of the Environmental Protection Agency (EPA). The EPA performs statistical analysis and comparison with known values and analytical values obtained from other participating laboratories. A report and control chart are returned to each participant. The program thus enables each laboratory to document the precision and accuracy of its radiation data, identify instrument and procedural problems, and I compare its performance with that of other laboratories. The environmental laboratory is required to participate in a Commission-approved , interlaboratory Comparison Program and to submit GA Program Progress Summary , Reports to Detroit Edison on a bimonthly or quarterly basis. These reports contain s'ummary descriptions and performance data summaries on reference standards, : blank, blind, spiked, and duplicate analyses, as well as the USEPA and other Laboratory intercommission Programs, as applicable. A summary of the -
- Interlaboratory Comparison Program results obtained is required to be included in the Annual Radiological Environmental Operating Report pursuant to ODCM 5.9.1.7.
Participation in an approved interlsboratory Comparison Program ensures that an independent check on the precision and accuracy of the measurements of radioactive material in environmental sample matrices is performed as part of the OA Program for environmental monitoring in order to demonstrate that the results are valid for . the purpose of Section IV.8.2 of Appendix 1 to 10 CFR Part 50.- l c END OF SECTION 10.0 - 1 I
l ODCM-10.0 '
- R;visi:n 5 Page 10.0-3
'( TABLE 10.0-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM FERMI 2 SAMPLE LOCATIONS AND ASSOCIATED MEDIA KEY 1- T TLD Locations (Pg.10.0-4 to 10.0-6) 2- S Sediments Locations (Pg.10.0-7) 3- F Fish Locations (Pg.10.0-7) 4- M Milk Locations (Pg.10.0-8) 5- DW Drinking Water Locations (Pg.10.0-9) 6- SW Surface Water Locations (Pg.10.0-9) 7- GW Ground Water Locations (Pg.10.0-9) 8- API Air Particulate / lodine Locations (Pg.10.0-10) 9-- FP Food Products Locations (Pg.10.0-11) { l l
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. . +1*
DOCN-12.@' - Ocwtoten S Page 13.0-0 TA8t.E 10.0-1 Redtelosteel Environmentet Monttoring Progree. Feret 2 Sample Lecettene end Aseectated Medte Direct Redtatten toteerslesIca1 Stetten Sector /Aatuuth Dietence from ' 9eumeser Dtrettten Reecter (Appeem) Descrtpt1on medie Freeguency 71 NE/38 8 1.3 at Estrel Beach Pole on Leheshore. Direct Redtetion Q 23 Poles S of Lakeview (Special Area) T2 MNF/22' 1.2 at East of terminetton of Branchoeu St on post Direct Radiation 0 (Special Area) T3 N/9' 1.1 at Pole. NW Corner of Swan Boot Club Fence Direct Redtation Q (Spectet Area) T4 NNW/337' O.6 mt Site Boundary and Toll Rd. on Site Fence by Direct Redtstion O API s2 TS NW/313' O.6 mt $tte Soundary and Tott Rd. on Site Fence by Direct Redtetion ~0 API 83 T6 WNW/2938 D.6 mi Pole. NE Corner of Bridge over Toll Rd Otroct Radletion 0 - l TT W/2708 14.2 mi Pole. behind Doty Farm. 7512 N Custer Rd Olesct Medtetten Q (Control) T8 NW/30S' t.9 mt Pole on Post Rd near NE Corner of Olste Hwy Direct Radiation O and Post Rd T9 NNW/3348 1.5 mi Pole. NW Corner of Trembley and $ wen View Rd Direct Redtatten O T10 N/6' 2.1 et Pole. E Stdo of teesserent - 2 Polee $ Otrect Redtatten O of Chineverre T11 MNE/23e 6.2 mt Pole. NE Corner of teltlimen and Jefferson Direct Redletten 0 T12 NNE/29' 6.3 mt Pointe mouttlee Game Area - Field Office. Pole Direct Redtation Q near Tree. N Area of Perking Lot T13 N/356o 4.1 mt Lobo end Dtate Hwy - Pole on SW Corner Direct Radiation Q with Light T14 NNW/337' 4.4 m1 Lobo and Branden - Pole on SE Corner near RR Direct Redtation O T15 MW/3tS' 3.9 at Pole, behind Newport Post Office Direct Radletton O
e' . .s-m - ODCm-13.0 _& L'evisten 5
-Pese 10.@-0 YASLE 10.5-1 1
Redieiestcet Enwtrennenta9 menttering Progree. Formt 2 Sangeie LecetIene and Aeoec1eted Sted1e Direct Radletten l Meteoretesteel Stetten Sector /Aatuesth Dietence from Nuuher D1roct1en Roseter (Appree) Deser1ptten medie Fresquency TIS WNW/283' 4.9 at Pole. SE of War and Post Roe W rect Redtation 0 l 717 W/271' 4.9 et Pole. WE Corner of Medeau and Lepred near, Otreet Radletten 0 i Mets 11e Home Park T13 FSW/247' 4.8 at Pole. WE Corner of Montet and Hurd atreet medtetten O T19 SW/236' S.2 et ist Pole E of Formt Stron on Waterworks Rd. Direct Redtatten 0 , NE corner of intersection - Storting State ! Park Rd Entreece Drive / Waterworks (in Sterling State Park) T20 WSW/257' 2.7 et Pole. S Side of Witateme Rd - 8 Poles W of Otreet Redtatten 0 Otate Hwy (Special Area) j 721 WSW/239' 2.7 at Pole. N Side of Peert et Parkview - Woodland Direct Redtatten 0 - seech (Spectet Area) l ! T22 S/172' 1.2 at Pole. N $9de of Pointe Aun Peaux 2 Poles W of Direct Redtetion 0 Long - Sito Soundary T23 SSW/195 8 1.1 at Pole. 5 Side of Pointe Aun Peeum - 1 Pole W of Direct Redtetton 0 Huron next to Vent Pipe - Site Boundary 724 SW/22S' t.2 at Feral Gate along Pointe Aun P Rd - on Ofrect Redletion 0 l fence stre W of Gate - Site .eaum ounder, 72S WSW/2S1' 1.5 at Pole. Toll Rd - 13 Polee $ of Formt De Direct Redietten 0 726 WSW/259' 1.1 at Pole. Toll Rd. 6 Polee 5 of Formt Dr Direct Radletten 0 T27 SW/22S' 6.8 et Pole. NE Corner of neuttlen and Emot Front St Direct Radiation 0 (Spectet Area) 728 SW/229' 10.7 et pole. SE Corner of morter Creek and Direct Redtetten 0 LePlatsence (Control) T29 WSW/237' 10.3 mi Pole. E Stdo of S Dtate. 1 Pole 5 of Albein Direct Rediation 0 (Controt) 730 WSW/247' 7.8 et Pole. St. Nery*e Park Corner of Direct Redtatten 0 Elm and Wonroe St. S eide of parking lot next to river (Special Area)
ODCm-13.0 y Revicten S Pese 13.0-3
~i TAet2 13.0-1 Redie1ogiest Enwfrenmente1 menttertng Progree. Fern 1 2 SexpIe t.ecet1ene and Aseec1eted God 9a Direct Redletten Weteoretsgical Stetten Sector /Aatnesth Otetence from Mundsor DIroct1en Reecter (Apprem) Doecr1ptien med9m Frequency 731 .wSw/255' 9.8 at tot Pole w of Entrance Ortve uttton " Pat" Ofrect Redletion 0 Dunson Recreettonal Reserve - N Custer Rd (Control)
T32 WNW/295 8 10.3 mi Pole. Corner of Stony Creek and Finmet Ros Direct Redtatten 0 T33 Nw/317' 9.2 at Pole, w Stoo of Grafton Rd. 1 Polo N of Otrect Redtetion O Ash /Grafton Intersect 9en T34 NNW/338' 9.7 mt Pole, w Stdo of Port Creek. 1 Pole 5 of Direct Redtatten 0 wtit-Carleton Rd 735 N/359' . 6.9 al pote. S Stoe of 5 Huron Rtwer Dr across from Direct Rediatten 0 Race St (Spectal Area) T36 N/3S8e 9.1 mt Pole. NE Corner of Gibeelter and Cahtil Rds Direct Redtatten O T37 NNE/21' 9.8 mi Pole. S Corner of Adams and Gibretter (across Direct Redtetion O f rom Humtpug Merine) T39 WNW/2948 1.7 mt Westdence - 6594 N. Olute Hwy. Direct Redtatten O T39 S/1788 0.3 at SE Corner of Protected Aree Fence (PAF) Otreet Redtatten 0 740 S/170' O.3 mt Midway elong 094 - PAF El'act Redtatten O T41 SSE/181e 0.2 at Midway between OSA and Shtold well - PAF Direct Radietton 0 742 SSE/149' O.2 mt utowey along Shtold watt - PAF Otreet Redtetten O T43 SE/131' O.1 at ' Midway between Shield weit and Aus % ~iers - Otreet Radiation 0 PAF T44 CE/t?9' O.1 mt Opposite 05SF Door - PAF Direct Redtatten 0 T4S 12/968 0.1 mi NE Corner - PAF T46 ENE/878 0.2 mt NE Side Berge Sitp - on Fence Otreet Redietten 0 l
4 ODCD-13.0 Revicion 5 Pege 13.0-7
.O~
l - TABLE 13.D-1 Radleteetest Enwtrennentet wonttoring Preerse. Forest 2 Seagste t.ecet tene and Aenecteted Bodle t l Otreet Radletion l ' Deteoroteelcel Station Secter/Azts?tih Dietence from M er Dt roct 9 a:* Reecter (Appres) Descr9ptfen med1e Frequency i ! T47 S/185 0.1 et South of Turetne side. rollup door on PAF Otreet Redtetton 0 746 Sw/235 0.2 et 30 ft. from corner of AAP on'PAF T49 WSw/251 1.1 mt Corner of alte boundary fence north of NOC Otreet Redtatten Q elong Critteel Path Rd. TSO w/270 0.9 mt $tte boundary fence near metn gate by the Direct Redtatten O ' south Bulttt St. sign T51 N/3 0.4 at Site boundary fence north of North
- Direct Radietten O Cooting Tower T52 NNE/20 0.4 mt Site boundary fence et the corner of Direct Radietten O Arson and Tower T53 NE/55 0.2 ml Site boundary fence east of South 09 rect Redtation O Cooting Tower T54 S/199 0.3 mt Pole, nemt to Formt 2 viettore Center Otreet Redtetton 0 755 WSw/251 3.3 mt First pote oest of Frenchtown Fire Stetton Direct Redtatten 0 entrance. across from Sodt Elementary Schoot 756 w?w/255 4.9 at Pole. entrance to fofferson wlddle School on Direct Redtetion O Stoney Creek Rd.
T57_ W/260 2.7 mi Pole, north side of witttens Rd. scross from Direct Radiation O Jefferson High School entrance T58 WSw/249 4.9 at Pole, west of Hurd Elementary School werquee Direct Redleston Q T59 Nw/325 2.6 mt Pole, north of St. Charles Church Otreet Redtatten 0 entrance on Otate Hwy. TSO NNw/341 2.5 ml tot pole north of North Elementary School OtroCt Rediation O entrance on Diate Hwy. T61 W/266 10.1 at Pole. SW Corner of Stewart and Direct Redtetton Q Roteinwtile Rde. 762 Sw/232 9.T mt Pole, ww corner of Albein and Hutt Rdo. Direct Redtetion C 763 WSW/245 9.6 mt Pote. Corner of Dunber and Telegraph Rdo. Direct Radiatio'n Q i
- - . 00CE-13.0 - :p Rowisten 5 '
Pope 10.0-8 TABLE 10.0-1 Redtelegical Epwtrennentet monttering Program. Formt 2 Sample Lecettene and Aseecte*ed Media Ftsh and Sediment Weteerslog1ca1 StatIen Sector /Amt=uth Sietence from Nuneser DIroct1en Rescter (Appren) Deseriptien tendie Freelvency SEDIteENTS S-1 SSE/19S* 0.9 et Pointe Aun Peeus. Shore 11ne to 500 ft Sediment SA offshore eighting directly to Land Base Water Tower S-2 t/81' O.2 at Fer=1 2 Otecharge. opprom 200 ft offshore Sediment SA S-3 NE/39' 1.1 at Estret Beach, opprom 200 ft offshore. off Sedtwnt SA North shoreline where Seen Creek and Lake Erte meet 3-4 WSW/241' 3.0 mt Indten Tretts Community Beech Sodtment SA S-S NME/20' 11.7 at Deco's Trenton Channet Power Plant intake Sediment SA erea (Control) FISH F-1 MNE/31' 9.5 at Celeron Intend (Controt) Fish SA F-2 E/96' O.4 ml Fermt 2 Discharge (approm 1200 ft offshore) Fish SA F-3 WSW/238' 4.8 mi- Brest Bay Iserine Area (Control) Floh SA _ _ , _ _ _ _ _ < _ - = , - - .. -= _ m ___ _ _ . _ _ _ . _ _ _ _-m___
DOCWa tLQ ' - p Rowtaten 5
, Pope 13.C \' TABLE 13.fr-1 medietogice1 Enwtrennental IsontterIng Preseman. Formt 2 Semple Lecettone and Aeoec1eted Bedte WlIh/Grose meteorelegical Stetten Sector /Antauth Statence from . 90usener O4roctien Reecter (Apprem) Descr1ptten medie Frequency M-2 NW/319' 5.4 mt Resum Form - 2705 E Lobo Mtik M-Su W-9 tmw/2998 9.9 mt Calder Detry - 9334 Finrot Rd Mith M-SM l
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.. ODCW-13.0' m Leviaton 5 Page 10.@-1's TABLE 10.D-1 5tedtelegtest Enytronmentet Mont tering Program. Formt 2 Sample Locet tone and Aseectated IAedte Water IBetoerelostceI Statten Sect or/Ar tsasth Dietence from Number Direction Reacter (Appren) Deecrtption Bedte Frequency DRINMING WATER DW-t S/1748 1.1 et Monroe Water Stetton N Side of Pointe Aum Drinking Water M Peaux 1/2 Block W of Long Rd DW-2 N/8' 18 $ et Detrott Water Stetton Drinking Water M 14700 Woren Rd. Allen Park (Control)
DW-3 55E/180' O.3 mt Fermt 1 Rs= Lake Water Intake Structure Drinking Water M SURFACE WATER l SW-2 MNE/20 8 11.7 mi DECO *e Trenton Channel Power Plant Intake Surface Water M Structure (Screenhouse a1) (Control) SW-3 55E/157' O.2 at DECO *e Formt 2 Generet Service Water Surface Water N Intake Structure SITE WELLS GW-1 S/175' O.4 mi Approm 100 ft W of Lake Erle. EF-1 Perking lot Groundwater Q near goe fired peakers GW-2 55W/208' 1.0 ml 4 ft 5 of Pointe Aus Pesun (PAP) Rd Fence Groundwater Q 427 ft W of where PAP crosses over Stoney
, Point's Western Dthe GW-3 SW/226' 1.0 mt 143 ft W of PAP Rd Gate. 62 ft N of Ground = ster Q PAP Rd Fence GW-4 WNW/2998 0.6 mt 42 ft 5 of Lengton Rd. 8 ft E of Groundwater 0 Toll Rd Fence
- _ 000s-10.@. g~
Cavt3 ten 9 .. _
. Pese 13.G-11'
+
TABLE 13.0-1 Redielos1cei Env1renmonta1 IBenIter1ng Program. Form 1 2 Sempte Locotiene end Assee1eted Wed1e Air Particulate Air Iodine Weteorologlcel Etetten Sector / Azimuth Dietence from Mumber Dtroctien Reecter (Appren) Descr1ptien Medie Fraquency API-1 ME/39' 1.4 mt Estrel Beach Pole on Lakeshore. 18 Poles 5 of Redtotodine W Lakeview (Neerest Consnuni t y w i t h highes t X/Q) Particulates W API-2 MNW/337' O.6 mi Site Soundry and Tott Road. on Site Fence by Radictedtne W T-4 Porticulates W API-3 NW/313' O.6 mt Site Boundry and Toll Road, on Site Fence by Redtotodine W T-5 Particulates W , API-4 W/270 8 14.2 mt Pole, behind Doty Farm - 7512 N Custer Road Redtetodine W (Control) Particulates W API-5 5/191 0 1.2 ml One pole ocuth of Pointe Aun Pesuu Rd on Redlotedtne W Erte St Porticulates W a _ _ _ - _ _ _ _ . . _ _ _ . . - _ _ _ _ _ _ .___- .____ _--w . ~ g
.. 00CW-10.G . w Revtalen S ,
Page 19.C-12 TABLE 13.0-1 Redfelegical Environmental Monttoring Program. Formt 2 Sample Locatione and Aseecleted MedEL Food Products meteorological Stetten Sector /Artmuth Distence from Number Directlen Reector (Appron) Description Wedle Frequency FP-1 NNE/21o 3.9 mt 9501 Turnothe Htchmey Food Products M (when aveilable) FP-3 NNE/128 1.1 mt 6441 Granchesu Food Products N (when avattable) FP-6 WNW/290 8 14.5 mt 8200 Getrman (Control) Food Products M (when avattable) FP-7 WNW/302 8 0.7 mt 6200 Lengton* Food Products u (when avettable)
- calculated critical receptor W
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, 00CIS-10.0 flow t a t en 5 -
m' ' Pege 12.0-13 TABLE 12.0-1 Redie1ogIcel'Enw1rerweenta1 teen 1ter1ng Program. Fermi 2 Sample Lecet1ene end Aeooc1eted Wedie Land Use Census C1eseet Ree1donces , Iteteorological Oletence from Sector Reector (Approm) Description NE tot al 6760 Lakeshcae NME 1.1 et 6460 Broncbeau N 1.1 mi t" f.) Branchesu M*nt 1.1 mt 5701 Post NW t.0 mt 6577 Lerous WNW 0.7 et 6200 Lenstr a W 1.1 mt 6001 Toll WSW 1.4 mt 4981 Pointe Aun Peaum SW 1.3 mt 5194 Pointe Aun Peeun S$W 1.1 et 5820 Pointe Aun Peaus S 1.0 mt 4834 Long ESE-SSE Lake Erle L l i _ _ _ _ _ _ _ - - _ _ _ _ _ . - . . - - - - - - - - - - - - ~ - - - - - - - - - - - - - - - -
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, Rewtut9 . Page 1 *s . G 1 ASt.E 10.0-2 Radtelegical Envi rorument al Isonttoring Program. Formt 1 Sample Locatione and Assectated leedle water Stetten aseteer t ogical Museer Sector /Artouth Oletence from Directten Reacter (appron.) Desertptten Modte Frequency Surface Water South Lagoon S/190' O.5 et Shorettne behind fuel ott storage tank SW SA Swan Creek Nw/56' 1.9 mt Area below bridge N. of Diste 68*, and Swan SW SA Creek Rd. (Public Access area) Reactor NE/O' O.4 mt Area where overflow canal meets Swan Creek SW SA ChanneI Lake water $5E/159' O.4 mt Formt t inlet E. of pumphouse SW SA dr3 Ree City Water City of S/169 8 1.2 mt Wonroe water Station. N. side of Point Aus RW SA Wonroe Peaur. 1/2 block W. of Long Rd. Fermt 1 55E/159' O.4 mt Fermt Unit 1 Raw Lake water intake structure RW SA City of NME/28' 29.5 mt Detroit water Treatment Plant Detrott 14700 moran Rd.. Allen Park RW SA t Note Distances were taken from Formt 2 Reactor Center Line.
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-. Pete it .15 TABLE 80.0-2 Radiological Environmental toonttertne Program. For,al 1 semple Lacettens and Assectated Wedte Sediments Stetten testeertogical Number Sector /Artmuth Dietence freen Directlen Reactor (appros) Descriptton IIedte Freesencir South Lagoon $/190' O.5 mt Shorettne behind fuel ott storage tank Sediment SA Reactor Channel N/O' O.4 mt. Area where overflow canal meets Swan Creek Sediment SA Swan Creek N/w/56' 1.9 at Area below bridge N. of htute Hwy. and $ wen Creek Rd. (pubite access area) Sediment SA S
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Nuclxr Prtducti:n - F;rmi 2 C DCM-APP- A Cffsita D:s3 Calcul:tirn M:nual Rivi21:n 4 Page A-1 APPENDlX A: TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS { L10UID EFFLUENT RELEASES Overview To simplify the dose calculation process, it is conservative to identify a controlling, dose-significant radionuclide and to use its dose conversion factor in the dose calculations. Using the total release (i.e., the cumulative activity of all radionuclides) and this single dose conversion factor as inputs to a one-step dose assessment yields a dose calculation method which is both simple and conservative. Fermi 2 does not have a large data base of previous releases of radioactive liquid effluents upon which to base the determination of the controlling, dose-significant isotope. The Fermi 2 UFSAR, Table 11.2-9 presents the estimated annual releases from liquid effluents as calculated using the NRC OALE computer code, (NUREG-0016, Revision 1). Site specific dose conversion factors (Alo) from ODCM Table 6.0-1 were multiplied by the UFSAR estimated annual release quantity to determine a relative dose significance. Table A-1 presents the results of this relative dose evaluation. Because Cs-134 is the controlling nuclide for the total body dose and has the highest dose conversion factor among the nuclides evaluated for that dose, the use of its dose conversion factor in the simplified dose assessment method for evaluating the total body dose is demonstrably conservative. Selection of the appropriate dose conversion factor for the maximum organ dose is not so straightforward. Inspection of Table A-1 shows that the thyroid dose is the controlling ( organ dose, and it follows that the lodines are the controlling radionuclides. However, this identification is based upon the FSAR estimate of annual releases. To be adequately conservative when using this simplified method, it is appropriate to select the largest dose conversion factor from among all the radionuclides evaluated to assure that offsite doses are not mistakenly underestimated. For the UFSAR Table 11.2-9 isotopes evaluated, there are a few radionuclides with a higher dose conversion factor than 1-133 for the thyroid dose. Further inspection of Table A-1 shows that P-32 is the major contributor to the dose to the bone, which is the second highest organ dose. P-32 has a high dose conversion factor (1.39 E + 06 mrom/hr per uCl/ml) and would provide additional conservatism if used as the simplifying dose conversion factor. However, analysis for P-32 is not required. P-32 decays by beta emission without any accompanying characteristic gammas. Use of the P-32 dose conversion factor is therefore inappropriate. The next largest dose conversion factor of the evaluated radionuclides is Cs-134 for the dose to the liver at 7.09 E + 05 mrem /hr per uCi/ml. (The dose to the liver is the third largest organ dose.) As Cs-134 is easily measured with gamma spectroscopy, has a long half-life, and a high organ dose conversion factor, it is used as the controlling radionuclide for the simplified maximum organ dose assessment. ARMS - INFORMATION(h,b SERkU b __ -,. mm** M M sw a Date approved: 7-3c, '7 2. Release authorized by: W,/ d % [ 7- () w fe w Change numbers incorporated: LCR 91-104-ODM , I DSN 0A6N. '*8//& Rev 4 Date /ftJ{rtAS Y $ , $2/ DTC TMPLAN File 1715.02 Recipient #d$
ODCM-APP-A R: vision 4 Page A-2 (. Simplified Method For evaluation of compliance with the dose limits ODCM 3.11.1.2, the following simplified equations may be used: Total Body D 1.67 E - 02
- VOL ,A Cs-134th) * {0 i g = DF e Z (A-1) where:
Dtb = dose to the total body (mram) VOL = volume of liquid effluents released (gal) DF = average circulating water reservoir decant line flow (gal / min) Z = 5, near field dilution factor (derived from Regulatory Guide 1.109) A(C s-134.tb)
=
5.80 E + 05 mrem /hr per uCi/ml, the total body ingestion dose factor for Cs-134 Ci = total concentration of all radionuclides (uCi/ml) I 1.67 E - 02 = 1 hr/60 min Substituting the value for the Cs-134 total body dose conversion factor, the equation simpilfies to: D tb = 9.69 E + 03 . VOL , [ C, DF e 2 (A-2) Maximum Organ D = 1.67 E - 02 . VOL , A(Cs-134, liver)
- i DF e Z (A-3) where:
Dmax = maximum organ dose (mram) A(Cs-134, liver) = 7.09 E + 05 mram/hr per uCl/mi, the liver ingestion dose factor for Cs-134 4 I s.
ODCM-APP-A R3 vision 4 Pege A-3 ( Substituting the value for the Cs-134 liver dose conversion factor, the equation simplifies to: D = 1.18E + 04 . VOL , [C, DF
- 2 (A-4)
Tritium is not included in the limited analysis dose assessment for liquid releases, because ; the potential dose resulting from normal reactor releases is relatively negligible. Furthermore, the release of tritium is a function of operating history and is essentially unrelated to radwaste system operations. 4 l
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- Nucl:ar Prrdutti:n - Fermi 2 CDCM-APP-D I Offsite D s3 C:lculati:n Manual R;visl5n 4 '
Page B-1 a - APPENDIX B: TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS ( GASEOUS RADWASTE EFFLUENTS ov.rvi.. Dose evaluations for releases of gaseous radioactive effluents may be simplified by the use of an effective dose factor rather than radionuclide-specific dose factors. These effective dose factors are applied to the total radioactive release to approximate the various doses in the environment; i.e., the total body, gamma-air, and beta-air doses. The effective dose factors are based on the typical radionuclide distribution in the gaseous radicactive effluents. This approach reduces the analyses to a single multiplication (Keff. Metf, or Neff) times the quantity of radioactive gases released, rather than
, individual analyses for each radionuclide and summing the results to determine the dose.
Yet the approach provides a reasonable estimate of the actual doses since under normal operating conditions there is relatively little variation in the radionuclide distribution. Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations: Keff = [( Kj e fi ) (B-1) where: g Keff = the effective total body dose factor due to gamma emissions from all noble gases released (mrem /yr per uCi/m , effective)
=
Ki the total body dose factor due to gamma epissions from each noble gas radionuclide i released (mrem /yr per uCi/m , from Table 7.0-2) fi = the fractional abundance of noble gas radionuclide i relative to the total noble gas activity (L + 1.1 M)eff a [((L +i 1.1 Mg)
- fg)
(B-2) where: (L + 1.1 M)ett = the effective skin dose factor due to beta and gargma emissions from all noble gases released (mrem /yr per uCl/m , effective) (Li + 1.1 M 1) = the skin dose factor due to beta and gamma emissions frog each noble gas radionuclide i released (mrom/yr per uCi/m , a from Table 7.0-2)
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ODCM-APP-B Revision 4 - Pago B-2 b M off a [(M ei fg) .
'(8-3)'
where: Moff = the effective' air dose factorgue to gamma emissions from all noble gases released (mrad /yr per uCi/m , effective) Mi = the air dose factor due to ggmma emissions from each noble gas radionuclide 1 - released (mrad /yr per uCi/m , from Table 7.0-2) N eff = [(N) a fg) (B-4) - where: Neff = the effective air dose factorgue to beta emissions from all noble gases - released (mrad /yr per uCl/m , effective) Ni = the air dose factor due to bgte emissions from each noble gas radionuclide i released (mrad /yr per uCi/m , from Table 7.0-2) Normally, past radioactive effluent data would be used for the determination of the effective dose factors. Fermi 2, however, does not have a sufficient operating history at or near full power to provide a reasonable data base for determination of the typical radionuclide 4 distribution in gaseous effluents. Therefore, the UFSAR estimate of radionuclide concentrations at the site boundary is used as the initial typical distribution. The effective dose factors derived from this distribution are presented in Table B 1. Application To provide an additional degree of conservatism, a factor of 2.0 is introduced into the dose , calculation when the effective dose factor is used. This conservatism provides additional . assurance that the evaluation of doses by the use of a single effective dose factor will not significantly underestimate any actual doses in the environment. For evaluating compliance with the dose limits of ODCM 3.11.2.2 the following simpilfied equations may be used: 7 = 2.0
- 3.17 E - 08
- X/O e M off . [Og D
(B-5) and i D
= 2.0 e 3.17 E - 08
- X/O e Neff a [Qi
; (B-6) where: "
D 1 4, y = air dose due to gamma emissions for the cumulative release of all noble gases (mrad) 4. t D = sir dose due to beta emissions for the cumulative release of all noble
- gases (mrad) i
- ODCM-APP-B Revisisn 4 Page B-3
( X/O ' = atmospheric dispersion to the controlling site boundary (sec/m3 )
- Meff - = 2.7 E + 03, effective gamma-air dose factor (mrad /yr per uCi/m3)
Neff . = 2.3 E + 03, effective beta-air dose factor (mrad /yr per uCi/m3)
-Qi
= cumulative release for all noble gas radionuclides (uCl) 3.17 E - 08 = conversion factor (yr/sec) 2.0 = conservatism factor to account for the variability in the affluent data Combining the constants, the dose calculation equations simplify to:
D 7 = 1.71 E - 04
- X/O * [Ol-(B-7) ,
and D = 1.46 E - 04
- X/O * [Ol 8 (B-8)
The effective dose factors are used for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods when the computer or ODCM software may' be unavailable to perform a detailed dose assessment. t
- ODCM-APP-B Revision 4 Page B-4 , TABLE B-1 Effective Dose Factors - Noble Gas Effluents Total Body Skin Dose Gamma Air Beta Air Dose Factor Factor Dose Factor Dose Factor artope Fractional
- Kagg (L+1.1 Magg) Megg Negg Abundance. (mrom/gr per (mrom/gr per uCi/m ) uCi/m ) (mrad uCl/m/yg)per (mrad uCl/m/yg)per Kr-85m 0.10 1.2E+ 02 2.8E+02 1.2E+02 2.0E+02 Kr-85 0.01 1.6E-01 1.4 E+ 01 1.7E-01 2.0E+01 Kr-88 0.04 5.9E+02 7.6E+02 6.1E+02 1.2E+02 Kr-89 0.06 1.0E+ 03 1.7E+ 03 1.0E+03 6.4E+02 X3-133 0.67 2.0E+ 02 4.7E+02 2.4E+02 7.0E+ 02 X3-135 0.02 3.6E+01 7.9E+01 3.8E+01 4.9E+01 XO-137 0.02 2.8E+01 2.8E+ 02 3.0E+01 2.5E+ 02 Xe-138 0.07 6_2 E + 02 1.0E+03 6.4 E+ 02 3.3E+02 TOTAL 2.6E4 03 4.6E+03 2.7E+03 2.3E+ 03 Radionuclide distribution derived from Fermi 2 UFSAR, Section 11.3, Table 11.3-6. Kr-90, Kr-91,
' Xe-139, and Xe-140 have been excluded from the UFSAR distribution because of short half-lives and subsequent decay during environmental transport. Kr-87, Xe-131m, and Xe-133m have been excluded because of their negligible fractional abundance.
END l l s.
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.. 1 DETROIT EDISON - FERMI 2 OFFSITE DOSE CALCULATION MANUAL.
Implementation Plan These revisions go into effect upon approval. Current Revision Status by Section Section Number Revision Number Approval Date 00 7 1.0 5 01/25/94 09/21/93 l [9Y 2.0 5 09/21/93 30 7 4.0 6 01/25/94 l l[T Y 09/21/93 5.0 5 09/21/93 60 6 7.0 6 01/25/94 l 8/7 Y 09/21/93 B.0 5 09/21/93 9.0 5 09/21/93 10.0 5 09/21/93 ; Appendix A 4 08/07/92 Appendix B 4 08/07/92 ' 1 ARMS - INFORMATION SERVICES Date approved: /- 2 i~ - 7 4 Release authorized by: r , Change numbers incorporated: 93-040-O D M [ DSN Rev 7 Dat. JAN 2 6 1994 DTC TMPLAN File 1715.02 Recipient I l
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1.0 INTRODUCTION
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PART I - RADIOLOGICAL EFFLUENT CONTROLS 2.0-2 2.0 DEFINITIONS 3.0-1 3.0 CONTROLS AND SURVEILLANCE REQUIREMENTS 3.0-2 3/4.0 Controls and Surveillance Requirements Applicability
. 3.0-4 3/4.3.7.11 Radioactive Liquid Effluent Monitoring Instrumentation 3
3.0-9 3/4.3.7.12 Radioactive Gaseous Effluent Mor.itoring
' Instrumentation 3.0-18 3/4.11.1.1 Liquid Effluents Concentration 3.0-22 3/4.11.1.2 Uquid Effluent Dose
, 3.0-23 3/4.11.1.3 Liquid Waste Treatment
; 3.0-24 3/4.11.2.1 Gaseous Effluents Dose Rate '
3.0-28 3/4.11.2.2 Gaseous Effluents Dose - Noble Gases 3.0-29 3/4.11.2.3 Gaseous Effluents Dose - lodine-131, lodine-133, Tritium, and Radionuclides in Particulate Form 3.0-30 3/4.11.2.4 Offgas Treatment System 3.0-31 3/4.11.2.5 Ventilation Exhaust Treatment System 3.0-32 3/4.11.2.8 Venting or Purging { 3.0-33 3/4.11.4 Radioactive Effluents Total Dose 1 3.0-34 3/4.12.1 Radiological Environmental Monitoring Program 3.0-46 3/4.12.2 Land Use Census ' 3.0-47 3/4.12.3 Interlaboratory Comparison Program 4.0-1 4.0 BASES 5.0-1 5.0 ADMINISTRATIVE CONTROLS ,
, s.
5.0-2 5.9.1.7 Annual Radiological Environmental Operating Report 5.0-2 5.9.1.8 Annual Effluent Release Report 5.0-4 5.15 Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems PART ll - CALCULATIONAL METHODS 6.0-2 6.0 UQUID EFFLUENTS , 6.0-2 6.1 Radiation Monitoring instrumentation and Controls l 6.0-2 6.1.1 Offsite Dose Calculation Manual (ODCM) 3.3.7.11 8 Requirement '
! 6.0-3 6.1.2 Non-ODCM Required Monitor 6.0- 3 6.2 Sampling and Analysis of Uquid Effluents 6.0-4 6.2.1 BATCH Releases 60-4 6.2.2 CONTINUOUS Releases 6.0- 5 6.3 Liquid Effluent Monitor Setpoints 6.0-5 6.3.1 Tank Effluent Line Monitors I
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Page Section IU .M 6.0-8 6.3.2 Circulating Water Reservoir Decent Line Mediation - Monitor (D11-N402) . ... . ,_ S.0-9 6.3.3 Generic Conservative Alarm Sotpoint for D11-N402 6.0-9 6.3.4 Alarm Setpoint for GSW and RHR System Radiation Monitors 6.0-10 6.3.5 Alarm Response - Evaluating Actual Release Conditions 6.0-10 -6.3.6 Liquid Radwaste Setpoint Determination With Contaminated Circulating Water Pond 6.0-11 6.4 Contaminated GSW or RHR System - Quantifying and Controlling Releases - 6.0-12 6.5 Uquid Effluent Dose Calculation - 10 CFR 50 6.0-12 6.5.1 MEMBER OF THE PUSUC Dose - Uquid Effluents 6.0-14 6.5.2 Simplified Uquid Effluent Dose Calculation 6.0-15 6.5.3 Contaminated GSW System - Dose Calculation 6.0-15 6.6 Uquid Effluent Dose Projections 7.0-1 7.0 GASEOUS EFFLUENTS 7.0-1 7.1 Radiation Monitoring instrumentation and Controls 7.0-1 7.1.1 Effluent Monitoring - Ventilation System Releases 7.0-1 7.1.2 Main Condenser Offgas Monitoring 7.0- 2 7.1.3 Reactor Building Ventilation Monitors (Gulf Atomic) 7.0- 2 7.2 Sampling and Analysis of Gaseous Effluents 7.0-2 7.2.1 Containment PURGE 7.0-2 7.2.2 Ventilation System Releases 7.0-3 7.3 Gaseous Effluent Monitor Setpoint Determination 7.0- 3 7.3.1 Ventilation System Monitors 7.0-5 7.3.2 Setpoint Determination with No Nuclides Detected 7.0-5 7.3.3 Gaseous Effluent Alarm Response - Evaluating Actual Release Conditions 7.0-6 7.4 Containment Drywell VENTING and PURGING 7.0-6 7.4.1 Release Rate Evaluation 7.0-7 7.4.2 Alarm Setpoint Evaluation 7.0-7 7.5 Quantifying Releases - Noble Gases 7.0-7 7.5.1 Sampling Protocol 7.0-8 7.5.2 Release Concentration Determination for Reactor Building Exhaust Plenum 7.0-9 7.5.3 Calculation of Activity Released 7.0-9 7.6 Site Boundary Dose Rate - Radiciodine and Particulates 7.0-10 7.7 Noble Gas Effluent Dose Calculations - 10 CFR 50 7.0-10 7.7.1 UNRESTRICTED AREA Dose - Noble Gases d i b 3J ,O -' *'b '9 * ( - '
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- l 7.0-11 7.7.2 ~
7.0-11 7.4 Simplifled Dose Calculation for Noble' Gases: < 7.0-11 Radiolodine 7.8.1 and Particulate Dose Calculations - 10 CFR 50 UNRESTRICTED AREA Dose - Radiciodine and Particulates 7.0-13 7.8.2 Simplified Dose Calculation for Radiolodines and 7.0-13 Particulates , 7.9 7.0-14 Gaseous Effluent Dose Projection ] 7.10 Waste Oil incineration i 8.0-1 8.0 SPECIAL DOSE ANALYSES i 8.0-1 8.1 ! 8.0-1 8.2 Doses Due to Activities inside the SITE BOUNDARY 8.0-2 Doses 8.2.1 to MEMBERS OF THE PUBLIC - 40 CFR 190 8.0- 3 Ettluent Dose Calculations 8.2.2 Direct Exposure Dose Determination 8.0- 3 8.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data 9.0- 1 90 1 ASSESSMENT OF LAND USE CENSUS DATA 9.0- 1 9.1 9.0-3 9.2 Land Use Census as Required by ODCM 3.12.2 Land Use Census to Support Realistic Dose Assessment , i 10.0-1 10.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 10.0-1 10.1 Sampling Locations 10.0-1 10.2 Reporting Levels 10.0-2 10.3 Interlaboratory Comparison Program APPENDICES . A-1 A B-1 B Technical Basis for Effective Dose Factors Liquid Effluent Releases Technical Basis for Effective Dose Factors Gaseous Radwas TABLES 2.0-6 2.1 2.0-7 2.2 Surveillance Frequency Notation Operational Conditions 3.0- 5 3.3.7.11-1 3.0-7 4.3.7.11-1 Radioactive Liquid Effluent Monitoring instrumentation Radioactive Liquid Effluent Monitoring instrumentation 3.0-10 Surveillance Requirements 3.3.7.12-1 Radioactive Gaseous Effluent Monitoring instrumentation
---wem-
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- TABLE OF CONTENTS (continued) s: . s. s .
Page Section 3.0-14 4.3.7.12-1 Radioactive Gaseous Effluent Monitoring lastrumentation Surveillance Requirements 3.0-19 4.11.1.1.1-1 Radioactive Uguld Waste Sampling and Analysis Program 3.0-25 4.11.2.1.2-1 Radioactive Gaseous Waste Sampling and analysis Program 3.0-36 3.12.1-1 Radiological Environmental Monitoring Program 3.0-42 3.12.1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 3.0-43 4.12.1-1 Detection Capabilities for Environmental Sample Analysis 6.0-17 6.0-1 Fermi 2 Site Specific Liquid Ingestion Dose Commitment Factors, Ago 6.0-19 6.0-2 Bioaccumulation Factors (BFi) 7.0-15 7.0-1 Values for Evaluating Gaseous Release 9ates and Alarm Setpoints 7.0-16 7.0-2 Dose Factors for Noble Gases 7.0-17 7.0- 3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations 7.0-18 7.0-4 Gaseous Effluent Pathway Dose Commitment Factors 80-7 8.0- 1 Assumptions for Assessing Doses Due to Activities inside SITE BOUNDARY 8.0-8 8.0- 2 Recommended Exposure Rates in Lieu of Site Specific Data 10.0-3 10.0-1 Radiological Environmental Monitoring Program, Fermi 2 Sample Locations and Associated Media 10.0-14 10.0-2 Radiological Environmental Monitoring Program. Fermi 1 Sample Locations and Associated Media A-4 A-1 Relative Dose Significance of Radionuclides in Liquid Effluents B-4 B-1 Effective Dose Factors - Noble Gas Effluents FIGURES 3.0-48 3.0-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents 6.0-20 6.0-1 Uquid Radioactive Effluent Monitoring and Processing r, Diagram 7.0-33 7.0-1 Gaseous Radioactive Effluent Monitoring and Ventilation Systems Diagram 10.0-16 10.0-1 Radiological Environmental Monitoring Program Sampling Locations - Site Area 10.0-17 10.0-2 Radiological Environmental Monitoring Program Sampling Locations - Greater than 10 Miles 10.0-18 10.0-3 Radiological Environmental Monitoring Program Sampling Locations - within 10 Miles 10.0-19 10.0-4 Radiological Environmental Monitoring Program Sampling Locations - Site Area (Lake Erie side) 10.0-20 10.0-5 Fermi 1 Sampling Locations END OF SECTION 0.0
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O '"Offsite Oose Calculation Manual Revision 7 Page 3 0-1 h an St'r. rION 3 0 CONTROLS AND SURVE!LLANCE REQUIREMENTS ARMS - INFDRMATION SERVICES Date approved: /-67 f Release authorized by: N M ,
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Change numbers incorporated: 93-040-ouH osu Rev 7 nate JAN 261994 DTC TMPLAN File 1715.02 Recipient
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<..@@ Offsite Does calculation Manual Revision 7 Page 3 0-2 1
3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY [ s., 00NTa0LS 3 0.1' . Compliance with the succeeding Controls is required during'the ' OPERATIONAL CONDITIONS or other conditions specified therein; except' that upon failure to meet the control, the associated ACTION requirements shall be met. 3 0.2 Noncompliance with a Control shall exist when the requirements of the control and associated ACTION requirements are not met within the s-specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. 303 When a Control is not act, except as provided in the associated ACTION-requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the control does not apply
- by placing it, as applicable, in:
- 1. At least STARTUP within the next 6 hours,
- 2. AL.least HOT SHUTDOWN within the following 6 hours, and 3 At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective seasures are compit *d that permit operation under the ACTION requirements, the ACTION may be taker. in accordance with the specified time lirits as ocasured free the time of failure to.seet the Control. Exceptions to these requirements are stated in the individual-Controls. . This Control is not applicable in OPERATIONAL CONDITIONS 4 or 5. - 3 0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall
.not be made when the conditions for the Controls are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with the ACTION-requirements when conformance to thes permits continued operation of the facility for an unilaited period of time. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. . Exceptions to these requirements are stated in the individual Controls.
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APPLICABILITY SURVE!LLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual Controls unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed
; 25 percent of the specified surveillance interval.
4.0 3 Failure to perform a Surveillance Requirement within tne allowed surveillance interval, defined by Surveillance 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Control. The time limits of the ACTION requirements are applicable at the time it
- is identified that a Surveillance Requirement has not been performed..
The ACTION requirecents may be delayed for up to 24 hours to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not t>e made unless the Surveillance Requirement (s) associated with the Control have been performed within the applicable surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONA!. CONDITIONS as required to comply with ACTION requirements.
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. . , . , . Ilu'elear. Production - Fers! 2 ODCH-3 0 ' ~ W " '"-~
J O,M,", 'Offsite' Dose Calculation Manual Revision 7 Page 3 0-4 INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION CDNTROLS 3 3 7.11 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3 3 7.11-1 sna11 be OPERABLE with their alarm / trip setpo.tnts set to ensure that the limits of Control 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATIONAL MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring
- instrumentation channel alarm / trip setpoint less conservative than required by the above control, lanediately suspend the release of radioactive liquid effluents menitored by the affected channel, or declare the channel inoperable, or enange the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent nonitoring instr m ntation channels OPERABLE, take the ACTION shown in Table 3 3 7.11-1 Restore the inoperable instrumentation to OPERAELE status within 30 days and, if unsuccessful, explain why this
- inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.
- c. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1 4 3 7.11 Each radioactive liquid effluent monitoring instrumentation channel shall be denonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4 3 7.11-1. I
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TABLE WOTATIONS ACTION 110 - With the number of channels OPERABLE less than required by the Hinlaus Channels OPERABLE requirement, effluent releases from this pathway may continue provided that prior to initiating a release:
- a. At least two independent samples are analyzed in
- accordance with Control 4.11.1.1.1, and
- b. At least two technically quallfled Individuals Independently verify the release rate calculations and discharge line valving; i
j
' Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 111 - With the number of channels OPERABLE less than the Minimus Channels OPERABLE requirement, radioactive effluent releases via this pathway say continue provided that grab samples are collected and analyzed at least once per 12 hours for gross radi oact ivi t at least 10 y (beta or gamma) at a lower limit of detection of alcrocurle/al, for Cs-137. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 112 - With the number of channels OPERABLE less than required by the Hinlaus Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estlaated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. Otherwise, suspend release of radioactive effluents via this pa thway. ACTION 113 - Suspend release of radioactive effluents via this pathway (1) - Prior to initiating a release, a) at least two independent samples are analyzed in accordance with Cont. ol 4.11.1.1.1, b) the release rate calculations are reviewed by two technically ! quallfled individuals, and c) the discharge line valving is independently verified by a technically quallfled Individual. In addition, the Condensate Storage Tank Discharge Honitor will be continually monitored to ensure that the tank is isolated promptly if the conitor alaras.
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Offsite' Dose Calculation Manual ODCN-3 0 Revision 7 Page 3 0-8 TABLE 4.3.7.11-1 (Continued) TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if any of the following conditions exists: 1. Instrument indicates acasured levels above the alara/ trip setpoint. 2. CircuiL failure. (2) The CHANNEL annunciation FUNCTIONAL occurs if any of theTEST shall following also demonstrate conditions exists: that control room
- 1.
l Instrument indicates seasured levels above the alarm setpoint.
- 2. Circuit failure.
3 Instrumen' indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION shall be performed using National Institute of Standards and Technology traceable sources. These standards shall permit calibrating the system over the range of energy and occurrences. espected during normal operation and anticipated operational seasurement
. For subsequent. CHANNEL CALIBRATION, sources that have been related and to the initial Technology calibration traceable or are National Institute of Standards shall be used.
(4) CHANNEL of release. CHECK shall consist of verifying indication of flow duririg periods CHANNEL CHECK shall be made at least once per 24 hours on days on which continous, periodic, or batch releases are made. (5) The CHANNEL annunciation occursFUNCTIONAL TEST shall if any of the following also demonstrate conditions exists: that control roo 1. Instrument indicates seasured levels above the alarm setpoint.
- 2. Circuit failure.
3 Instrument Indicates a downseale failure. (6) The if anyCHANNEL FUNCTIONAL of the following TEST shall also demonstrate that local alarm oc conditions exists: 1) instrument indicates i measured levels above the alarm setpoint, 2) circuit failure, or 3) instrument indicates a downscale failure. i e
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Page 3 0-9 i t. INSTRUMENTATION ' RADIOACTIVE CASEOUS EFFLUENT HONITORING INSTRUMENTATION CONTROLS 3 3 7.12 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3 3 7.12-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Control 3.11.2.1 are not. exceeded. The alarm / trip setpoints of these channels, with the exception of the offgas monitoring system, shall be determined and adjusted in accordance with the methodology and parameters in the ODCH. APPLICABILITY: As shown in Table 3 3 7.12-1 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alars/ trip setpoint less conservative than req' tired by the above Control, lamediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the AC"I' ION shown in Table 3 3 7.12-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessfu), esplain why this inoperability was not corrected in a timely alnner in the next Annual Radioactive Effluent Release Report.
- c. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4 3 7.12 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNFL FUNCTIONAL TEST operations at the frequencies shown in Table 4 3 7.12-1 l l
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Page 3 0-13 TABLE 3.3.7.12-1 (continued) TABLE NOTATIONS
# At all times. .
## Not used.
*** During operation of the main condenser air ejector.
f During operation of the standby gas treatment system.
## Also included in Technical Specifications Table 3 3 7.5.1 Ites 13.a.
ACTION STATEMENTS ACTION 121 - With the number of channels OPERABLE less than required by the ! Minlaus Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 122 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within 8 hours samples are continuously collected with auxiliary sampling equipment as required in Table 4.11.2.1.2-1. ACTION 123 - Vith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 124 - Not used. ACTION 125 - With the number of channels OPERABLE less than required by the Minlaum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 4 hours and these samples are analyzed for gross activity within 24 hours. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 126 - With the number of channels OPERABLE less than required by the l Mininum Channels OPERABLE requirement, releases via this pathway ' to the environment may continue for up to 7 days provided that*
- a. 1he offgas system is not bypassed, and l D. The reactor building exhaust plenum noble gas effluent (downstreaa) monitor is OPERABLE; Otherwise, be in at least HOT STANDBY within 12 hours.
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1Nuc1 Production - Fermi 2 ':4 544 k' ' pjj..,A00ffs iy Does Calculation Manual Revision 7 Page 3 0-17
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TABLE 4.3.7.12 1 (Continued) TABLE NOTATIONS At all times.
** Not used.
During operation of the main condenser air ejector. During operation of the standby gas treatment system. (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alara annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alara setpoint.
- 2. Circuit failure.
3 Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode (alarm or type).
(2) The initial CHANNEL CALIBRATION shall be performed using National Institute of Standards and Technology traceable sources. These standards shall permit calibrating the system over the range of energy and seasurement expected during normal operation and anticipated operational occurrences. For subsequent CHANNEL CALIBRATION, sources that have been related and to the initial Technology calibration traceable shall be or used.are National Institute of Standards (3) Not used. * (4) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation occurs on hign level and that control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm setpoints.
- 2. Circuit failure.
3 Instrument indicates a downscale failure. i 4. Instrument controls not set in the operate mode (alars or type).
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< Nuclear Production - Fermi 2 !JM , ? 2-5 0 0C W hId M ^" N og b KN..sji.A.'Offsite'DoseCalculationManual ,
Revision 7 Page 3 0-17 TABLE 4.3.7.12-1 (Continued) ! TABLE NOTATIONS At all times. t ** Not used. During operation of the main condenser air ejector. f During operation of the standby gas treatment system. (1) The CHANNEL annunciation FUNCTIONAL occurs if any of theTEST shall conditions following also demonstrate exists: that control room alar 1. Instrument indicates measured levels above the alara setpoint.
- 2. Circuit failure.
3 Instrument indicates a downscale failure. 4. Instrument controls not set in operate mode (alara or type). (2) The initial CHANNEL CALIBRATION shall be performed using National Institute of Standards and Technology traceable sources, h ese standards shall permit calibrating the system over the range of energy and seasurement occurrences. expected during normal operation and anticipated operational For subsequent CHANNEL CALIBRATION, sources that have been related and to the initial Technology calibration traceable shall be or are National Institute of Standards used. (3) Not used. * (4) he CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation occurs on hign level and that control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm setpoints.
- 2. Circuit failure.
3 Instrument indicates a downscale failure. 4. Instrument controls not set in the operate mode (alara or type).
mA'
-10uclear Production - Fermi 2- ODCN-3!O @ < 'hW *Q W @'iff .. .. E'ld GQ c[M Offsite' Dose Calculation ~ Manual Revision 7 Page 3 0-18 3/4.11 RADIOACTIVE EFFLUENTS i
3/4.11.1 LIQUID EFFLUENTS - COWCENTRATION CONTROLS 3 11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 3 0-1) shall be limited to ten times the concentration values specified in 10 CFR Part 20.1001-20.2401, Appendia B. Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved concentrationshallbelimitedto2X10~grentrainednoblegases,the sicrocuries/al total activity. APPLICABILITY: At all times. ACTION: With the concentration of radioactive saterial released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits. SURVE!LLANCE REQUIREMENTS 4.11.1.1.1 Radioactive 11guld wastes shall be sampled ano analyzed according to the sampilng and analysis program of Table 4.11.1.1.1-1. 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCH to assure that the concentrations at the point of release are maintained within the limits of Control 3 11.1.1. 1 ' I
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@...N... 5 Revision 7 Page 3 0-19 TABLE 4.11.1.1.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM -
Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)a Type Frequency Frequency Analysis (uC1/al) A. Batch P P Releaseb : Each Batch Each Batch Principal ca==a 5 X 10~7 Waste Esitters c Sample Tanks (3) or I-131 1 1 10~6
\ Condensate Storage P M Dissolved and 1 X 10-5 Tank One Batch /M Entrained Cases (Gamma Emitters)
P M H-3 1 1 10-5 Each Batch Cceposite Cross Alpha 1 Y 10~I P Q Sr-89 Sr-90 5 I 10~0 Each Batch Composite d Fe-55 1 I 10-6 B Continuous Releases
- M Principajna=== 5 X 10~I Circulating NA Emitters Vater Ccaposite d System (if I-131 1 x 10~6 contaminated)
W H Dissolved and 1 X 10-5 Grab Samp'e Entrained Cases (Camma Emitters) H H-3 1 y 10-5 Composite d Cross Alpha 1 I 10-7 0 Sr-89, Sr-90 5 I 10-8 NA Composite d Fe-55 1 X 10-6 i I l l i l l l , iPi . # f 't ji[' V Q .' j', f*(. JRp,$ efJ$ K k ) $p 4 4 4. * . A p p S P-** v.c4, yrf $ r . g
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"The 1.1.D la defined, for purposes of these controls, as the smallest concentration of radioactive material In a sample that will yield a net count,
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above systes background, that will be detected with 955 probability with only 55 probability of falsely concluding that a blank observation represents a "real" signal. For a particular seasurement system, which may include radiochemical separation: 4.66 ab LLD = E
- V
- 2.22 x 106 . y . ,,p (. y g)
Where: LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 100 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and L for plant effluents is the elapsed Line between the midpoint of sample collection and time of counting. l Typical values of E, V, Y, and t should be used in the calculation. It snould be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement systen and not as an a_ posteriori (after the fact) limit for a particular measurement. , b A batch release is the discharge of 11guld wastes of a discrete volume. < Prior to sampling for analyses, each batch shall be isolated, and then ' thoroughly alred by a method described in the ODCH to assure representative su pling. Iutch 11guld disenarge may be made from only one tank at a time, i 1 i i
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Tl. i cThe principal exclusively are: ganaa esitters for which the LLD specification applies Ce-141. and Ce-144.Mn-54. Fe-59, co-58. Co-60, Zn-65. No-99, Cs-134. Cs-137 considered. his does not mean that only these nuclides are to be Other peaks that are identifiable, Logether with those of the i Effluent Release Report pursuant to Control 5 9.1.8.above nuclides, d A composite sample is one in which the quantity of 11guld samples is proportional to the quantity of liquid waste discharged and in which the method of sampling liquids capinyed results in a specimen that is representative of the released. obtained his may be accomplished through composites of grab samples rior to discharge af ter the tanks have been recirculated.
'A continuous volume; e.g., release is the discharge of liquid wastes of a nondiscrete continuous release.from a volume of a system that has an input flow during the -
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Mer.P.? J0ffsite Does Calculation Manual. Revision 7 Page 3 0-22 . RADIDAC'!VE EFFLUENTS . Id5EE (' , co m 0cs 3 11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in 12 quid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 3 0-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 areas to the total body and to less than or equal to 5 areas to any organ, and
- b. During any calendar year to less than or equal to 3 areas to tne total body and to less than or equal to 10 areas to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical-Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the ilsit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in l compliance with the above limits. This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking-water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.'
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE RDQUIREMENTS I 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in , accordance with the methodology and parameters in the ODCM at least once per 31 i days. "
' Applicable only if drinking water supply is taken from the receiving water body within 3 s!!es of the plant discharge.
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CO m 0LS 3 11.1 3 The 11guid radwaste treatment system shall be OPERABLE and appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the 11guld effluent, from each reacter unit, to UNRESTRICTED AREAS (see Figure 3 0-1) would exceed 0.06 area to the total body or 0.2 area to any organ in any 31-day period. APPLICABILITY: At all times. ACTION:
- a. With radioactive liquid waste being discharged and in excess of the above lialts and any portion of the liquid radwaste treatment systen not in operation, prepare and subalt to the Consission within 30 days pursuant to Technical Specification 6.9.2 a Special Report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without complete treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3 Summary description of action (s). taken to prevent a .
recurrence.
- b. '!he provisions of Controls 3 0 3 and 3 0.21 are not applicable.
SURVEILLANCE REQUIREMENTS 11.11.1 3 1 Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM. 11.11.1 3 2 1he installed 11guld radwaste treatment system shall be demonstrated OPERABLE by meeting Controls 3 11.1.1 and 3 11.1.2. r 50
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3/4.11.2 CASEOUS EFFLUENTS M;f a DOSE RATE
. CONTROLS 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 3.0-1) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 arems/yr to the total body and less than or equal to 3000 arems/yr to the skin, and
- b. For lodine-131. lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 arems/yr to any organ.
APPLICABILITY: At all times. ACTION: With the dose rate (s) etceeding the above limits, immediately restore the release rate to within the above limit (s). SURVEILLANCE REQUIREMENTS I 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCH. 4.11.2.1.2 The dose rate due to iodine-131. todine-133 tritium, and all other radionucildes in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in th;s ODCM by obtaining representative samples and performing analyses in accordance with the sampling
; and analysis program specified in Table 4.11.2.1.2-1.
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TABLE NOTATION 3, -
"The LLD is defined, for purposes of these controls, as the smallest
, concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 55 probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 sb LLD
- E
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LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, su is the standard deviation of the background c.#unting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, Aistheradioactivedecayconstantfortheparticularradionuclide,and t for plant effluents is the elapsed time between the midpoint of sample collection and time of counting. Typical values of E, V, Y, and L should be used in the calculation. It should be recognized that the LLD is defined as an a, priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) 11 alt for a particular acasurement. he principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, l Ie-135 and re-138 in noble gas releases and Mn-54, Fe-59, Co-58, co-60, Zn-65, l Mo-99, 1-131, Cs-134, cs-137, ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual l Radioactive Effluent Release Report pursuant to Control 5.9.1.8.
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' i .5 I N 9 Si' cSampling and analysis shall also be performed following sh'u tdown, startup, or a THERHAL POWER change exceeding 15% of RATED THERHAL POWER within a 1-hour k period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3 d
Tritium grab samples shall be taken at least once per 24 hours when either the reactor well or the dryer-separator storage pool is flooded.
" Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool, f
The ratio of the sample flow rate '.a the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculstion made in accordance with Controls 3 11.2.1, 3.11.2.2, and 3 11.2 3 BS amples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at leard. 7 days following each shutdown, st.artup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hnur and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requircaent does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 conc entration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3 h Required when the SGTS is in operation. I In OPERATIONAL CONDITIONS 1, 2, 3, and 4, the applicable portion of primary j containment shall be sampled and analyzed within 8 hours prior to the start of any PURCING. l 1 J In OPERATIONAL CONDITIONS 1, 2, 3. and 4, when the primary containment I atmosphere radiation monitoring system is declared INOPERABLE or is in alarm condition, the applicable portion of primary containment shall be sampled and analyzed within 8 hours prior to the start of any VENTING or PURGING and at least once per 12 hours during VENTING or PURGING through other than SGTS.
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-i 3 11.2.2 The air dose due to noble gases released in gaseous effluents, from j each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 3 0-1)
) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 meads for beta radiation and,
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At ull times. ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce t.he releases and the proposed corrective actions to be '
taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicaole. ,
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with ) the methodology and parameters in the ODCM at least once per 31 dr,p. l b me. 1 9
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CASEOUS EFFLUDrf3 ,,J
..Ji DOSE - IODINE-131,10 DINE-133. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM CONTROLS P
3 11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131. iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 3 0-1) shall be limited te the following:
- a. During any calendar quarter: Less than or equal to 7 5 areas to any organ and,
- b. During any calendar year: Less than or equal to 15 areas to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of iodine-131, lodine-133, tritius, and radionuclides in particulate form with half-lives ,
greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and subalt to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and definet the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits, ,
- b. The provisions of Controls 3 0..t J 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2 3 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, lodine-133, tritium, and radionuelides in particulate form with half-lives greater than 8 days shall be determined in , sccordance with the methodology and parameters 'in the ODCM at least once per 31 tbys. 4 I I 1 I
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CDfTR0(3 . M. 3.11.2.4 operation. The OFT-CAS TREATMENT SYSTEM shall be OPERABLE APPLICABILITY: operation. Whenever the main condenser steam jet air ejectors are in ACTION: a. With the OFF-CAS TREAINENT SYSTEM inoperable for more than 7 prepare and submit b the commission within 30 days, pursuant to , Technical Specification 6.9.2, a Special Report that includes the following information: 1. Identification of the inoperable equipment or subsystems and the reason for the Inoperability. 2. Action (s) taken OPERABLE status, to andrestore the inoperable equipment to 3 Summary recurrence,description of action (s) taken to prevent a b. The provisions of Controls 3.0 3 and 3.o.4 are not applicable. c. The provisions of Control 4.0.4 are not applicable. SURVEILLANCE PEOUIREMENTS 4.11.2.4 The OFF-CAS TREATMENT SYSTEM shall be demonstrated OPERABL meeting Controls 3 11.2.1, 3.11.2.2, and 3 11.2 3 l l
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TENTILATION EIHAUST TREATMENT SYSTLM _ !. ; CONTROLS e 3 11.2.5 The VENTILATION EXHAUST TREATHENT SYSTEM as described in the ODCM l shall be OPERABLE and appropriate portions of the system shall be used to j reduce radioactive materials in gaseous waste prior to their discharge when the j projected doses due'to gaseous effluent releases from the site to UNRESTRICTED AREAS (see Figure 3 0-1) would exceed 0 3 aren to any organ in any 31-day period. APPLICABILITY: At all times. ACTION:
- a. With radioactive gaseous waste being discharged in excess of the above limits and any portion of the VENTILATION EXHAUST TREATMENT SYSTEM not in operation, prepare and submit to the Commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3 Suasary description of action (s) taken to prevent a recurrence.
- b. The provisions of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from the site shall be projected at : least once per 31 days in accordance with the methodology and parameters in the ODCM, when any portion of the VENTILATION EIHAUST TREATMENT SYSTEM is not in use. 4.11.2.5.2 The VENTILATION EXHAUST TREATMENT SYSTEM shall te demonstrated OPERABLE by meeting Controls 3 11.2.1, 3 11.2.2, and 3 11.2 3 1 i
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RADI0 ACTIVE EFFLUENTS VENTING OR PURGING " ' r CONTROLS 3 11.2.8 VENTING or PURGING of the primary containment shall be through the standby gas treatment system or the reactor building ventilation system. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4 ACTION:
- a. With the requirements of the above control not satisfied, suspend all VENTING or PURGING of the primary containment.
- b. The provision of Controls 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.S.1 The applicable portion of primary containment shall be sampled and analyzed per Table 4.11.2.1.2-1 of Control 3 11.2.1 within 8 hours prior to the start of any PURGING. 4.11.2.8.2 If the primary containment radiation monitoring system is INOPERABLE or is in alara condition, the applicable portion of primary containment shall be sampled and analyzed per Table 4.11.2.1.2-1 of Control 311.2.1 within 8 hours prior to the start of snd at least once per 12 hours during VENTING or PURGING of primary containment through other than the standby gas treatment systes. 4.11. 8.3 The primary containment shall be determined to be aligned for VENTING or PURGING through the standby gas treatment syutes or the reactor building ventilation system within 4 hours prior to start of and at least once ; per 12 hot.rs during VENTING or PURGING of the containment. ! l 4.11.2.8.4 Prior to use of the vent / purge systen through the standby gas treatment systes assure that; l
- a. Both standby gas treatment system trains are OPERABLE whenever the vent / purge systes is in use, and
- b. Whenever the vent / purge system is in use during OPERATIONAL CONDITION 1 or 2 or 3. only one of the standby gas treatment system trains may be used.
4.11.z.8.5 Prior to VENTING or PURGINJ, assure that at least one of the following monitors is CPERABLE: the primary containment atmosphere radiation monitor, the reactor building ventilation exhaust radiation monitor (at least one division), or the SPING monitor corresponding to the release path (the reactor build!ng exhaust plenum radiation monitor or t.he standby gas treatment system radiati .n monitor, Division 1 or 2).
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RADIOACTIVE EFFLUENTS r : 3/4.11.4 'IOTAL DOSE . CONTROLS 3 11.4 The annual (calendar year) dose or dose consituent to any member of the public (as defined in 40 CFR Part 190) due to releases of radioactivity and to radiation from uranius fuel cycle sources shall be limited to less than or equal to 25 areas to the total body or any organ, except the thyroid, which
.shall be limited to less than or equal to 75 areas.
APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls 3 11.1.2a., 3 11.1.2b., 3 11.2.2a., 3 11.2.2b. 3 11.2 3a., or 3 11.2.3b., calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Control 3 11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special deport that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits anc includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a'nenber of the public from uraniun fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report.
It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of_ the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
- b. Tne provisions of Controls 3 0 3 and 3 0.4 are not applicablo.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.11.1.2, 4.11.2.2, and 4.11.2 3, and in accordance with the methodology and parameters in the ODCH. 4.11.4.2 Cumulative dose contributions from direct radiation
- rom the reactor units and from outside storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Control 311.4, ACTION a.
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3/4.12 RADIO!AGICAL ENVIRONMENTAL HONITORING I 3/4.12.1 HONITORING PROGRAM
. . ... ai CONTROLS iw -
1 3 12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3 12.1-1. APPLICABILITY: At all times. . I ACTION:
- a. With the radiological environmental sonitoring program not being conducted as specified in Table 312.1-1, prepare and submit to the Coesission, in the Annual Radiological Environmental Operating Report required by Control S.9.1.7, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30
) days, pursuant to Technical Spe :ification 6.9.2, a Special Report ,' that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose' to A MEMBER OF THE PUBLIC is less than the calendar year limits of Controls 3 11.1.2, 3 11.2.2, and 3 11.2 3 When more than one of the radionticlides in Table 3 12.1-2 are detected in the sampling medium, this report shall be subaitted if: concentration (1) + concentration (2) + ...>~ 1.0 reporting level (1) reporting level (2) When radionuclides other than those in Table 3 12.1-2 are detected and are the result of plant effluents, this rer crt shall be submitted if tne potential annual dose
- to A MEMBER OF "HE PUBLIC frca all radionuclides is equal to or greater than the calendar year limits of Controls 3.11.1.2, 3 11.2.2, and 3.11.2 3 This report is not required if the seasured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report pursuant to Control 5.9.17. '
- c. With milk or fresh lesfy vegetable samples unavailable from one or more of the sample locations required by Table 312.1-1, identify specific locations for obtaining replacement samples and add them to the radiological environssntal monitoring program within 30 days.
1he specific locations fror which samples were unavailable may then
'The methodology used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
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~ c0ffsite' Dose' Calculation Manual Page 3 0-35 7.. . .
4 RADIOLOGICAL ENVIRONMENTAL MONITORING C0FFROLS (Continued) be deleted from the monitoring program. Pursuant to Control 5 9.1.8i identify the cause of the unavailability of samples and identify the - new location (s) for obtaining replacement samples in the next Annual-Radioactive Effluent Release Report pursuant to Control 5 9.1.8 and also include in the report a revised figure (s) and table for the ODCM ' reflecting the new location (s).
- d. The provisions of Controls 3.0 3 and 3 0.4 are not applicable.
SURVEIL!,ANCE REQUIREMENTS 4.12.1 The radiological environaental monitoring sampics shall be collected pursuant to Table 312.1-1 from the specific locations given in the table and figure (s) in the ODCH, and shall be analyzed pursuant to the requirements of Table 3 12.1-1 and the det ection capabilities required by Table 4.12.1-1.
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- 2. AIRDORNE .' .
hadletodine and Sa.vtes from 5 tocattons. particulates Continuous sempler operetton -sth en=pte neototodine Canettaters ',,
- n. 3 samples from close to collectton esemly, or I-131 enalyala weekly. Q~
the 3 5ITE BOuuDAsty loca- more frequently if . . ttone, in different sectors, * ." .. r eau t r ed try dust of the highest ca lcut e t ed tending. Perticulate Someter '<' annues sworege ground- Gross beto rectoectivity level D/Q. analystsfollopfg - filter changet
- b. I sample from the vicinity of a community having the highest Gamme lectopte snelysis' calculated annual average of composite (by j' groundtowel D/Q. location) quarterly.
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plant mastes have been + discharged.
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'Offsite' Dose Calculation ~ Manual ~
- Revision T'"' ,
Page 3 0-40
, , i TABLE 3.12.1-1 (Continued)
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TABLE NOTATIONS
" Specific parameters of distance and direction sector from the centerline of ,
one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 3 12.1-1 in a table and figure (s) in the ODCH. Refer to N11 REG-0133. " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979 Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, salfunction of automatic , sampling equipment and other legitimate reasons. .If specimens are unobtainable due to sampling equipment salfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Contol 5.9.1.7. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 i days in the radiological environmental monitoring program. Pursuant to Control l 5.9.1.8, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next ; Annual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s). bone or more iristruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in plhee of, or in addition to, integrating dosimeters. For the purpose of this tuble, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a > packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optinua dose information with minimal radin,;. cThe purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that have valid background data may be substituted. dAirborne particulate sample filters shall be analyzed for gross beta i radioactivity 24 hours or more af ter sampling to allow for radon and tho'ron daughter decay. If gross beta activity in air particulate samples is greater i than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
'Caram isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents:=from the
' - - n facility.
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e y. TABLE 3.12.1-1 (continued)
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TABLE NOTATION ,
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- 1he "upstrean sample" shall be taken at a distance beyond significant influence of the discharge. The
- downstream" sample shall be taken in an area beyond but near the mixing zone. " Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence.
E Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquot at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly). h Croundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge
,7roperties are suitable for cont.asination. l j
I The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCH.
)!r harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly.
food products.Attention shall be paid to including samples of tuberous and root i
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TABLE 4.12.1-1 (Continued) s, .
TABLE NOTATIONS ,
, j "This list does not mean that only these nuclkbes are to be considered.
Other paah that are identifiable, together with those of the above nuclides, I shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Control 5 9.1.7. bRequired detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13 cThe LLD is deff ned, for purposes of these Controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above systes background, that will be detected with 955 probability with only 55 probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 sh E*V 2.22
- Y exp (- At)
Where: LLD is the "a prior!" lower limit of detection as defined above, as , picocuries per unit mass or volume, j i sb is the standard deviation of the background counting rate or of the ! counting rate of a blank sample as appropriate, as counta per minute, E is the counting efficiency, as counts per disintegration, . V is the sample size in units of tass or volume, 2.22 is the number of disintegrations per minute per picoeurie, Y is the fractional radiochemical yield, when applicable, I histheradioactivedecayconstantfortheparticularradionuclide,and ) L for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical values of E, V, Y, and t should be used in the calculation. i l l l
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r '1 i TABLE 4.12.1-1 (Continued)
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H l It should be recognized that the LLD is defined as an a, priori (before the fact) limit representing the capability of a seasuresent system and not as an a posteriori (af ter the fact) limit for a particular seasurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under . routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Control 5.9.1.7 d LLD for drinking water samples, i i 4 1 i k i l
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~ .ODCW 340 $ . M- . , 4,.p .f M.c-r "N N0ffsitF Dose ~Calculati'on~ Manual Revisio'n'7 Page 3 0-46 i i
RADIOLOGICAL ENVIRONMENTAL MDWITORING 3/4.12.2 LAND USE CENSUS -
,s E.
CONTROLS l 3 12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 reteorological sectorsofthenearestallkgnimal,tgenearestresidenceandthenearest garden
- of greater than 50 m (500 ft ) producing broad leaf vegetation.
APPLICABILITY: At all times. i ACTION:
- a. With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Control 11.11.2 3, identify the new location (s) in the next Annual Radioactive Effluent Release Report, pursuant to control 5.9.1.8.
- b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Control 3 12.1, add the new location (s) to the radiological environmental monitoring program within 30 days.
The sampling location (s), excluding the control station location, having the lowest calculated dose.or dose commitment (s), via the same , exposure pathway, say be deleted from this menitoring program after 1 October 31 or the year in which this lend use census was conducted. Pursuant to Contrcl 5 9.1.8, identify the new location (s) in the next Annual Radioactive ti.ffluent Release Report and also include in the report a revised figure (s) and table for the ODCH reflecting the new location (s).
- c. The provisions of Control 3 0 3 and 3 0.4 are not applicable.
SURVEILLANCE REQU!REMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, visual survey, aerial survey,.or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 5.9.1.7.
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.. Controls for broad lear vegetation sampling in Table 312.1-1, Part 4.c, shall be followed, including analysis of contrni samples.
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Nuclear Production - Fermi 2 . , ~ .;e [fdi."pp"i;<f9ffsite Dose Calculation ^ Manual ODCM-3 0' Revision 7 Page 3 0-47 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM CONTROLS 3.12 3 Analyses shal' be perforned on radioactive aaterials supplied as part of an Commission. Interlaboratory Comparison Program that has been approved by the APPLICABILITY: At all times. ACTION: a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Control 5.9.1.7. b. The provisions of Controls 3.0 3 and 3 0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Cosparison Program shall be described in the ODCH. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 5.9.1.7.
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