ML20217D632
| ML20217D632 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/31/1997 |
| From: | Quennoz S PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PGE-1006-97, VPN-021-98, VPN-21-98, NUDOCS 9804240411 | |
| Download: ML20217D632 (181) | |
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Portland General Electric l One World Trade Center 121 SW Salmon Street Portland OR 97204 April 21,1998 VPN-021-98 Trojan Nuclear Plant Docket 50-344 <
License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 PGE-1006-97. Troian Nuclear Plant Annual Radiological Environmental Monitoring Renort for 1997 i This letter transmits Portland General Electric Company's Trojan Nuclear Plant Annual Radiological Environmental Monitoring Report for the calendar year 1997. This report is submitted in accordance with Trojan Permanently Defueled Technical Specification (PDTS) 5.8.1.2 and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to Title 10 CFR 50.
Sincerely, ;
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Stephen M. Quennoz Trojan Site Executive Enclosure c: R. A. Scarano, NRC Region IV L. It Thonus, NRC, NRR D. Stewart-Smith, OOE 9804240411 971231 PDR ADOCK 05000344 PDR hs s I
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Natural gas. Electricity. Endless possibilities.
HRP-042-98 To: Distribution From: II. R. Pate Date: April 21,1998
Subject:
Transmittal of PGE-1006-97, Trojan Nuclear Plant Annual Radiological Environmental Monitoring Report for 1997
[
Enclosed is your copy of PGE-1006-97, Trojan Nuclear Plant Annual Radiological Environmental Monitoring Report for 1997.
l Please acknowledge receipt of your copy by completing the lower portion of this transmittal and returning it to the location given below.
HRP/CKC Enclosure 4/21/98 ACKNOWLEDGMENT l PGE-1006-97 Trojan Nuclear Plant Annual Radiological Environmental Monitoring Report for 1997 I hereby acknowledge receipt of Controlled Copy Number (s) of the subject document. All changes have been made in accordance with the instructions, and superseded pages have been destroyed.
Signature of Copy Holder Date Return to: Pat Schaffran, TCB-3/ Licensing Trojan Nuclear Plant 71760 Columbia River Highway Rainier, Oregon 97048
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Trojan Nuclear Plant Radiological Environmental Monitoring Report fer Calendar Year 1997 PORTLAND GENERAL ELECTRIC COMPANY f }
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PGE-1006-97 I
TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT i
January through December 1997 l
April 1998 l
Prepared by PORTLAND GENERAL ELECTRIC COMPANY i
With Analyses By Thermo NUtech ALBUQUERQUE, NEW MEXICO l l
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TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT TABIF OF CONTENTS Section Title Page
-TABLE OF CONTENTS ............................... -i-LIST O F TA B LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -lii-LIST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -iv-ABSTRACT........................................ -v-1.0 INTROD UCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 SAMPLING AND PROGRAM PROCEDURES . . . . . . . . . . . . . . . . . 2-1 2.1 SAMPLING LOCATIONS .............................. 2-1 2.2 SAMPLING PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2.1 Air Particulate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2.2 Ambient Radiation Measurements Using TLDs . . . . . . . . . . . . . 2-1 2.2.3 Well Wa ter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2.4 Drinking Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2.5 Shoreline Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 3.0 ANALYTICAL PROCEDURES AND COUNTING METHODS . . . . . . 3-1 3.1 ANALYTICAL DETECTION LIMITS AND UNCERTAINTY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 AIR PARTICULATES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.3 DRINKING AND WELL WATER . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.4 SHORELINE SOIL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.5 AMBIENT RADIATION MEASUREMENTS . . . . . . . . . . . . . . . . . . 3-2 3.6 Q UALITY CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2
3.7 REFERENCES
FOR ANALYTICAL PROCEDURES . . . . . . . . . . . . . 3-2
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TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT TABLE OF CONTENTS Section Title Page 4.0 RESULTS AND DISCUSSION ........................... 4-1 4.1 SAMPLES FROM TIIE TERRESTRIAL ENVIRONMENT . . . . . . . . . 4 4.1.1 Air Particulates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1.2 Well Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1.3 Ambient Radiation Levels .......................... 4-1 4.2 SAMPLES FROM THE AQUATIC ENVIRONMENT . . . . . . . . . . . . 4-2 4.2.1 Drinking Water Samples .................... ...... 4-2 4.2.2 Shoreline Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.3 S UM M ARY O F RES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 5.0 COMMENTS ON AND TERMS USED IN DATA TABLES . . . . . . . . 5-1 APPENDIX A, PGE-1021, OFFSITE DOSE CALCULATION MANUAL, AMENDMENT 14
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9 TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT LIST OF TABLES Number Title 2-1 Sampling Locations and Frequency by Type 3-1 Program Analyses and Reported Detection Levels 3-2 1997 EPA and DOE Interlaboratory Comparison Program Results 3-3 1997 Quality Control Analyses Summary 4-1 Average Gross Beta Concentrations for Air Particulates 4-2 Average Ambient Gamma Radiation Levels 4-3 Average Gross Beta Concentrations for Drinking Water from Columbia River 4-4 Radiological Environmental Monitoring Program Summary 5-1 Gross Beta in Air Particulate Filters 5-2 Summary - Gross Beta in Air Samples 5-3 Gamma Emitters: Concentrations in Air Particulate Filters 5-4 Radioactivity in Well Water 5-5 Ambient Gamma Radiation Levels 5-6 Radioactivity in Drinking Water 5-7 Radioactivity in Shoreline Soil
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9 TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT LIST OF FIGURES Number Title 2-1 Sampling Locations 1
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1 ABSTRACT This report presents the data obtained through the analyses of environmental samples collected through the Portland General Electric Trojan Nuclear Plant Radiological Environmental Monitoring Program for the period January 1,1997, through December 31,1997.
1 Most of the radionuclide analyses on the environmental samples resulted in non-detectable values for radionuclides that could be released from the Trojan Nuclear Plant. In no case did radioactivity that could be attributed to the Trojan Nuclear Plant exceed the Reporting Levels of the Offsite Dose Calculation Manual for Trojan.
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1.0 INTRODUCTION
The Trojan Nuclear Plant, a 1130 megawatt-electric pressurized water reactor, first achieved criticality on December 15,1975. On January 27,1993, Portland General Electric decided to permanently shut down the Trojan Nuclear Plant. This report presents the analytical data from the Radiological Environmental Monitoring Program with appropriate interpretation for 1997.
The analytical contractor during this period has been Thermo NUtech, Albuquerque, New Mexico. In comparing data obtained during this period with those from previous periods, care should be taken to ensure that differences in procedures among the contractors are considered.
Information concerning the Radiological Environmental Monitoring Program prior to this period may be found in earlier reports.
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2.0 SAMPLING AND PROGRAM PROCEDURES 2.1 SAMPLING LOCATIONS Fifteen (15) sampling locations were used in the Radiological Environmental Monitoring Program from January 1,1997, through December 31,1997. These sampling locations are shown in Figure 2-1 Table 2-1 includes a listing of the sites, their distance from Trojan, and the type and frequency of sample collection.
During 1994 a review of the environmental sample results from 1977 through 1993 was conducted. In general, the review confirmed that radioactivity attributable to Trojan Nuclear Plant during power operations was not detected in the environmental samples. Therefore, since the production of radioactivity had ceased when the reactor was permanently shut down, and from that point forward, the radioactivity in both liquid and gaseous effluents continued to decrease, it was evident that the environmental sampling requirements could be reduced.
Therefore, revisions to the Radiological Environmental Monitoring Program were submitted to the Oregon Department of Energy (now known as the Oregon Office of Energy) on September 22,1994, for approval. The revisions to the program were approved on December 12,1994.
2.2 SAMPLING PROCEDURES 12.1 AIR PARTICUMIE 1 Air particulate sampling was performed weekly. The samples were gathered with a low-volume air sampling device which is designed to draw a constant flow rate regardless of i the pressure drop across the filter. The sampling devices were set to maintain one cfm. The sample pump, metering devices, and timer were in a weatherproof housing. The filter was located in a sample housing that is connected to an air inlet about one meter above the ground.
Glass fiber filters were used to collect particulate matter.
The glass fiber filter was removed from the air sampler and placed in a two-inch plastic petri dish. Air flow readings and other data required to compute the levels of radioactivity were recorded and submitted to the analysis laboratory along with the samples.
I 2.2.2 AMBIENT RADIATION MEASUREMENTS USING TLDs
[ Thermoluminescent dosimeters (TLDs) were placed for field exposure and collected on a quarterly frequency. The TLDs were placed about one meter above ground level in plastic
- containers. The time of collection, the exposure period, and any abnormal conditions such as moisture in the holders, damage done by animals, etc., were recorded when the TLDs were retrieved. Care was taken to minimize exposure to the TLDs between collection and delivery to the laboratory. Trip TLDs were carried with the field TLDs during transport to and from the field.
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2.2.3 WFI.T. WATER Well water was collected quarterly from the tap that leads off the pump. The line was purged for about five minutes prior to collection. Sixty milliliters were drawn from the one-gallon sample for tritium analysis. The remainder of the sample was put in a one-gallon polyethylene bottle and acidified with concentrated hcl. The bottles were securely sealed and labeled, and collection data forms were prepared specifying site, date collected, volume, and sample type.
2.2.4 DRINKING WATER \
Four-week composite samples of municipal drinking water were collected for Rainier (Sample l Location 8) and St. Helens (Sample Location 9) at their respective intake structures on the Columbia River. Rainier is downstream of the Trojan Nuclear Plant while St. Helens is upstream. At each location, a compositing sampler took a sample every two hours and aliquots of this four-week composite were sent for analysis. From these aliquets,60 milliliters are sent for tritium analysis and two one-gallon polyethylene bottles are acidified with concentrated hcl and sent for the other analyses. The bottles were securely sealed and labeled, and collection data forms were prepared specifying site, date collected, volume, and sample type.
2.2.5 SHORFIINE SOIL '
Shoreline soil samples of about one quart in volume were taken twice a year. The samples were taken from a one square foot area at a depth of between one and four inches. Vegetation l and large rocks were removed from the sample before it was placed in a plastic container. ;
The containers were securely sealed and labeled. The sample site identification number, date collected, and volume obtained were recorded on the collection data forms.
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3.0 ANALYTICAL PROCEDURES AND COUNTING METHODS Samples are analyzed for the various radioactive components by standard radiochemical methods. These methods are equal to, and in most cases, identical with, those of the U. S.
Department of Energy [ Health and Safety Laboratory (HASL) Procedures Manual, HASL-300, see references, Section 3.7], or those of the U. S. Environmental Protection Agency (EPA). l Analyses of individual sample types, general methods, and routine analytical sensitivities are discussed below. The analytical program and sensitivity requirements are given in Table 3-1.
3.1 ANALYTICAL DETECTION LIMITS AND UNCERTAINTY In environmental radiological analyses the dominant known uncertainty is usually the sample count rate. This uncertainty is calculated by standard methods (HASL-300), and is reported at the 95 percent confidence level (20). The lower limit of detection (LLD) is defined as the smallest concentration of radioactive material in a sample that will yield a net indication, above system background, that will be detected with 95 percent probability with only five percent probability of falsely concluding that a blank observation represents a real signal.
Analytical data for samples for which concentrations are less than or equal to the LLD are preceded by the symbol " < " unless otherwise specified.
3.2 AIR PARTICULATES Gross beta concentrations are measured with low background, window-type (0.85 mg/cm2 in thickness), proportional counting systems. The LLD for gross beta measurements is less than 2
or equal to 0.01 pCi/m assuming a collected air volume of 285 m /3 week. l Gamma isotopic analyses are performed with germanium detectors. The LLD requirements for gamma scans are given in Table 3-1.
3.3 DRINKING AND WELL WATER Gross beta analysis of water samples is performed by evaporation of a measured aliquot of the sample, digestion, planchetting of the processed sample and radiometric assay by the low-background beta counters mentioned in Section 3.2, with an LLD of 1 pCi/ liter. Tritium analysis is performed on water samples to the required LLD of 1000 pCi/ liter by liquid scintillation counting. Gamma isotopic analysis is performed using germanium detectors. The LLD requirements for gamma scans are given in Table 3-1 l
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3.4 SHORELINE SOIL Samples are oven-dried and results reported based on dry weight. Gamma emitters are measured with germanium detectors. The LLD requirements for gamma scan are given in Table 3-1.
3.5 AMBIENT RADIATION MEASUREMENTS Quarterly ambient gamma radiation measurements were made using TLDs supplied by a vendor. Each environmental dosimeter is composed of two CaF 2:Dy (TLD-200) elements and two LiF:Mg,Ti (TLD-100) elements, all of which are 0.035 inches thick. The CaF2 :Dy elements are shielded by 80 mg/cm2 ABS plastic,0.010 inches of tantalum and 0.002 inches of lead. The LiF:Mg,Ti elements are shielded by 80 mg/cm2 ABS plastic only.
Environmental dosimeters retrieved from the field are sent to the vendor for processing on a quarterly basis, i
! 3.6 OUALITY CONTROL A large number of the analyses performed by the analysis laboratory are for quality control purposes. The analysis laboratory participates in Environmental Protection Agency (EPA) and Department of Energy (DOE) interlaboratory comparison programs for environmental measurements. Reports of quality control analyses are presented monthly to PGE.
Results of EPA and DOE interlaboratory comparisons for 1997 are given in Table 3-2. In l those cases where the laboratory failed the performance evaluation study, the laboratory l performs an investigation to determine the cause and corrective action as required. Table 3-3 summarizes the spiked sample results for the year 1997.
3.7 REFERENCES
FOR ANALYTICAL PROCEDURES
- 1. American Public Health Association, American Water Works Association and Water Pollution Control Federation (1971): Standard Methods for the Examination of Water l and Wastewater. Thirteenth edition, pp 583-632; 12th edition, pp 325-352. APHA, 1740 Broadway, New York, NY 10019.
- 2. Department of Health, Education and Welfare, Public Health Service: Radioassay l Procedures for Environmental Sarr j_ National Center for Radiological Health (1967), Sec.1, pp 36-115.
- 3. Atomic Energy Commission: Regulatory Guide 4.3 (September 1973).
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- 4. Health and Safety Laboratory, Atomic Energy Commission: HASL Procedures Manual (now known as EML of the Department of Energy). HASL,376 Hudson Street, New York, NY 10014.
- 5. National Environmental Research Center, Environmental Protection Agency; Handbook of Radiochemical Analvtical Methods. Program Element 1HA 325. Of6ce of Research and Development, Las Vegas, NV 89114.
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TABLE 3-1 PROGRAM ANALYSES AND REPORTED DETECTION LEVELS Program Analysis Reported Detection Limits (LLDV*I Air Particulate-gross beta 0.01 pCUm' Air Particulate-gamma scan 0.05 pCi/m3 Cs-134 0.06 pCi/m3 Cs-137 Water-gross beta 1 pCi/ liter Water-tritium 1000 pCi/ liter Water-gamma scan 15 pCi/ liter Mn-54 15 pCi/ liter Co-58 15 pCi/ liter Co-60 30 pCi/ liter Zn-65 30 pCi/ liter Zr-95 15 pCi/ liter Nb-95 15 pCi/ liter Cs-134 18 pCi/ liter Cs-137 Shoreline Soil-gamma scan (dry) 0.15 pCi/g Cs-134 0.18 pCi/g Cs-137 Direct Radiation 0.04 mR/ day or less l'1 Reported detection level or LLD is defined in Section 3.1 t
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. l TABLE 3-2 Sheet 1 of 2 1997 EPA AND DOE INTERLABORATORY COMPARISON PROGRAM RESULTS Sample Study EPA TNU/A Deviation Tvoe Analysis Date Value pCi/l Value oCi/l (known)
Water Beta January 97 14.7 5.0 12.9 1.8 -0.64 Pass Water H-3 March 97 7900 790- 7105 i 82 -1.74 Pass Water Beta April 97 102 15 94.5 1.5 -0.86 Pass Water Co-60 April 97 21.0 5.0 20.7 i 0.6 -0.12 Pass Water Cs-134 April 97 31.0 i 5.0 26.0 1.0 -1.73 Pass Water Cs-137 April 97 22.0 i 5.0 23.0 0.0 0.35 Pass l Water Co-60 June 97 18.0 5.0 20.3 0.6 0.81 Pass !
Water Zn-65 June 97 100 10 103 2 0.46 Pass Water Cs-134 June 97 22.0 5.0 22.3 i 0.6 0.12 Pass Water Cs-137 June 97 49.0 5.0 48.3 2.5 . -0.23 Pass Water Ba-133 June 97 25.0 i 5.0 23.7 0.6 -0.46 Pass Water Beta July 97 15.1 5.0 17.7 1.1 0.90 Pass Water H-3 August 97 11010 1101 9850 i 180 -1.83 Pass Water ' Beta October 97 143.4 21.5 125.5 t 1.0 -1.44 Pass Water Co-60 October 97 10.0 5.0 9.67 0.6 -0.12 Pass Water Cs-134 October 97 41.0 5.0 36.7 0.6 -1.50 Pass Water Cs-137 October 97 34.0 5.0 35.0 1.0 0.35 Pass Water Beta October 97 48.9 5.0 46.8 t 0.3 -0.74 Pass Water Co-60 November 97 27.0. 5.0 28.0 1.7 0.35 Pass Water Zn-65 November 97 75.0 8.0 75.7 3.5 0.14 Pass Water Cs-134 November 97 10.0 5.0 11.0 1.0 0.35 Pass Water Cs-137 November 97 74.0 5.0 76.0 1.7 0.69 Pass Water Ba-133 November 97 99.0 i 10 87.0 2.0 -2.08 Pass l
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TABLE 3-2 Sheet 2 of 2 1997 EPA AND DOE INTERLABORATORY COMPARISON PROGRAM RESULTS 1
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Sample Study DOE TNU/A Tyne Analysis Date Value pCi/ Filter Value pCi/ filter Ratin Filter Ce-144 April 97 424.3 i 27.0 489.2 26.5 1.15 Pass Filter Co-57 April 97 291.9 27.0 340.5 11.2 1.17 Pass Filter Co-60 April 97 135.4 8.1 151.9 9.6 1.12 Pass Filter Cs-134 April 97 294.6 27.0 332.4 12.2 1.13 Pass l Filter Cs-137 April 97 235.1 i 21.6 266.5 11.6 1.13 Pass Filter Mn-54 April 97 205.9 16.2 250.5 11.3 1.22 Pass ,
Filter Sb-125 April 97 332.4 i 27.0 416.2 21.4 1.25 Pass I Filter Beta April 97 12.2 0.8 10.6 2.1 0.88 Pass Filter Ce-144 October 97 516.2 18.9 562.1 i 20.5 1.08 Pass l Filter Co-57 October 97 340.5 11.6 389.2 11.6 1.13 Pass Filter Co-60 October 97 289.2 29.5 291.9 10.8 1.00 Pass Filter Cs-134 October 97 762.2 19.7 759.5 23.0 0.99 Pass Filter Cs-137 October 97 197.6 6.8 222.4 8.4 1.12 Pass Filter Mn-54 October 97 181.6 7.3 200.5 7.8 1.10 Pass Filter Sb-125 October 97 435.1 i 21.4 508.1 i 19.7 1.16 Pass Filter Beta October 97 81.1 3.8 69.2 5.9 0.85 Pass Sampw Study DOE TNU/A
@t Analyiz.s Date Value pCi/g Value pCi/g Ratia Soil Cs-137 April 97 22.3 0.4 22.4 0.7 1.00 Pass Soil K-40 April 97 9.1 0.2 8.7 0.8 0.97 Pass Soil Cs-137 October 97 21.9 1.1 23.8 0.7 1.08 Pass Soil K-40 October 97 8.5 i 0.5 7.9 1.5 0.92 Pass l
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l TABLE 3-3 i 1997 QUALITY CONTROL ANALYSES
SUMMARY
l The table below summarizes results of samples run for process quality control purposes during the subject year. These listings are in addition to such measurements as detector backgrounds, check source values, radiometric-gravimetric comparisons, system calibrations, l etc. Detailed listings of each measurement are maintained at the analysis laboratory and are available for inspection if required.
Spiked Samnles Nuclide Number of Within 2 Sigma Between 2-3 Analyzed Determs. of Known Sigma of Known l
Alpha 91 84 7 l Beta 124 113 11 l H-3 74 73 1 l
C-14 4 4 0 Sr-90 19 18 1 Tc-99 15 15 0 Cs-137 127 124 3 Pb-210 8 8 0 Po-210 28 25 3 Ra-226 36 28 8 Ra-228 17 17 0 Th-228 112 101 11 Th-230 112 102 10 Th-232 112 105 7
]
U-234 82 75 7 U-235 82 80 2 U-238 82 74 8 Np-237 2 2 0 i Pu-239 78 72 6 Am-241 7 7 0 Uranium 122 121 1 i
4.0 RESULTS AND DISCUSSION 4.1 SAMPLES FROM THE TERRESTRIAL ENVIRONMENT 4.1.1 AIR PARTICULATES The gross beta air particulate data obtained during 1997 were comparable to the data obtained during the years 1982 through 1996 (except May 1986) and the preoperational period. Gross l beta concentrations for air particulates for sampling periods in 1997 remained generally ;
at low levels.
l Average concentrations 'with their average standard deviations for the years 1997 and before are presented in Table 4-1 for both onsite and offsite locations. Due to revisions of the Radiological Environmental Monitoring Program, air samples were only collected at onsite ;
locations during 1997.
In October 1980, the People's Republic of China tested a nuclear device in the atmosphere.
For this reason, the increased average concentrations in 1981 were due to increased fallout levels from the October 1980 Chinese test and not from operation of the Trojan Nuclear 4 Plant. The larger average standard deviation for the 1986 data was due to the in: ceased gross beta activity for May 1986 resulting from the Chernobyl reactor accident near Kiev, Ukraine.
For 1997, the measurement of gamma emitting radionuclides in monthly composites of air -
particulate filters resulted in no detectable activity.
Data for these air monitoring samples are listed in Chapter 5 Tables 5-1,5-2, and 5-3.
4.1.2 WEI I WATER Well water samples were collected on a quarterly basis. Tritium levels were below the sensitivity requirements of the program. Gamma emitting radionuclides were not detected in well water samples. The data are presented in Chapter 5, Table 5-4.
4.1.3 AMBIENT RADIATION LEVELS Gamma radiation levels (mR/ day) for dosimeter measurements at locations in the environs around the Trojan Nuclear Plant during 1997 are shown in Chapter 5, Table 5-5.
The elevated radiation levels from Location 15 were due to radioactivity contained in the refueling water storage tank (RWST) and radioactive materials stored in outside areas of the Restricted Area during 1997. The_ dosimeter at this location is on the Industrial Area fence, 4-1 i
which is immediately adjacent to Trojan's Restricted Area boundary. The elevated radiation levels were limited to a section of the East Industrial Area fence. This area is on a bluff that overlooks the Restricted Area and is also in direct line of sight with the RWST.
All of the other dosimeter measurements obtained within the Controlled Area showed no increase in ambient radiation levels. Therefore, the measurements from Location 15 were not included in the determination of the onsite mean ambient radiation level used for comparison with the mean ambient radiation level for the control locations. However, the increase in the radiation levels was evaluated by performing a dose assessment for a member of the public present in the area adjacent to the East Industrial Area fence. This dose assessment has been included in Trojan's Annual Radiological Effluent Release Report for-1997. j 1
The mean and standard deviation for the Trojan onsite measurements during 1997 were O.08i0.02 mR/ day. This is less than, but not significantly different from, the mean and standard deviation of 0.09 0.02 mR/ day for the control locations. Average gamma radiation ,
levels with their standard deviations for the years prior to, and including,1997 are presented !
in Table 4-2 for both onsite and control locations. )
l 4.2 SAMPI ES FROM THE AOUATIC ENVIRONMENT 4.2.1 DRINKING WATER SAMPIM Drinking water samples were collected from municipal water supply locations on the Columbia River that are downstream (Sample Location 8) and upstream (Sample Location 9) of the Trojan site. The ' samples were composited on a monthly basis and analyzed for gross beta activity, tritium, and gamma emitters. The data are presented in Chapter 5, Table 5-6.
l No radioactivity attributable to operation of the Trojan Nuclear Plant was detected in any of )
the water samples.
)
- Table 4-3 presents the annual average of the gross beta activity for the two water sample sites from 1980 through 1997. These samples were not collected prior to 1980. The annual average values do not differ significantly over the years.
4.2.2 SHORELINE SOIL l i
Shoreline soil samples were collected from a location on the bank of the Columbia River near the Trojan site. Analyses were performed for gamma emitters. The data are presented in Chapter 5, Table 5-7. None of the shoreline soil samples showed detectable levels of gamma emitters.
4-2
4.3
SUMMARY
OF RESULTS Table 4-4 presents a summary of the radioactivity analysis results for each medium or pathway sampled during 1997 for the Radiological Environmental Monitoring Program. The format of Table 4-4 is that which is required by ODCM Control 5.1.1.
A review of Table 4-4 shows that none of the radioactivity measurements, averaged over a quarter year period, were larger than the Reporting Levels defined by ODCM Control 3.3.1.
Air particulate samples were collected from two onsite locations during 1997. Location 13 is at the North site boundary and Location 12 is South of the plant at the meteorology tower.
The gross beta annual mean concentrations of 0.022 0.009 pCi/m' for Location 12 and 0.022i0.011 pCi/m' for location 13 were both less than, but not significantly different from, the five year (1990-1994) mean concentration of 0.025 0.014 pCi/m' for the control location.
For the ambient radiation measurements, the mean value for the control locations was not significantly different than the mean values for the Trojan onsite locations.
For the radioactivity measurements in drinking water, the annual mean for the gross beta determination was higher (though not significantly) for the upstream or control location (St.
Helens) than it was for the downstream location (Rainier).
As is shown by Table 4-4, there is no indication that the operations of the Trojan Nuclear Plant had a radiological impact on the environs around the Plant.
l l
l
)
l l
4-3
TABLE 4-1 AVERAGE GROSS BETA CONCENTRATIONS FOR AIR PARTICULATES
- (10-2 pCi/m')
i Trojan l Year ~ h c on Ore _ Washinoton Preop 2 2 22 3 2 1976 2i6 3 8 2i4 1977 3 4 44 5 2 1978 2 2 2 1 2 1 1979 11 1 1 1 1 1980 3i4 3 4 2i4 1981 11 i 2 11 i 4 11 1 1982 2 5 27 2i6 1983 22 2 2 2 2 l 1984 2i2 2 2 2 2 1985 2i2 2 1 2 1 1986 3 7 3 6 3 7 1987 lil 1 1 1 1 1988 1 1 1 1 1 1 l
1989 2 2 2i2 2 2 .
1990 2 1 2i1 2 1 I 1991 2 1 2 1 2 1 1992 2 1 2 1 2 1 l 1993 3 2 3i2 3 2 1994 32 3 1 3 1 1995 2 1 *
- 1996 2 1 *
- 1997 2 1 *
- o Due to revisions of the Radiological Environmental Monitoring Program, air samples are no longer collected at offsite locations.
1
]
TABLE 4-2 1
AVERAGE AMBIENT GAMMA RADIATION LEVELS mR/ Day
_Y. car Trojan Oregon _ Washington Site 1976 0.13 0.14 0.13 1977 0,13 0.15 0.14 1978 0.11 0.13 0.13 1979 0.11 0.02 0.14 i 0.02 0.13 0.03 l 1980 0.11 i 0.02 0.14 0.02 0.12 i 0.01 l 1981 0.11 i 0.03 0.14 0.02 0.12 i 0.02 1982 0.14 i 0.03 0.16 0.02 0.15 i 0.02 j i 1983 0.12 0.02 0.14 i 0.02 0.13 i 0.01 l 1984 0.12 i 0.03 0.13 0.02 0.12 i 0.02 1985 0.12 i 0.03 0.14 i 0.02 0.12 i 0.02 1986 0.12 i 0.03 0.14 0.03 0.12 i 0.02 l_ 1987 0.13 i 0.03 0.15 0.03 0.12 i 0.02 1988 0.12 i 0.02 0.14 0.02 0.12 i 0.02 i
1989 0.11 0.02 0.14 0.02 0.12 i 0.02 1990 0.11 0.02 0.13 0.03 0.11 0.02 1991 0.11 i 0.02 0.13 0.02 0.13 0.02 1992 0.10 0.03 0.13 0.03 0.12 i 0.02 1993 0.10 i 0.03 0.12 i 0.03 0.10 i 0.03 l 1994 0.19 i 0.03 0.22 0.03 0.20 i 0.03 1995 0.08 i 0.02 0.11 0.01
- 1996
- 0.09 i 0.02 0.11 i 0.02 1997 0.08 i 0.02 0.09 0.02
- I o Due to revisions of the Radiological Environmental Monitoring Program, ambient gamma radiation levels are no longer measured in the state of Washington.
4
TABLFE I AVERAGE GROSS BETA CONCENTRATIONS j FOR DRINKING WATER FROM COLUMBIA RIVER (Units: pCi/l)
No. 8 -Rainier No. 9 - St, Helens Year (Downstream) (Upstream) 1980 2 2 2i1 1981 2 1 3 1 1982 3 2 4 2 1983 3 2 4 2 1984 3i2 4 2 I 1985 3 2 4 1 1986 3 2 3i2 1987 3 2 4i1 1988 4i2 63 1989 3 2 4 2 1990 2 3 5 3 1991 3 3 li2 1992 2 1 3 1 1993 2 1 3 1 1994 2i1 3 1 1995 2 0.4 3 1 1996 2 0.4 3il 1997 1.7 0.5 2.7i0.8 l
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5.0 COMMENTS ON AND TERMS USED IN DATA TABLES Dry Weight A reporting unit used for shoreline soil in which the amount of sample is taken to be the weight of the sample after removal of moisture by drying l in an oven at about 110 C for about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
pCi/m' A reporting unit used with air particulate data which refers to the radioactivity content expressed in picocuries per unit volume of air expressed in cubic meters passed through the glass fiber filter. Note i i
l that the volumes are not corrected to standard conditions.
Gamma Emitters Samples were analyzed by high resolution germanium gamma or spectrometry. The resulting spectrum is analyzed by a computer Gamma Isotopic program which scans about 50 to 2000 kev and lists the energy peaks of any nuclides present in concentrations exceeding the sensitivity limits set for that particular experiment. '
Error Terms Figures following " " are error terms based on counting uncertainties at the 20 (95 percent confidence) level unless otherwise specified.
l Values preceded by the " < " symbol were below the stated l
concentration as defined by the notation associated with Table 4.3-1 of Trojan's Offsite Dose Calculation Manual.
l l
l l
l l 5-1 1
TABLE 5-1 Sheet 1 of 2 GROSS BETA IN AIR PARTICULATE FILTERS Location 12 Location 13 Collection Volume Gross B Volume Gross B Date (m') (pCi/m') (m') (pCi/m')
1/07/97 280 0.012 i 0.002 280 0.012 i 0.002 1/14/97 280 0.022 i 0.002 280 0.021 0.002 1/21/97 285 0.045 0.003 285 0.050 i 0.003 1/28/97 285 0.020 0.002 285 0.023 0.002 2/04/97 280 0.023 i 0.002 285 0.024 i 0.002 2/11/97 290 0.039 0.003 290 0.041 i 0.003 2/18/97 280 0.018 i 0.002 280 0.023 i 0.002 2/25/97 285 0.015 0.002 285 0.012 0.002 3/04/97 280 0.017 i 0.002 235' O.015 0.002 3/11/97 290 0.016 i 0.002 290 0.017 i 0.002 3/18/97 280 0.014 i 0.002 285 0.014 i 0.002 3/25/97 285 0.017 i 0.002 285 0.017 i 0.002 4/01/97 290 0.016 i 0.002 290 0.017 i 0.002 4/08/97 280 0.015 i 0.002 280 0.015 0.002 4/15/97 280 0.019 i 0.002 280 0.021 i 0.002 4/22/97 290 0.020 i 0.002 290 0.019 0.002 4/29/97 285 0.016 i 0.002 285 0.016 i 0.002 5/06/97 275 < 0.014 280 0.033 i 0.01 5/13/97 285 0.026 0.009 285 < 0.013 5/20/97 290 0.024 0.01 280 0.017 0.008 5/27/97 285 0.014 i 0.007 285 0.019 i 0.007 6/03/97 285 < 0.011 285 < 0.011 6/10/97 285 < 0.011 285 0.009 i 0.005 6/17/97 290 < 0.011 290 0.013 0.006 6/24/97 280 < 0.009 280 0.010 i 0.006 7/01/97 285 0.019 i 0.007 285 0.016 i 0.006 6
7/08/97 285 0.020 i 0.009 160 0.042 i 0.015 Partial sample due to circuit breaker trip 6
Power to sampler intermpted due to construction activities Partial sample due to maintenance on power supply
TABI E 5-1 Sheet 2 of 2 GROSS BETA IN AIR PARTICULATE FILTERS Location 12 Location 13 Collection Volume Gross B Volume Gross B Date (m') (pCi/m') (m') (pCi/m')
6 6 7/15/97 285 < 0.012 6 6 7/22/97 290 <0.010 6 6 6 6 7/29/97 8/05/97 290 0.014 i 0.002 290 0.014 0.002 8/12/97 280 0.017 i 0.002 280 0.017 i 0.002 8/19/97 290 0.016 0.002 290 0.016 0.002 8/26/97 275 0.022 i 0.002 270 0.021 0.002 6 6 9/02/97 295 0.016 i 0.002 9/09/97 280 0.022 0.002 280 0.020 i 0.002 9/16/97 285 0.022 i 0.002 285 0.022 i 0.002 9/23/97 290 0.023 0.002 290 0.023 i 0.002 9/30/97 285 0.030 i 0.002 285 0.034 i 0.002 10/07/97 280 0.021 i 0.002 280 0.019 0.002 10/14/97 285 0.018 t0.002 285 0.020 i 0.002 10/21/97 290 0.029 i 0.002 290 0.026 i 0.002 10/28/97 280 0.025 i 0.002 280 0.027 i 0.002 11/04/97 290 0.017 i 0.002 290 0.020 i 0.002 11/11/97 280 0.030 i 0.003 280 0.028 i 0.002 l 11/18/97 285 0.054 i 0.003 285 0.044 0.003 11/25/97 285 0.017 i 0.002 285 0.014 i 0.002 l 12/02/97 290 0.015 i 0.002 290 0.017 i 0.002 i 12/09/97 300 0.041 0.003 175* 0.065 i 0.004 !
12/16/97 290 0.025 i 0.002 235 0.015 i 0.002 12/23/97 280 0.014 0.002 280 0.013 i 0.002 12/30/97 285 0.021 i 0.002 285 0.019 0.002 Partial sample due to circuit breaker trip 6
. Power to sampler interrupted due to construction activities Partial sample due to maintenance on power supply
TABLE 5-2
SUMMARY
- GROSS BETA IN AIR SAMPLES i
1 l
l- oCi/m 3 Mean + 10 Maximum Minimum l Trojan Onsite Stations I Location 12 0.022 i 0.009 0.054 0.012 Location 13 0.022 i 0.011 0.065 0.009 l
l l
l l
r i .
TABLE 5-3 l GAMMA EMITTERS: CONCENTRATIONS IN AIR PARTICULATE FILTERS l (Monthly Composites)
Collection (pCi/m')
Dates Location 12 Location 13 12/31/96-1/28/97 < LLD <, LLD j 1/28/97-2/25/97 < LLD < LLD l 2/25/97-3/25/97 < LLD < LLD 3/25/97-4/29/97 <LLD < LLD 4/29/97-5/27/97 < LLD < LLD 5/27/97-6/24/97 <LLD < LLD
- 6/24/97-7/22/97* < LLD < LLD 7/29/97-8/26/97 < LLD < LLD 8/26/97-9/23/97 < LLD < LLD 9/23/97-10/21/97 < LLD < LLD 10/21/97-11/18/97 < LLD < LLD i
- No samples for period of 7/22/97 to 7/29/97 due to interruption of power to samplers.
Composite for period 6/24/97 to 7/22/97 for Location 13 consists of two samples.
l l
l l LLD: 0.05 pCi/m 3Cs-134 0.06 pCi/m' Cs-137 i
i l
l l
i TABLE 5-4 RADIOACTIVITY IN WELL WATER Collection pCi/l 1 Date Location 11 Gamma Tritium Emitters 3/11/97 < 1000 < LLD 6/10/97 < 1000 < LLD 9/9/97 < 1000 < LLD 12/9/97 < 1000 < LLD l LLD: 15 pCi/l Mn-54, Co-58, Co-60, Nb-95, Cs-134 l 18 pCi/l Cs-137 30 pCi/l Zn-65, Zr-95 1000 pCi/l H-3 l
l l
(
l TABLE 5-5 t
AMBIENT GAMMA RADIATION LEVELS mR/ Day l First Quarter Second Quarter Third Quarter Fourth Quarter Location 12/26/96-3/27/97 3/27/97-6/26/97 6/26/97-9/29/97 9/29/97-12/29/97 1 0.09 0.08 0.08 0.10 2 0.08 0.04 0.04 0.06 3 0.08 0.07 0.05 0.09 o 4 0.08 0.07 0.06 0.08 5 0.10 0.07 0.08 0.10 6 0.10 0.08 0.08 0.10 7 0.08 0.06 0.% 0.07 8 0.10 0.09 0.08 0.10 9 0.12 0.09 0.07 0.10 13 0.08 0.06 0.04 0.08 14 0.09 0.07 0.07 0.08 15 0.14 0.17 0.13 0.20 i
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TABLE 5-7 i
RADIOACTIVITY IN SHORELINE SOIL (Semiannual Collections) pCi/g (dry) 4 i
Location 10
{
Collection Ganuna l Date - Emitters 3/11/97 < LLD )
9/9/97'l1
< LLD l'1 Split sample with the State of Oregon LLD: 0.15 pCi/g Cs-134 0.18 pCi/g Cs-137 1
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TROJAN NUCLEAR PLANT RADIOLOGICAL ENVIRONMENTAL i MONITORING REPORT l
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OFFSITE DOSE CALCULATION MANUAL I
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Amendment 14 i
l ll Portland General Electric Company 121 SW Salmon Street Portland OL 97204 i I
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TROJAN NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL CONTENTS j PART I i
RADIOLOGICAL EFFLUENT CONTROL PROGRAM AND ENVIRONMENTAL MONITORING PROGRAM Section Title Pace
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . 1-1 2.0 DEFINITIONS Action . . . . . . . . . . . . . . . . . . . . . . . . 2-1 Channel Functional Test . . . . . . . . . . . . . . . 2-1 Channel Calibration . . . . . . . . . . . . . . . . . 2-1 Channel Check . . . . . . . . . . . . . . . . . . . . 2-1 Frequency Notation . . . . . . . . . . . . . . . . . . 2-1 Liquid Radwaste Treatment System . . . . . . . . . . . 2-1 l Member (s) of the Public . . . . . . . . . . . . . . . 2-2 Offsite Dose Calculation Manual . . . . . . . . . . . 2-2 Operable - Operability . . . . . . . . . . . . . . . . 2-2 Purge - Purging . . . . . . . . . . . . . . . . . . 2-2 Site Boundary . . . . . . . . . . . . . . . . . . . . 2-2 l Solidification . . . . . . . . . . . . . . . . . . . . 2-2 Source Check . . . . . . . . . . . . . . . . . . . . . 2-3 )
I Unrestricted Area . . . . . . . . . . . . . . . . . . 2-3 Ventilation Exhaust Treatment Systems . . . . . . . . 2-3 1 1 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . 3/4 0-1 3/4.1 INSTRUMENTATION Radioactive Liquid Effluent Instruanentation . . . . . 3/4 1-1 Radioactive Gaseous Effluent Instrumentation . . . . . 3/4 1-5 1
1 3/4.2 RADIOACTIVE EFFLUENTS Liquid Effluents - Concentration . . . . . . . . . 3/4 2-1 Dose . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-4 Liquid Waste Treatment . . . . . . . . . . . . . . . . 3/4 2-5 Gaseous Effluents - Dose Rate . . . . . . . . . . . . 3/4 2-6 Gaseous Effluents - Dose, Noble Gases . . . . . . . . 3/4 2-9 Gaseous Effluents - Dose - Tritium and Radionuclides in Particulate Form . . . . . . . . . 3/4 2-10 Ventilation Exhaust Treatment . . . . . . . . . . . . 3/4 2-11 l
l i Amendment 11 (December 1994)
TROJAN NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL CONTENTS Section Title Pace Total Dose . . . . . . . . . . . . . . . . . . . . . 3/4 2-13 Solid Radioactive Waste . . . . . . . . . . . . . . 3/4 2-14 3/4.3 RADIOLOGICAL ENVIRONMENTAL MONITORING Monitoring Program . . . . . . . . . . . . . . . . . 3/4 3-1 Interlaboratory Comparison Program . . . . . . . . . 3/4 3-7 BASES 3/4.1 INSTRrYNTATIOG . . . . . . . . . . . . . . . . . . . B3/4 1-1 Radioactive Liquid Effluent Instrumentation . . . . . B3/4 1-1 Radioactive Gaseous Effluent Instrumentation . . . . B3/4 1-1 3/4.2 RADIOACTIVE EFFLUENTS . . . . . . . . . . . . . . . . B3/4 2-1 Liquid Effluents - Concentration . . . . . . . . . . B3/4 2-1 Liquid Effluents - Dose . . . . . . . . . . . . . . . B3/4 2-1 Liquid Waste Treatment . . . . . . . . . . . . . . . B3/4 2-2 Gaseous Effluents - Dose Rate . . . . . . . . . . . . B3/4 2-2 Gaseous Effluents - Dose, Noble Gases . . . . . . . . B3/4 2-3 Gaseous Effluents - Dose - Tritium and Radionuclides in Particulate Foru . . . . . . . . . B3/4 2-3 Ventilation Exhaust Treatment . . . . . . . . . . . . B3/4 2-4 Total Dose . . . . . . . . . . . . . . . . . . . . . B3/4 2-4 Solid Radioactive Waste . . . . . . . . . . . . . . . B3/4 2-5 3/4.3 RADIOLOGICAL ENVIRONMENTAL MONITORING , . . . . . . . B3/4 3-1 Monitoring Program . . . . . . . . . . . . . . . . . B3/4 3 1 Interlaboratory Comparison Program . . . . . . . . . B3/4 3-1 5.0 ADMINISTRATIVE CONTROLS 5.1 Reporting Requirements . . . . . . . . . . . . . . . 5-1 5.1.1 Annual Radiological Environmental Monitoring Report . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1.2 Annual Radioactive Effluent Release Report . . . . . 5-1 5.1.3 Special Reports . . . . . . . . . . . . . . . . . . . 5-2 j 5.2 Major Changes to Radioactive Waste Treatment Systems . . . . . . . . . . . . . . . . . . . 5-3 Amendment 13 ii (December 1995) 1
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- t. TROJAN NUCLEAR PLANT l OFFSITE DOSE CALCULATION MANUAL l
TABLES PART I i Number Title 2.2 Frequency Notation 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation 4.1-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance
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Requirements 3.1-2 Radioactive Gaseous Effluent Monitoring Instrumentation j 4.1-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements
- 4. 2 l ~ Radioactive Liquid Waste Sampling and Analysis Program
! 4.2 2 Radioactive Gaseous Waste Sampling and Analysis Program l^
j 3.3-1 Radiological Environmental Monitoring Program 3.3 2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 4 3-1 Maximum Values for the Lower Limits of Detection (LLD) 5.1 1 Radiological Environmental Monitoring Program Summary l
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iii Amendment 11 l (December 1994)
- . 1 TROJAN NUCLEAR PLANT l OFFSITE DOSE CALCULATION MANUAL l CONTENTS PART II CALCULATIONAL METHODS AND PARAMETERS Section Title Pace
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . 1-1 2.0 LIQUID EFFLUENT DOSE CALCULATIONS . . . . . . . . . . 2-1
2.1 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . 2-1 2.2 ODCM CONTROL 3.2.1.1 . . . . . . . . . . . . . . . . . 2-4 2.3 ODCM CONTROL 3.2.1.2 . . . . . . . . . . . . . . . . . 2-6 2.3.1 Method 1 . . . . . . . . . . . . . . . . . . 2-6 2.3.2 Method 2 (Optional) . . . . . . . . . . . . . . . . 2-7 2.4 ODCM CONTROL 3.2.1.3 . . . . . . . . . . . . . . . . . 2-8 2.5 ODCM REPORTING REQUIREMENT 5.1.2 . . . . . . . . . . . 2-9 2.5.1 Ceneral Methodology . . . . . . . . . . . . . . . . . 2-9 2.5.2 Plant / Site Specific Assumptions . . . . . . . . . . . 2-9 3,0 CASEOUS EFFLUENT DOSE CALCULATIONS . . . . . . . . . . 3-1
3.1 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . 3-1 3.2 ODCM CONTROL 3.2.2.1 . . . . . . . . . . . . . . . . 3-3 1.3 ODCM CONTROL 3.2.2.2 . . . . . . . . . . . . . . 3-5 3.4 ODCM CONTROL 3.2.2.3 . . . . . . . . . . . . . . . . 3-6 1.4.1 Method 1 . . . . . . . . . . . . . . 3-6 l
1.4.2 Method 2 (Optional) . . . . . . . . . . . . . . . 3-6 1.5 ODCM CONTROL 3.2.2.4 . . . . . . . . . . . . . 3-7
- !.).1 Noble Cases . . . . . . . . . . . . . . . . 3-7 l
i 3.5.2 Particulates and Tritium . . . . . . . . . . . . . 3-7 i
5.6 ODCM CONTROL 3.2.2.5 - TOTAL DOSE . . . . . . . 3-8 1.6.1 Surveillance Requirements . . . . . . . 3-8 1.62 Methodology . , . . . . . . . . 3-8 I
( Amendment 11 iv l (December 1994)
TROJAN NUCLEAR PLANT j OFFSITE DOSE CALCULATION MANUAL CONTENTS PART II Section Title Pace 3.7 ODCM REPORTING REQUIREMENT 5.1.2 . . . . . . . . . . . 3-10 3.7.1 General Methodology . . . . . . . . . . . . . . . . . 3-10 3.7.2 Plant / Site-Specific Assumptions . . . . . . . . . . . 3-10 4.0 EFFLUENT MONITOR SETPOINT CALCULATIONS . . . . . . . . 4-1 4.1 LIQUID EFFLUENT MONITORS . . . . . . . . . . . . . . . 4-1 4.1.3 Liquid Radwaste Discharge P4cnitor (PRM-9) . . . . . . 4-2 4.2 GASEOUS EFFLUENT MONITORS . . . . . . . . . . . . . . 4-4 4.2.1 Setpoint Calculations for Noble Gas Effluent Channels (PRM 2C) . . . . . . . . . . . . . . . . . 4-4 4.2.2 Setpoint Calculations for Particulate Channels (PRMs 1A, 2A) . . . . . . . . . . . . . . . .. . . . 4-5 4.2.3 Condensate Demineralizer Building Effluent Monitoring
. . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 5.0 ENVIRONMENTAL MONITORING . . . . . . . . . . . . . . . 5-1 6.0 TROJAN PROCESS CONTROL PROGRAM FOR SOLID RADIOACTIVE WASTE . . . . . . . . . . . . . . . . . . 6-1 6.1 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . 6-1 l 6.2 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION . . . . . . . . . . . . . . . 6-1 6.2.1 Scope . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3.2 Program Elements . . . . . . . . . . . . . . . . . . . 6-1 6.3 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH-INTEGRITY CONTAINERS . . . . . . . . 6-3 v Amendment 14 (April 1997)
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TROJAN NUCLEAR PLANT .
OFFSITE DOSE CALCULATION MANUAL CONTENTS PART II l Section Title Pace 6.3.1 Scope . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.3.2 Program Elements . . . . . . . . . . . . . . . . . . . 6-3 i
6.4 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHER WASTES . . . . . . . . . . . . . . . 6-4 6.4.1 Scope . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.4.2 Program Elements . . . . . . . . . . . . . . . . . . . 6-4 6.5 SUPPORTING DOCUMENTS . . . . . . . . . . . . . . . . . 6-4 6.6 PROGRAM CHANGES . . . . . . . . . . . . . . . . . . . 6-6 I APPENDIX A DERIVATION OF NOBLE GAS FACTORS (K, L, M, N) . . . . . A-1 APPENDIX B DERIVATION OF PARTICULATE DOSE FACTORS . . . . .. . . B-1 APPENDIX C METEOROLOGY . . . . . . . . . . . . . . . . . . . . . C-1 APPENDIX D METHODOLOGY FOR DETERMINING DOSES TO PERSONS UTILIZING UNRESTRICTED AREAS WITHIN THE SITE EXCLUSION AREA BOUNDARY . . . . . . . . . . . . . . . D-1 APPENDIX E DELETED . . . . . . . . . . . . . . . . . . . . . . . E-1 APPENDIX F BASIS FOR CURIE RELEASE VALUES UTILIZED IN LIQUID EFFLUENT SURVEILLANCE REQUIREMENTS . . . . . . . . . . F-1 APPENDIX G QUALITY ASSURANCE REQUIREMENTS FOR THE ENVIRONMENTAL AND EFFLUENT MONITORING PROGRAM . . . . . . . . . . . G-1 Amendment 11 vi (December 1994)
TROJAN NUCLEAR PLMIT OFFSITE DOSE CALCULATION MANUAL TABLES PART II Number Title Section 2.0 2-1 Liquid Effluent Adult Ingestion Dose Factors (mrem /hr per Ci/ml)
Section 4.0 4-1 Historical Particulate Releases 4-2 Effluent Pathway Flow Rates Used for Method 1 Setpoint Calculations 4-3 Particulate Channel Detector Efficiencies Section 5.0 5-1 Sampling Locations and Frequency by Type Accendix A A-1 Noble Cas Dose Factors Anoendix B B-1 Dose Factors for Controlling Exposure Location B-2 Particulate Dose Factors Appendix C C-1 Historical Meteorology Data C2 liistorical Meteorological Data Anoendix D D1 Correction Factor for Persons Utilizing Unrestr.icted Areas Within the Site Exclusion Area Boundary Anoendix F F1 Calculated Aquatic Dose Due to Liquid Releases 1
vii Amendment 11 (December 1994)
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! 0FFSITE DOSE CALCULATION MANUAL I
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2-1 Liquid Radwaste Treatment System
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Title page 14 I
i 11 11 13 iii and iv 11 v 14 vi through viii 11 ix through x 14 PART I 1.0 1-1 14 2.0 2-1 11 2-2 14 2-3 through 2-5 11 2-6 14 3/4.0 3/4 0-1 and 3/4 0-2 11 3/4.1 3/4 1-1 11
- 3/4 1-2 through 3/4 1-4 14 3/4 1-5 and 3/4 1-6 11 j 3/4 1-7 and 3/4 1-8 14 l 3/4 1-9 11 3/4 1-10 14 3/4.2 3/4 2-1 14 3/4 2-2 11 3/4 2-3 and 3/4 2-4 14 3/4 2-5 11 3/4 2-6 through 3/4 2-13 14 !
3/4 2-14 11 l 3/4.3 3/4 3-1 13 l 3/4 3-2 through 3/4 3-4 11 3/4 3-5 14 3/4 3-6 11 i 3/4 3-7 14 3/4.1 3/4 1-1 11 3/4.2 B3/4 2-1 through B3/4 2-5 14 3/4.3 B3/4 3-1 13 5.0 5-1 through 5-2 14 5.2 5-3 11 5.3 5-4 14 5-5 14 PART II 1.0 1-1 11 3.0 2-1 14 2-2 9 2-3 1 2.2 2-4 14 l ix Amendment 14 l (April 1997)
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I LIST OF EFFECTIVE PAGES PGE-1021 offsite Dose calculation Manual Section Effective Paces : u e . ie e 2.2 2-5 10 2.3 2-6 10 2-7 14 2.4 2-8 10 2.5 2-9 10 Table 2-1 N/A 1 Figure 2-1 N/A 14 3.0 3-1 9 3.1 3-2 11 3.2 3-3 and 3-4 11 3.3 3-5 11 3.4 3-6 11 3.5 3-7 11 3.6 3-8 10 1 3-9 8 3.7 3-10 14 Figt, re 3-1 N/A 11 4.0 4-1 14 4.1 4-2 and 4-3 14 4.2 4-4 through 4-6 14 4-7 11 Table 4-1 through 4-3 N/A 11 5.0 5-1 11 Table 5-1 N/A 11 Figure 5-1 N/A 13 6.0 6-1 10 6.2 6-2 11 6.3 6-3 11 6.4 6-4 13 6.5 6-5 14 6.6 6-6 14 Appendix A A-1 3 Appendix A A-2 11 Table A-1 N/A 11 Appendix B B-1 through B-9 11 Table B-1 N/A 11 Table B-2 N/A 11 Appendix C C-1 11 Table C-1 and Table C-2 N/A 11 Appendix D D-1 8 Appendix D D-2 11 Table D-1 N/A 11 Appendix E E-1 11 Appendix F F-1 10 Table F-1, Sheet 1 N/A original Table F-1, Sheet 2 N/A 4 Appendix G G-1 11 Amendment 14 x (April 1997)
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1.0 INTRODUCTION
l The Offsite Dose Calculation Manual (ODCM) contains the Radioactive Effluent Controls Program required by Trojan Technical Specifications. This Program includes the Radiological Effluent Controls and their associated Surveillance Rsquirements, plus the methodology and parameters to be used for the i calculation of offsite doses resulting from radioactive gaseous and liquid l effluents, and for the calculation of gaseous and liquid effluent monitoring
! 'clarm/ trip setpoints. The implementation of this Program ensures compliance l l with the requirements of 10 CFR 50.36a, Subpart D of 10 CFR 20, Appendix I of i 10 CFR 50, and 40 CFR 190. The dose calculation methodology is based on l Plcnt-specific applications of the dose models contained in Regulatory Guida 1.109 (Rev. 1, 10/77) and/or the simplified models presented in I
NUREG 0133 (10/78).
Th3 ODCM contains the Radiological Environmental Monitoring Program required by Trojan Technical Specifications. This Program consists of monitoring at tions and sampling programs designed to confirm the dose estimates made undar the Radiological Effluent Controls Program and to meet the requirements of Appendix I to 10 CFR 50. The Radiological Environmental Monitoring Program of the ODCM also includes requirements to participate in an interlaboratory I comparison program.
Tha ODCM contains the Process Control Program (PCP) for solid radioactive l wectes which is required by Trojan Technical Specifications. The ODCM also
! contains administrative controls regarding the content of the annual l l Rtdiological Environmental Monitoring Report and the annual Radioactive I
! Effluent' Release Report which are required by Trojan Technical Specifications end administrative controls regarding major changes to radioactive waste treatment systems.
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1-1 Amendment 14 (April 1997)
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2.0 DEFINITIONS l
l Tha defined terms J.n this section appear in capitalized type and are cpplicable throughout these controls.
i ACTION 2.1 ACTION shall be that part of a Control that prescribes remedial measures rcquired under designated conditions.
CHANNEL FUNCTIONAL TEST 2.2 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel.as close to the primary sensor as practicable to verify OPERABILITY, including alarm and/or trip functions.
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CHANNEL CALIBRATION 2.3 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the I
channel output such that it responds with the necessary range and accuracy to i known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm cnd/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 2.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where ;
possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels l mancuring the same parameter. I 1
FREOUENCY NOTATION 2.6 The FREQUENCY NOTATION specified for the performance of Surveillance Rsquirements shall correspond to the intervals defined in Table 2.2.
l LIOUID RADWASTE TREATMENT SYSTEM 2.8 LIQUID RADWASTE TREATMENT SYSTEM is the system used to reduce radioactive msterials in liquid effluents by filtering, demineralizing, and providing holdup or decay of the radioactive wastes for the purpose of reducing the total radioactivity prior to release to the environment.
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2-1 Amendment 11 (December 1994)
1 2.0 DEFINITIONS MEMBER (S) OF THE PUBLIC 2.9 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
OFFSITE DOSE CALCULATION MANUAL 2.10 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and l
liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the i Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Trojan Technical Specifications and (2) descriptions of the information that should be included in the Annual I Radiological Environmental Monitoring and annual Radioactive Effluent Release Reports required by Trojan Technical Specifications.
OPERABLE - OPERABILITY 2.11 A system, 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, cooling or seal water, lubrication or other auxiliary 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).
PURGE - PURGING 2.13 PURGE or PURGING is the process of discharging air from the containment utilizing the Containment Purge Supply and Purge Exhaust Systems.
SITE BOUNDARY 2.14 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
SOLIDIFICATION l
l 2.15 SOLIDIFICATION shall be the convers.in of radioactive wastes from liquid i
systems to a form that meets shipping and burial ground requirements.
l l Amendment 14 2-2 (April 1997)
! 2.0 DEFINITIONS SOURCE CHECK l 2.16 A SOURCE CHECK shall be the qualitative assessment of channel response j when the channel sensor is exposed to either an installed detector check cource or to a background radiation level if background exceeds the installed chock source strength.
UNRESTRICTED AREA 2.17 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l
eccass to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any cras within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
VENTILATION FYMAUST TREATMENT SYSTEMS 2.18 The VENTI 1ATION EXHAUST TREATMENT SYSTEMS are those systems designed and
! installed to reduce the gaseous radioactive material in particulate form in affluents by passing the ventilation exhaust from the Fuel and Auxiliary l Buildings and the Containment purge through HEPA filters prior to release to tha environment. Such systems are not considered to have any effect on noble ges affluents, i
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I 2-5 Amendment 11 (December 1994)
TABLE 2.2 Frequency Notation Notation Freauancy D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W At least once per 7 days M At least once per 31 days Q At least once per 92 days R At least once per 18 months P Completed prior.to each release N/A Not applicable 1
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Amendment 14 2-6 (April 1997) l l
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3/4 CONTROLS AND SURVEILLANCE REOUIREMENTS 3/4.0 APPLICABILITY CONTROLS
-3.0.1 Controls and ACTION requirements shall be applicable during the conditions specified for each Control. I ;
1 3.0.2 Adherence to the requirements of the Control and/or associated ACTION within the specified time interval shall constitute compliance with the Control. In the event the control is restored prior to expiration of the cpacified time interval, completion of the ACTION statement is not required.
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3/4 0-1 Amendment 11 (December 1994)
a APPLICABILITY SURVEILLANCE REOUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the conditions
.specified for individual Controls unless otherwiss scated 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 the allowed surveillance interval, defined by Control 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 requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into a specified applicability condition shall not be made unless the Surveillance Requirement (s) associated with the Control have been performed within the stated surveillance interval or as otherwise specified.
l Amendment 11 3/4 0-2 (December 1994)
INSTRUMENTATION RADIOACTIVE LIOUID EFFLUENT INSTRUMENTATION CONTROL 3.1.1 The radioactive liquid effluent monitoring instrumentation channels chown in Table 3.1-1 shall be OPERABLE with their alarm / trip setpoints set to cnzure that the limits of Control 3.2.1.1 are not exceeded.
APPLICABILITY: As shown in Table 3.1-1.
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure the limits of Control 3.2.1.1 are met, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.1-1. With the inoperable channels not returned to OPERABLE status within 30 days, identify the cause of the inoperable channels in the annual Radioactive Effluent Release Report in lieu of any other report.
SURVEILLANCE REOUIREMENTS 4.1.1.1 The setpoints shall be determined in accordance with procedures as dsscribed in Part II of the ODCM and shall be recorded.
4.1.1.2 Each radioactive liquid effluent monitoring instrumentation channel chill be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the fraquencies shown in Table 4.1-1. I i
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s TABLE 3,1-1 Table Notation Sheet 2 of 2 ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Table 4.2-1, and l
l 2. At least two technically qualified members of the Facility Staff l independently verify the release rate calculations and discharge valving, and j 3. Initiate corrective action to return the channels to operable l status; or Otherwise suspend release of radioactive effluents via this pathway.
ACTION 2 Deleted. l l
l ACTION 3 Deleted. ]
l l ACTION 4 With the number of channels OPERABLE less than required by the l- Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least l once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
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INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION CONTROL 3.1.2 The radioactive gaseous effluent monitoring instrumentation channels chown in Table 3.1-2 shall be OPERABLE with their alarm / trip setpoints set to snsure that the limits of Control 3.2.2.1 are not exceeded.
APPLICABILITY: As shown in Table 3.1-2.
ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than the value shown in Table 3.1-2, adjust the setpoint to within the limit without delay or declare the channel inoperable.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.1-2. With the inoperable channels not returned to OPERABLE status within 30 days, identify the cause of the inoperable channels in the annual Radioactive Effluent Release Report in lieu of any other report.
SURVEILLANCE REOUIREMENTS 4.1.2.1 The setpoints shall be determined in accordance with procedures as dsceribed in Part II of the ODCM and shall be recorded.
4.1.2.2 Each radioactive gaseous effluent monitoring instrumentation channel chnll be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frsquencies shown in Table 4.1-2.
3/4 1-5 Amendment 11 f (December 1994) 1
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a TABLE 3.1-2 Sheet 3 of 3 Table Notation
- During releases via this pathway.
- Set to assure the limits of 3.2.2.1 are met.
ACTION 5 Deleted.
ACTION 6 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue prc11ded grab samples are taken at least once per 7 days and these sar.ples are analyzed for gross activity within 1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l I
ACTION 7 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken and analyzed once per 7 days when Plant release rates are anticipated to be stable. If transient release rates are anticipated, a sufficient number of additional samples will be taken to document release rates.
ACTION 8 With 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
0,CTION 9 With the number of channels OPERABLE less than required by the
, minimum channels OPERABLE requirement, effluent releases via this pathway will be stopped.
Amendment 14 3/4 1-8 (April 1996)
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3 TABLE 4.1 2 Sheet 3 of 3 During or prior to releases via this pathway.
include verification l1) Th2 CHANNEL FUNCTIONAL TEST shall, where applicable, th;t automatic isolation of the affected pathway and/or control room annunciator occurs if:
- c. Instrument indicates above alarm trip setpoint
- b. Instrument indicates a downscale failure
- c. Controls not in OPERATE mode (2) DelOted.
]3) CHANNEL FUNCTIONAL TEST consists of verification of sampler flow through tha sampler.
Amendment 11 3/4 1-11 (December 1994)
4 3 /4. 2 RADIOACTIVE EFFLUENTS 3/4.2.1 LTOUID EFFLUENTS CONCENTRATION CONTROL 3.2.1.1 The concentration of radioactive material released in liquid effluents from the site to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2.
APPLICABILITY: At all times. ,
ACTION:
- c. With the concentration of radioactive material released from the site to UNRESTRICTED AREAS exceeding the above limits, immediately restore l concentration to within the above limits. I SURVEILLANCE REOUTREMENTS 4.2.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.2-1.
4.2.1.1.2 The results of radioactive analysis shall be used with the l calculational methods in Part II of the ODCM to assure that the concentration at the point of release is limited to the values in control 3.2.1.1.
I l
I l
1 3/4 2-1 Amendment 14 (April 1997)
a e
TABLE 4.2-1 Sheet 1 of 2 Radioactive Liquid Waste Sampling and Analysis Program l
l Minimum Type of Sample / Lower Limit of I Liquid Release Sampling' Analysis Activity Detection (LLD)
Type Frequency Frequency Analysis (pCi/ml)a Batchd P P Grab Sample / 5x10-7b Waste Principal Gamma Release Emitters
l P Q Compositec / 5x10*8 Sr-89, Sr-90 P Q CompositeC / 1x10-6 Fe-55 Amendment 11 3/4 2-2 (December 1994)
___-_______________.1_______.
TABLE 4.2-1 Sheet 2 of 2 Table Notation
- a. The lower limit of detection (LLD) is defined in Table Notation a. of Table 4.3-1 of Control 4.3.1 with the exception of at. at in this case is the elapsed time between midpoint of sample collection and time of counting.
)
- b. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased provided that such an increase will not result in a discharge of that radionuclide which is greater than the effluent concentration value specified in 10 CFR 20, Appendix B, Table 2, Column 2, in the. diluted stream.
, c A composite sample is one in which the quantity of liquid sampled is
( proportional to the quantity of liquid waste discharged and in which the
! nathod of sampling employed results in a specimen which is representative of l the liquids released.
l l
- d. A batch release is the discharge of liquid wastes of a discrete volume. Prior j to sampling for analysis, each batch shall be isolated, and then thoroughly mixed to assure representativa sampling.
- e. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, co-58, co-60, Zn-65, Cs-134, Cs-137, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses should not be reported as l being present at the LLD' level. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the annual Radioactive Effluent Release Report. ,
l l
l f
3/4 2-3 Amendment 14 (April 1997)
RADIOACTIVE EFFLUENTS DQ.EE CONTROL 3.2.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive matorts]s in liquid effluents released to UNRESTRICTED AREAS shall be limited:
During any calendar quarter to s 1.5 mrem to the total body and to s 2.5 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of radioactive material in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days upon determination, pursuant to control 5.1.3, a Special Report in lieu'of any other report, which identifies the cause (s) for exceeding the limit (s) and defines the corrective actions to be taken to prevent recurrence and to reduce the releases to below the design objectives. This Special Report shall also include (1) the results of radiological analyses of the drinking water cource, and (2) the radiological impact in finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.
SURVEILLANCE REOUIREMENTS Dome Calculations. Cumulative doses due to liquid releases to' I.4.2.1.2.1
. UNRESTRICTED AREAS shall be determined in accordance with Part II of the ODCM at least once per 31 days when the cumulative liquid activity release, excluding tritium and dissolved gases,. exceeds 2.5 Ci/qtr. The cumulative liquid activity release will be determined at least once per 31 days.
Amendment 14. 3/4 2-4 (April 1997)
RADIOACTIVE EFFLUENTS LIOUID WASTE TREATMENT CONTROL 3.2.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be maintained and used to rsduce the radioactive materials in liquid wastes prior to their discharge whtn the liquid activity release excluding tritium and dissolved gases to UNRESTRICTED AREAS when averaged over a calendar quarter would exceed 1.25 Ci/qtr.
APPLICABILITY: At all times.
ACTION:
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, the following information shall be provided in the annual Radioactive Effluent Release Report:
- 1. Identification of equipment not OPERABLE and the reason for inoperability.
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status.
- 3. Summary description of action (s) taken to prevent a recurrence.
SURVEILLANCE REOUIREMENTS 4.2.1.3 Cumulative liquid activity releases excluding tritium and dissolved gases to UNRESTRICTED AREAS shall be determined at least once per 31 days.
l 3/4 2-5 Amendment 11 (December 1994)
RADIOACTIVE EFFLUENTS 3/4.2.2 GASEOUS EFFLUENTS DOSE RATE CONTROL 3.2.2.1 The dose rate to areas at or beyond the SITE BOUNDARY due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:
- a. The dose rate limit for noble gases shall be s 500 mrem /yr to the total body and s 3000 mrem /yr to the skin, and
- b. The dose rate limit for tritium and radionuclides in particulate form with' half-lives greater than 8 days shall be s 1500 mrem /yr to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With dose rate (s) exceeding the above limits, immediately. decrease the release rate to comply with the limit (s) given in control 3.2.2.1.
SURVEILLANCE REOUIREMENTS 4.2.2.1.1 The release rate of noble gases in gaseous effluents shall be such that 2.0 On, Ky 5 1 and 0.33 Qn, (Ly + 1.1 Ny) 1 1.
For Kr-85, the limiting release rate is s 0.176 Ci/sec.
4.2.2.1.2 The release rate of tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be such that 0.67 Qy Ri s 1 by using the results of the sampling and analysis program specified in Table 4.2-2.
4.2.2.1.3 The above release rates are determined in accordance with the methodology and parameters in Part II of the ODCM.
1 Amendment 14 3/4 2-6 (April.1997) l
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4 TABLE 4.2-2 Sheet 2 of 2 Table Notation i
1 f a. The lower limit of detection (LLD) is defined in Table Notation a. of l Table 4.3-1 of Control 4.3.1 with the exception of ot. At in this case is the elapsed time between midpoint of sample collection and time of
- counting.
l
- b. For certain radionuclides with low gamma yield or low energies, or for l
l certain radionuclide mixtures, it may not be possible to measure I radionuclides in concentrations near the LLD. Under these circumstances, l the LLD may be increased provided that such an increase will not result in l
a discharge of that radionuclide which is greater than the effluent concentration value specified in 10 CFR 20, Appendix B, Table 2, Column 1, l in plant effluents.
- c. Analysis shall also be performed following occurrences which could alter the mixture of radionuclides. !
i i d. Samples shall be taken and analyzed at the specified minimum frequency when there is a discharge through each release point.
l
- e. The principal gamma emitters for which the LLD specification will apply
! are exclusively the following radionuclicas: Kr-85 for gaseous emissions I and Mn-54, co-58, co-60, Zn-65, Cs-134, Cs-137, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, l together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the annual Radioactive Effluent Release Report.
l i
l l
l l
I Amendment 14 3/4 2-8 (April 1997) ;
I
. l RADIOACTIVE EFFLUENTS DOSE. NOBLE CASES CONTROL 3.2.2.2 The air dose in areas at or beyond the SITE BOUNDARY due to noble l gessa released in gaseous effluents shall be limited to the following:
During any calendar quarter to s 5 mrad for gamma radiation and 5 10 mrad for beta radiation.
APPLICABILITY: At all 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 upon determination, pursuant to Control 5.1.3, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to prevent recurrence and to reduce releases to below the design objectives.
SURVEILLANCE REOUIREMENTS _ _ _
4.2.2.2.1 Release Rate Calculations: The average release rate of noble gases from the site during any calendar quarter shall be such that:
50 Q 7 N < 1 or 25 Q Tv M y
- 1 For Kr-85, the limiting release rate is < 1.54E-3 Ci/sec.
4.2.2.2.2 The above release rates are determined in accordance with the msthodology and parameters in Part II of the ODCM at least once per 31 days.
3/4 2-9 Amendment 14 (April 1997)
l l
l l
BADIOACTIVE EFFLUENTS l
I DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FOPl4 CONTROL l I 3.2.2.3 The dose to a MEMBER OF THE PUBLIC from tritium and radionuclides in I I
l particulate form with half-lives greater than 8 days in gaseous effluents I released to areas at or beyond the SITE BOUNDARY shall be limited to the following:
During any calendar quarter to s 2.5 mrem to any organ.
APPLICABILITY: At all times.
l l
ACTION:
- a. With the calculated dose from the release of radioactive materials in particulate form, or radionuclides other than noble gases in gaseous
! effluents exceeding the above limit, prepare and submit to the Commission within 30 days upon determination, pursuant to control 5.1.3, a Special Report wnich identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to prevent recurrence and to reduce releases to below the design l objectives.
l g ILLANCE REOUIREMENTS 4.2.2.3.1 Release Rate Calculations: The average release rate of tritium and l l radionuclides in particulate form with half-lives greater than 8 days in l I gaseous effluents released to areas at or beyond the SITE BOUNDARY during any calendar quarter shall be such that:
j 100 Qy R < 1 l
l 4.2.2.3.2 The above release rates are determined in accordance with the methodology and parameters in Part II of the CDCM at least once per 31 days.
Amendment 14 3/4 2-10 (April 1997)
RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT CONTROL 3.2.2.4 The VENTILATION EXHAUST TREATMENT SYSTEMS shall be maintained and u Id to reduce radioactive materials in gaseous waste prior to their discharge wh:n the doses due to tritium and radionuclides in particulate form with half-livas greater than 8 days in gaseous effluent releases to areas at or beyond tha SITE BOUNDARY when averaged over a calendar quarter would exceed 1.25 mrem to any organ. The gaseous effluent air doses due to gaseous effluent releases l
to creas at or beyond the SITE BOUNDARY when averaged over a calendar quarter I ch-11 not exceed 2.5 mrad for gamma radiation and 5.0 mrad for beta radiation.
APPLICABILITY: At all times.
ACTION:
- a. With gaseous waste being discharged for more than 31 days without treatment and in excess of the above limits, discuss in the annual Radioactive Effluent Release Report the following information:
- 1. Identification of equipment not OPERABLE and the reason for ]
inoperability.
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status.
- 3. Summary description of action (s) taken to prevent a recurrence. ,
l SURVEILLANCE REOUIREMENTS 4.2.2 4.1 The average release rate of noble gases from the site during any l calendar quarter shall be such that:
100 0v7 Ny < 1 or 50 Qn, My < 1 For Kr-85, the limiting release rate is < 7.69E-4 Ci/sec.
3/4 2-11 Amendment 14 (April 1997)
RADIOACTIVE EFFLUENTS VENTILATION EXHAUS? TREATMENT SURVEILLANCE REOUIREMENTS (Continued) 4.2.2.4.2 The average release rate of tritium and radionuclides in l particulate form with half-lives greater than 8 days in gaseous effluents I released to areas at or beyond the SITE BOUNDARY during any calendar quarter shall be such that:
200 Q R <1 4.2.2.4.3 The above release rates are determined in accordance with the methodology and parameters in Part II of the ODCM at least once per 31 days.
Amendment 14 3/4 2-12 (April 1997)
a RADIOACTIVE EFFLUENTS TOTAL DOSE CONTROL 3.2.2.5 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except thyroid, which shall be limited to less than or Squel to 75 mrems.
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 3.2.1.2, 3.2.2.2, or 3.2.2.3, calculations should be made including direct radiation contributions from the reactor unit and from i outside storage tanks to determine whether the above limits of control l 3.2.2.5 have been exceeded. If such is the case, prepare and submit to I the Commission within 30 days a Special Report pursuant to l 10 CFR 20.2203 (a) (4) and Control 5.1.3. I l
SURVEILLANCE REOUIREMENTS 4.2.2.5.1 Cumulative dose centributions from liquid and gaseous effluents shall be determined in cecordance with Surveillance Requirements 4.2.1.2, 4.2.2.2 and 4.2.2.3, and in accordance with the methodology and parameters in Part II of the ODCM.
4.2.2.5.2 Cumulative dose contributions from direct radiation from the reactor unit and from outside storage tanks shall be determined in accordance with the methodology and parameters in Part II of the ODCM. This requirement is only applicable under the conditions set forth in control 3.2.2.5.a.
i i
3/4 2-13 Amendment 14 (April 1997)
RADIOACTIVE EFFLUENTS 3/4.2.3 SOLID RADIOACTIVE WASTE CONTROL 3.2.3.1 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.
APPLICABILITY: At all times.
ACTION:
- a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid !
radioactive wastes from the site.
SURVEILLANCE REOUIREMENTS I
4.2.3 The PROCESS CONTROL PROGRAM, as defined in the ODCM shall be used to verify that processed wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, and boric acid solutions) meet the shipping and l burial ground requirements with regard to solidification and dewatering.
Amendment 11 3/4 2-14 (December 1994)
[
3 /4 . 3 RADIOLOGICAL ENVIRONMENTAL MONITORING l l
l 3 /4 . 3 .1 MONITORING PROGFAM
/
CONTROL i
3,3.1 A radiological environmental monitoring program as specified in Tablo 3.3-1 shall be conducted in accordance with written procedures.
(Rtductions in tha scope of this program shall be discussed with the Oregon Stato Health Division before implamenting the reduction.) )
I APPLICABILITY: At all times.
ACTION:
- c. With the radiological environmental monitoring program not being l conducted as specified in Table 3.3-1 and Part II of the ODCM, prepare and submit to the Commission, in the annual Radiological Environmental Monitoring Report, a description of the reasons for not conducting the program as required and the plans for preventing recurrence.
(Deviations are permitted from the required sampling schedule if i specimens are unobtainable due to hazardous conditions, seasonal )
unavailability, or to malfunctions of automatic sampling equipment.
l If the latter, every effort shall be made to complete corrective j action prior to the end of the next sampling period.) l 1
- b. With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 3.3-1 exceeding the limits of Table 3.3-2 when averaged over any calendar quarter, prepare l and submit to the Commission within 30 days from receipt of analysis results for the affected calendar quarter, a Special Report which !
l includes an evaluation of any release conditions, environmental I factors or other aspects which caused the limits of Table 3.3-2 to be exceeded. When more than one of the radionuclides in Table 3.3-2 are detected in the sampling medium, this report shall be submitted if:
Concentration (1) Concentration (2)
Limit Level (1) Limit Level (2)
This report is not required if the measured level of radioactivity was l not the result of plant effluents; however, in such an event, the condition shall be reported and described in the annual Radiological Environmental Monitoring Report.
3/4 3-1 Amendment 11 (December 1994)
a 1 e I
. 1 l
CONTROL (Continued) . !
l When Radienuclides other than those in Table 3.3-2 are detected and are the result of plant effluents, this report shall be submitted if the l
potentir.1 for annual dose to an individual is equal to or greater than i the calendar year limits of Control 3.2.1.2, 3.2.2.2 and 3.2.2.3.
SURVEILLANCE REOUTREMENTS 4.3.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.3-1 from the locations shown in Part II of the ODCM and chall be analyzed pursuant to the requirements of Tables 3.3-1 and 4.3-1.
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l l Table 3.3-2
! Reporting Levels for Radioactivity Concentrations in Environmental Samples Airborne l Particulates Water Analysis (pci/m 3) (pci/1)
H-3 SX10' 2X10'(*)
Mn-54 5X10 2 Co-58 SX10 2 Co-60 2X10 2 Zn-65 2X10 2 1
Zr-Nb-95 2X102 Cs-134 4 20 Cs-137 6 20 k
(a) For drinking water samples. This is a 40 CFR 141 value.
( l l
l i
i Amendment 11 3/4 3-4 I (December 1994) I
. 1 TABLE 4.3-1 Sheet 1 of 2 Maximum Values for the Lower Limits of Detection (LLD)*
Airborne particulate Sediment Water Analysis (pCi/m 3) (pci/kg, dry) (pCi/1) gross beta 1X10 2 4 (i b) 3H 2000 (1000b) 54Mn 15 50, 60Co 15 65Zn 30 95Zr 30 95Nb 15 13'Cs 5X10-2 150 15 137Cs 6X10-2 180 18 3/4 3-5 Amendment 14 (April 1997) i
TABLE 4.3-1 Sheet 2 of 2 Igble Notation a- The LLD is defined, for the 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 51 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) (V) (2.22) (Y) (e-hat) where LLD is the "a priori" lower limit of detection as defined above (as pCi 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 transformation per minute per picoeurie Y is the fractional radiochemical yield (when applicable)
A is the radioactive decay constant for the particular radionuclide at is the elapsed time between sample collection (or end of the sample collection period) and time of counting It should be recognized that the LLD is defined as an a criori (before the fact) limit representing the capability of a measurement system and not as an a nosteriori (after the fact) limit for a particular l measurement.
1 b- LLD for drinking water.
Amendment 11 3/4 3-6 (December 1994)
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.3.2 INTERLABORATORY COMPARISON PROGPJJ4 CONTROL 3.3.2 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by NRC.
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 Monitoring Report.
SURVEILLANCE REGUIREMENTS 4.3.2 The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the annual Radiological Environmental Monitoring Report pursuant to Control 5.1.1.
/
V 3/4 3-7 Amendment 14 (April 1997)
3 /4 .1 INSTRUMEWATION t
BASES 3 /4 .1.1 RADIOACTIVE LIOUID EFFLUEW INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid !-
effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with Part II of the CDCM to ensure that the alam/ trip will occur prior to exceeding the limits of 10 CFR 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 50.
3/4.1.2 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases, however, in no case will the limits of 10 CFR 20 be exceeded. The alarm / trip setpoints for these instruments shall be calculated in accordance with Part II of the ODCM. Tiva /
OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFA 50.
\
B3/4 1-1 Amendment 11 (December 1994)
3/4.2 RADIOACTIVE EFFLUENTS BASES /
3/4.2.1 LIOUID EFFLUENTS 3/4.2.1.1 CONCENTRATION This Control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to UNRESTRICTED AREAS will be less than the concentration levels specified in Appendix B, Table 2, Column 2, to 10 CFR 20. This limitation provides reasonable 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 50, to a MEMBER OF THE PUBLIC and (2) restrictions authorized by 10 CFR 20.1301(e) . This Control does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301(a). G' 3/4.2.1.2 DOSE This control is provided to implement the requirements of Section II. A, III.A and IV.A of Appendix I, 10 CFR Part 50.Section II.A of Appendix I, 10 CFR 50 specifies design objective dose to an individual defined as a MEMBER OF THE PUBLIC, from radioactive materials in liquid effluents released to UNRESTRICTED AREAS will be limited during any calendar year to s 3 mrem to the total body. The design objective for any organ will be limited to s 5 mrem during any calendar year in accordance with PGE Agreement with intervenors dated May 1972.Section IV.A to 10 CFR 50 specifies the limiting condition for operation as one-half the design objective annual exposure in one calendar quarter.
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 radioactive materials in liquid effluents will be kept "as low as reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations that are in Part II of the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of .
Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Part II of the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consistent with the methodology provided in Regulatory Guides 1.109 (Rev. 1) v and 1.113 (Rev. 1). The intermediate surveillance value of 2.5 Ci/qtr excluding tritium and noble gases is a release rate which has been B3/4 2-1 Amendment 14 (April 1997)
. 1 9
RADIOACTIVE EFFLUENTS BASES shown during 7 years of Trojan operation to be significantly below the Control value of 1.5 mrem /qtr total body and 2.5 mrem /qtr to any organ. Refer to PGE-1015, Annual Operating Report of Trojan Nuclear Power Plant for 1977, 1978, 1979, 1980, 1981, 1982 and 1983.
3/4.2.1.3 LIOUID WASTE TREATMENT This Control ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment.
The requirements 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 Control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and Design Objective Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as one quarter of the annual design objective set forth in Section II.A of Appendix I, 10 CFR 50, for liquid effluents (3 mrem /yr whole body; 5 mrem /yr maximum organ per PGE Agreement with Intervenors dated May 1972). The dose calculational procedures specified in Part II of the ODCM include sufficient factors of conservatism to ensure that the sum of both treated and untreated releases will not result in doses exceeding the design objectives. The surveillance value of 1.25 Ci/qtr excluding tritium and noble gases is a release rate which has been shown during the first 7 years of Trojan operation to be significantly below the Control value of 0,75 mrem /qtr total body and 1.25 mrem /qtr to any organ.
3/4.2.2 GASEOUS EFFLUENTS 3/4.2.2.1 DOSE RATE This Control provides reasonable assurance that radioactive material discharged in gaseoue effluents will not result in the exposure of a MEMBER OF f THE PUBLIC at or beyond the SITE BOUNDARY in excess of the design objectives
! of Appendix I to 10 CFR Part 50. It provides operational flexibility for l releasing gaseous effluents to satisfy the Section II.A and II.C design objectives of Appendix I to 10 CFR Part 50. For MEMBERS OF THE PUBLIC who nay, at times, be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta 3
dose rates above background to a MEMBFD OF THE PUBLIC at or beyond the SITE BOUNDARY to s 500 mrem / year to the total body or to s 3000 mrem / year to the
- skin. These release rate limita also restrict the corresponding dose rate l
above back.round to any organ to s 1500 mrem / year. This Control does not l a f ."tct the requirement to comply with the annual limitations of it C) :< 2 0.13 01 (a) .
Amendment 14 B3/4 2-2 (April 1997)
RADIOACTIVE EFFLUENTS BASES 3/4.2.2.2 DOSE. NOBLE GASES This Control is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR 50. As outlined in Section II.B of Appendix I, 10 CFR 50, the design objective air dose in areas at or beyond the SITE j BOUNDARY due to noble gases released in gaseous effluents will be limited l during any calendar year to s 10 mrad for gamma radiation and s 20 mrad for j bsta radiation. As outlined in Section IV.A to Appendix I to 10 CFR 50, the j limiting condition for operation is specified as one-half the design objective '
cnnual exposure in one calendar quarter. The ACTION statements provide the rsquired operating flexibility and at the same time implement the guidas set forth in Section IV.A of Appendix I to assure that the releases of radioactive i material in gaseous effitents will be kep*. "as low as is reasonably )
cchievable". The Surveillance Requir<sments implement che requirements in S2ction III.A of Appendix I, which 7;equires that eniculational procedures be bnzed on models and data such that che actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
Tha dose calculations established in Part II of the ODCM for calculating the .
dores due to the actual release rates of radioactive noble gases in gaseous l cffluents will be consistent with the methodology provided in Regulatory Guides 1.109 (Rev. 1) and 1.111 (Rev. 1). The ODCM equations provided for I datermining the air doses at the SITE BOUNDARY will be based upon the historical average atmospheric conditions.
3/4.2.2.3 DOSE - TRITIUM 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 CFR 50. As outlined in Section II.C of Appendix I, 10 CFR 50, the design objective dose to an individual from tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at or beyond the SITE BOUNDARY will be limited during any calendar year to s 15 mrem to any organ. This value is further reduced to 5 mrem / year maximum organ dose per PGE Agreement with Intervenors, dated May 1972. As outlined in Section IV.A to Appendix I, 10 CFR 50, the limiting condition for operation is specified es one-half the design objective annual exposure in one calendar quarter.
Th2 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 radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". The ODCM calculational B3/4 2-3 Amendment 14 (April 1997) s'
l
. i RADIOACTIVE EFFLUENTS BASES 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 an individual through appropriate pathways is unlikely to j be substantially underestimated. The dose calculations established in Part II ]
of the ODCM for calculating the doses due to the actual release rates of l particulates in gaseous effluents will be consistent with the methodology l provided in Regulatory Guides 1.109 (Rev. 1) and 1.111 (Rev. 1). The ODCM equations provided for determining these doses will be based upon the historical average atmospheric conditions. The release rate specifications for tritium and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man, in areas at or beyond the SITE BOUNDARY. The pathways which are examined in the development of these calculations are: (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 milk and '
meat by man, and (4) depe,sition on the ground with subsequent exposure of man.
3/4.2.2.4 VENTILATION EXHAUST TREATMENT This control ensures that the ventilation exhaust treatment systems will be used 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 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and design objective Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the systems were specified as one i
quarter of the annual design objective set forth in Sections II.B and II.C of Appendix I, 10 CFR 50, for gaseous effluents (20 mrad /yr beta air dose; 10 mrad /yr gamma air dose; 5 mrem /yr maximum organ dose per PGE Agreement with intervenors, dated May 1972). The dose calculational procedures specified in Part II of the ODCM include sufficient factors of conservatism to ensure that l the sum of both treated and untreated releases will not result in doses
! excaeding the design objectives.
l 3/4.2.2.5 TOTAL DOSE i
l This Control is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20.1301(d). The Control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which chall be limited to less than or equal to 75 mrems. It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if radioactive effluents remain within twice the dose design ob]ectives of Appendix I, and if direct radiation doses (including outside storage tanks, etc.) are kept small.
Amendment 14 B3/4 2-4 (April 1997)
F .
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! RADIOACTIVE EFFLUENTS BASES
. 3 /4 . '2 . 3 SOLID RADIOACTIVE WASTE
- Thio control ensures that radioactive wastes that are transported from the
- cite shall meet the solidification requirements specified by the burial ground licsnse of the respective states to which the radioactive material will be chipped.
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l B3/4 2-5 Amendment 14 (April 1997)
. 9 3/4.3 RADIOLOGICAL ENVIRONMENTAL MONITORING i
! BASES l l
! i 3/4.3.1 MONITORING PROGRAM The radiological monitoring program required by this Control provides l mencurements of radiation and of radioactive materials in those exposure l pnthwnys and for those radionuclides which lead to the highest potential
, rediction exposures of individuals resulting from station operation. This i monitoring program thereby. supplements the radiological effluent monitoring :
program by verifying that the measurable concentrations of radioactive I
matorials and levels of radiation are not higher than expected on the basis of l l ths cffluent measurementu and modeling of the environmental exposure pathways, j l
Th3 L1Ds for drinking water meet the requirements of 40 CFR 141.
3/4-3.2
. INTERLABORATORY COMPARISON PROGRAM
! Th3 rcquirement for participation in an Interlaboratory Comparison Program is
{provided.toensure.thatindependentchecksontheprecisionandaccuracyof l th2 miasurements of radioactive material in environmental sample matrices -are l psrformed as part of a quality assurance program for environmental monitoring
- in order to demonstrate that the results are reasonably valid.
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B3/4 3-1 Amendment 13 (December 1995) 1 I
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l 5.0 ADMINISTRATIVE CONTROLS 5.1 REPORTING REOUIREMENTS 5.1.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT I
Tha Annual Radiological Environmental Monitoring Report shall include I summaries, interpretations, and statistical evaluation of the results of the I radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as cppropriate), and previous environmental surveillance reporto and an l casassment of the observed impacts of the plant operation on the environment. I If harmful effects or evidence of irreversible damage are detected by the j monitoring, the report shall provida an analysis of the problem and a planned !
course of action to alleviate the problem. ;
i Th3 Annual Radiological Environmental Monitoring Report shall include I cummarized and tabulated results in the format of Table 5.1-1 of all rcdiological environmental samples taken during the report period. In the sysnt that some results are not available for inclusion in the report, the rsport shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a cupplementary report.
Tha reports shall also include the followings a summary description of the Radiological Environmental Monitoring Program including sampling methods for occh sample type, size and physical characteristics of each sample type, simple preparation methods, analytical methods, and measuring equipmer.t used; ,
o map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program required by Control 3.3.2.
Any changes to the ODCM made during the reporting period, shall be reported as provided in Trojan Technical Specifications.
5.1.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT I'
The Radioactive Effluent Release Reports shall include a summary of the quintities of radioactive liquid and gaseous effluents and solid waste .
ralcased from the unit as outlined in Regulatory Guide 1.21 (Rev. 1),
"Marsuring, Evaluating, and Reporting Radioactivity in $alid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Ef flaents from Light-Water-cooled Nuclear Power Plants", with de : summarized on a quarterly basis following the format of Appendix B thereof.
5-1 Amendment 14 (April 1997)
5.0 ADMINISTRATIVE CONTROLS 5.1 REPORTING REOUIREMENTS 5.1.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The Radioactive Effluent Release Report shall include an assessment of the radiation doses from radioactive effluents to MEMBERS OF THE PUBLIC due to their activities in UNRESTRICTED AREAS during the report period. All assumptions used in making these assessments (e.g., specific activity, cxposure time and location) shall be included in these reports.
The Radioactive Effluent Release Report shall include a copy of all licensee
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I event reports required by 10 CFR 50.73 (a) (2) (viii) .
The Radioactive Effluent Release Report shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory 9 Guide 1.21. In addition, thr, maximum noble gas gamma air and beta air doses k shall be evaluated in areas at or beyond the SITE BOUNDARY. The assessment of radiation doses shall be parformed in accordance with Part II of the ODCM.
5.1.3 SPECIAL REPORTS The originals of Special Reports shall be submitted to the Document Control Desk with a copy sent to the Regional Administrator, NRC Region IV, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference control,
- a. Radioactive Effluents, Controls 3.2.1.2, 3.2.2.2, 3.2.2.3, and 3.2.2.5.
- b. Radiological Environmental Monitoring, Control 3.3.1.
amendment 14 5-2 AApril 1997)
5.0 ' ADMINISTRATIVE CONTROLS
! 5.2 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMEN* SYSTEMS (Liquid, Gaseous and Solid)
Licansee initiated major changes
- to the radioactive waste treatment systems i (liquid, gaseous and solid):
l l e. A summary description of the change including discussion of the equipment, components and processes involved shall be reported to the l Commission. The change shall be reviewed and approved in accordance l 'with plant procedures.
l
- b. The following information shall be available for review:
- 1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; 1
l 2. Sufficient information to totally support the reason for the change; l
- 3. A 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 l
radioactive materials in liquid and gaseous effluents and/or quantity
! of solid waste that differ from those previously estimated in the license application and amendments thereto; I 5. An evaluation of the change which shows the expected maximum exposures to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;
- 6. An estimate of the exposure to plant operating personnel as a result of the change; and
- 7. Documentation of the fact that the change was reviewed and approved in accordance with plant procedures.
- Major changes to the radioactive waste treatment systems are permanent changes which would alter the capacity of handling radioactive wastes or differ in the method of treatment.
5-3 Amsndment 11 (December 1994)
5.3 CHANGES TO THE ODCM Changes to the ODCM shall be documented and records of reviews performed shall be retained. This documentation shall contain sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s); and a determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, and 40 CFR 190, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
Changes to the ODCM shall become effective after review and approval by an Independent Safety Reviewer and the approval of the Plant General Manager or designee; and shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as part of, or concurrent with, the Radiological Environmental Monitoring Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicated the date (i.e., month and year) the change was implemented.
i l
Amendment 14 5-4 (April 1997)
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i 1,0 TNTRODUCTIUN Because the introduction has been moved in front of Part I, this page has been intentionally deleted I
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l l
1 1-1 Amendment 11 (December 1994)
4' I
2.0 LIQUID EFFLUENT DOSE CALCULATIONS
2.1 INTRODUCTION
Cumulative quarterly doce contributions due to radioactive liquid effluents rolossed to UNRESTRICTED AREAS will be determined at least once per 31 . lays )
whzn the cumulative liquid activity release, excluding tritium and dissolved l gnose, exceeds 2.5 Ci/ quarter. These dose contributions will be calculated l
l for all radionuclides identified in liquid effluents released to the UNRESTRICTED AREA using the following general equation (Reference; NUREG-0133, pg. 15) :
E E I (2-1) 0 3"i fij t t C,gFg wh2re D3 , the cumulative quarterly dose commitment to any organ j, fromliquideffluentfortotaltimeperiod[ATg, in mrem l ATg_= the length of time over which Cg i and Fg are averaged, in hours Cg i - average concentration on nuclide i, during time period ATg, l in pCi/ml. The term C ig is the undiluted concentration of f radioactive material in liquid waste determined in accordance with Table 4.2-1 of Part I of the OtVM l l
A 13 = ingestion dose factor for any organ j, for each identified nuclide i, listed in Table 2-1, in mrem /hr per pCi/ml Fg = the near field average dilution factor for C gi during any liquid release.
2-1 Amendment 1.
(April 1997)
l The term Fg, the near field average dilution factor, is determined as follows for time period ATg:
liquid radioactive waste discharge volume 7 ,
t total Plant discharge volume x Plant dilution factor The Plant dilution factor accounts for mixing effects of the dilution pipe.
This value is determined in accordance with NUREG-0133, Page 16, as equal to:
1000 cfs = 19.26 average total Plant discharge The average total Plant discharge <>f 23,803 gpm is the historical average for l the years 1976-1985.
l The term A i j, the ingestian dose factors for any organ, are tabulated in Table 2-1. For simplicity and conservatism, a single maximum organ dose factor for each nuclide was calculated using the critical organ for each nuclide. The following equation was used in calculating the ingestion dose factors (
Reference:
MUREG-0133, pg. 16 ) :
U 7
A y =K o - " + U,B F, D F, u
(2-2) where A ij = composite does parameter for total body or maximum organ i
of an adult for nuclide i, in mrem /hr per pCi/ml ko = conversion factor, 1.14 x 105 , go6 pCi/pf A x 103 ul/kg + 8760 hr/yr Uw = 730 kg/yr, adult maximum annual water consumption rate.
(from Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5)
Amendment.9 2-2 (February 1993)
'4 Up = 21 kg/yr, adult maximum annual fish consumption rate (from Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5)
BFi = . bioaccumulation factor for nuclide i, in fish, pCi/kg per PCi/l (from Regulatory Guide 1.109,-Rev. 1, 10/77, Table A-1) 1 i
I DFi = dose conversion factor for nuclide i, for adults - total body or maximum organ in mrem /pCi (from Regulatory Guide 1.109, Rev. 1, 10/77, Table E-11) i Dw = dilution factor from the near field concentration to the potable water intake, 230 = 230,000 cfs average river flow + 1000 cfs near field dilution flow.
l 2-3 Amendment 1 (December 1984)
a 2.2 ODCM CONTROL 3.2.1.1 This section will be used to demonstrate compliance with ODCM Control 3.2.1.1 by providing the calculational methods to use with the results of radioactive analysis required by ODCM Surveillance Requirement 4.2.1.1.1.
Once the results of a radioactive analysis are obtained, the fractional ECV (f) should be calculated using the equation:
E i f=1 ECVi (2-3) where Ai = concentration of nuclide i in sample, pCi/ml ECVi = 10 CFR 20, Appendix B, Table 2, Column 2 effluent concentration value for nuclide i, pCi/ml The resulting fractional ECV must be adjusted for Plant dilution using the following equation:
fxF e (2-4)
P where C = fraction of ODCM 3.2.1.1 limit Fe = liquid radioactive waste discharge flow rate prior to dilution, gpm Fp = total Plant dilution flow rate, gpm Releases comply with ODCM Control 3.2.1.1 if the value of C is s1.0.
Amendment 14 2-4 (April 1997)
Nuclides which require analysis of monthly or quarterly c.omposite samples (e.g.. H 3, Fe-55, Sr-89, Sr-90) are not considered in the calculation required by ODCM Control 3.2.1.1 at the time of the release. When the results from these analyses are available, they will be used to confirm that those nuclides, averaged over the sample period, did not cause violation of coCM Control 3.2.1.1.
4 l
l 2-5 Amendment 10 (December 1993)
2.3 ODCM CONTROL-3.2.1.2 This section will be used to demonstrate compliance with ODCM Control 3.2.1.2 at least once per 31 days when the cumulative liquid activity release, excluding cricium and dissolved gases, exceeds 2.5 Ci/ quarter.
The intermediate surveillance value of 2.5 Ci/ quarter, excluding tritium and dissolved gases, is a release rate which has been shown during 7 yr of Trojan operation to be significantly below the ODCM Control value of 1.5 mrem / quarter total body and 2.5 mrem / quarter to any organ. This is demonstrated in Appendix F.
2.3.1 METHOD 1 The following Plant specific applications of Equation 2-1 will be used in Method 1 should the quarterly release exceed 2.5 Ci/ quarter:
b Total Body E E g
AT# A tr3xQs (2-5)
D73 i t
19.26V f Maximum Orean ATg C Dg=E g g i Agg(x Qgg (2-6) 19.26V g where .
DTB - cumulative quarterly total body dose incurred to date, mrem DMO -
cumulative quarterly maximum organ dose incurred to date, arem ATE
- the fth time period in a calendar quarter over which the dose is evaluated, hr (ie, dose for 7-day period has ATg - 7 days)
Aine ndme n t 10 2-6 I (December 1993)
I L _
I V) = volume of total Plant discharge flow for time o} , ml 19.26 = Plant-specific dilution factor A = total body dose parameter for nuclide i, mrem /hr per TBi pCi/ml A = maximum organ dose parameter for nuclide i, mrem /hr per moi pCi/ml l Qgi - activity released of nuclide i, over time period agp, i
pCi.
l l
l Nuclides which require analysis of monthly or quarterly composite samples (sg, Sr-89, Sr-90) are not considered in the calculation required by ODCM Surveillance Requirement 4.2.1.2.1 at the time of release. When the results from these analyses are available, they will be used to confirm 1
i that those nuclides, averaged over the sample period, did not cause the I
total liquid release to exceed 2.5 Ci/ quarter or the total calculated doses to exceed ODCM Control 3.2.1.2.
1 2.3.2 METHOD 2 fontional)
Should the dose limits of ODCM Control 3.2.1.2 be exceeded using Method 1, a more accurate dose calculation may be made using the methodology in Rtgulatory Guide 1.109 (Rev. 1, 10/77) to demonstrate compliance.
I i
l l
2-7 Amendment 14 (April 1997)
2.4 ODCM CONTROL 3.2.1.3 This section is used to demonstrate compliance with ODCM Control 3.2.1.3 at least once per 31 days.
The surveillance value of 1.25 Ci/ quarter, excluding tritium and dissolved gases, is a release rate which has been shown during the first 7 yr of Trojan operation to be significantly below the ODCM Control value of 0.75 mrem / quarter total body and 1.25 mrem / quarter to any organ. Should this surveillance value be exceeded, the radwaste treatment systems will be used.
A flow diagram of the liquid radwaste treatment system, as applicable to ODCM Control 3.2.1.3, is shown in Figure 2-1.
Amendment 10 2-8 (December 1993)
2.5 ODCM REPORTING REOUIREMENT 5.1.2 This section describes the method that will be used to calculate doses from liquid effluents, .as required by ODCM Reporting Requirement 5.1.2 (annual Radioactive Effluent Release Report).
2.5.1 GENERAL METHODOLOGY The models of Regulatory Guide 1.109 (Rev. 1, 1977) will be utilized, incorporating Trojan site-specific modeling parameters, to compute doses from liquid effluents for this ODCM Control. In addition to the four principal Regulatory Guide 1.109 liquid effluent dose pathways, a PCE developed swimming immersion dose pathway has been added to inc'lude radiation exposure to swimmers in the Columbia River. The PGE computer codes utilized in these calculations are documented, validated and controlled in accordance with written, quality-related procedures.
2 5.2 PLANT / SITE SPECIFIC ASSUMPTIONS j Hydrologic dilution factors will be based on actual river flow rates and effluent flow races during the reporting period. Drinking water and agricultural exposure pathways will assume dilution into the full river j
flow. Other exposure pathways will assume dilution into the Plant mixing ,
I zone, which is defined as that portion of the river from the Oregon shore I to a point 300 ft from the end of the active region of the diffuser pipe.
2-9 Amendment 10 (December 1993)
1-
- 6 i
l TABLE 2-1 LIQUID EFFLUENT ADULT INGESTION DOSE FACTORS j (ares /hr per pCi/ml) ;
)
Total Maximum Total Maximum Body Organ Body Organ )
Og Nuclide -
i i Nuclide i H-3 2.7E-1 2.7E-1 Ag-110m 0 3.6E+2 i
l~ F 18 1.7E+0 1.6E+1 Sn-113 3.3E+3 5 .'7 E+4 Ns-22 4.2E+3 4.2E+3 Sb-124 3.1E+0 2.2E+2 -
Ne 24 4.1E+2 4.1E Sb-125 1.2E+0 5.5E+1 Cr-51 1.3E+0 3.3E+2 Sb-126 1.2E+0 2.6E+2 Mn 54 8.4E+2 1.4E+4 Te-132 1.5E+3 7.4E+4 Mn.56 2.0E+1 3.6E+3 1-131 1.3E+2 7.1E+4 Fe 55 1.1E+2 6.6E+2 I-132 6.9E+0 6.9E+2 Fe'59
- 9.4E+2 8.2E+3 I-133 2.8E+1 1.4E+4 Co 57 3.5E+1 5.4E+2 I-134 3.8E+0 1.9E+2 Co 58 2.1E+2 1.9E+3 I-135 1.6E+1 2.8E+3 Co 60 5.7E+2 4.9E+3 Cs-134 5.8E+5 7.1E+5 Cu-64 4.7E+0 8.6E+2 Cs-136 8.9E+4 1.3E+5 Zn 3.4E+4 7.4E+4 Cs-137 3.5E+5 .5.3E+5 Rb 86 4.8E+4 1.1E+5 Cs-138 2.6E+2 5.3E+2-Rb 88 1.6E+2 2.9E+2 Ba-140 1.4E+1 '4.2E+2 Sr 89 '6.4E+2 2.3E+4 La-140 0 5.6E+3 Sr-90 1.'4 E+ 5 5.5E+5 Ce-141 0 6.7E+1 Zr 95 0 2.6E+2 Co-143 0 1.3E+2 Zr 97 0 8.7E+2 Ce-144 0 4.6E+2 Nb 95 1. 4 E+ 2 1.6E+6 Pr-147 0 2.5E+3 Mo 99 2.0E+1 2.5E+2 Nd-147 0 2.1E+3 -
Tc 99M. 0 1.5E+1 W-187 8.7E+1 8.1E+4 l Ru 103 2.0E+0 5.3E+2 Np-239 0 5.9E+2 Ru 106 8.5E+0 4.4E+3 i
Note: Zaro in this table is <1.0 except H 3. ,
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Amendment 1 (December 1984) l
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3.0 CASE 0VS EFFLUENT DOSE CALCULATIONS
3.1 INTRODUCTION
The noble gas dose rate contributions may be determined using the follow.
ing general equations (adapted from NUREC-0133, Sections 5.2.1 and 5.3.1):
Camma air dose rate, D{. mrad /yr D{=1000fN g xQ TV g (3-1)
Beta air dose race, D , mrad /yr Df-1000fM g xQ (3-2) l Skin dose race, D,, mrem /yr D, - 1000 gD (Lg + 1.1 x N.)
g x Qg (3-3) i Total body dose rate, DTB' "#**!7#
( }
D TB
~
i*STV g where Kg - total body dose factor due to gamma emissions for nuclide i, rem /yr per Ci/see Lt .- skin dose factor due to beta emissions for nuclide i, rem /yr per CL/sec Mg - air dose factor due to beta emissions for nuclide i, rad /yr per CL/sec 31 Amendment 9 (February 1993)
Ng - air dose factor due to gamma emissions for nuclide i, rad /yr (
per Ci/sec (note that these are " air" rads not "cissue" rads)
QTV t
- n ble gas activity release rate of nuclide i, Ci/sec 1000 - constant, mrad / rad or mrem / rem 1.1 . constant, the average ratio of tissue to air energy absorption coefficients with the units of rem /" air" rad.
Derivation of Kg, L ,iM ,.and i Ni are presented in Appendices A and C.
These values are listed in Table A-1.
The critium, and particulate (Tl/2 > 8 days) dose contributions may be determined using the following general equation:
(3-5)
D IPC-1000fR g x Qg where DIPC - dose rate at controlling exposure location, mrem /yr Rt - dose f actor for nuclides other than noble gases 'at the the site boundary for critical organ and age group, rem /yr per Ci/sec Qt - Particulate activity release race of nuclide 1. Ci/sec Derivation of Rt values is presented in Appendix B. These values are listed in Thble B-2.
I i
Anne ndme n t 11 32 (December 1994)
3.2 ODCM CONTROL 3.2.2.1 This section, together with Chapter 4, will be used to demonstrate compliance with ODCM Control 3.2.2.1.
Allowable release rates for batch and continuous releases will be computed Ouch that the dose rate limits of ODCM Control 3.2.2.1 are not exceeded.
The allowable release rate is the lowest of the three values computed as follows (based on Equations 3-3, 3-4, and 3-5):
Noble Cases 1
Qw 1 2.0R (3-6) y 1
On s 0. 3 3 (E, + 1.1 Ry) (3-7)
Tritium, and particulates (T1/2 > 8 days) 1
. Qy s (3-8) 0.67E t where E - total noble gas release rate, Ci/sec OTV " i O W, Qy -
E - total cricium, and particulate (T1/2 > 8 days) i Oy, Qw,'= noble gas release rate for nuclide i, ci/sec Qv, = particulate release rate for nuclide 1. Ci/see R,= (1/03) f Oy, K i Ev =
(1/Qw) Qw, L t 3-3 Amendment 11 (December 1994)
,- )
I
(
Ny * (1/Qw) t$ Ow, Ni 1.1 - constant, the average ratio of tissue to air energy absorption coefficients with the units of rem /" air" rad Ri= (1/Qy) iEOy, Ri l
l Kt- gamma total body dose factor for nuclide i, rem /yr per Ci/see Lt- beta skin dose factor for nuclide i, rem /yr per Ci/see Nt- gamma air dose factor for nuclide i, rad /yr per Ci/sec (NOTE: these are " air" rads not " tissue" rads.)
R.- particulate dose factor for nuclide i, rem /yr per Ci/sec (,
t Since Kr-85 is the only remaining noble gas, equations (3-6) and (3-7) can be solved. The noble gas release rate to determine compliance with ODCM Control 3.2.2.1 is:
Q7 y 5 0.176 Ci/sec Nuclides which require analysis of monthly or quarterly composite samples (e.g., H 3, Sr 89, Sr-90) are not considered in the calculation required by ODCM Surveillance Requirement 4.2.2.1.2 at the time of the release. When the results from these analyses are available, they will be used to confirm that,those nuclides, averaged over the sample period, did not cause violation of ODCM Surveillance Requirement 4.2.2.1.2.
l l
Arre ndmen t 11 3-4 (December 1994) l
3.3 ODCM CONTROL 3.2.2.2 This section will be used to demonstrate compliance with ODCM Control 3.2.2.2 at least once per 31 days.
Noble gas release races shall be evaluated at least once per 31 days using the more limiting of Equations 3-9 and 3-10, together with the dose factors in Table A-1:
50 07 y Ny <1 (3-9) or 25 Orv Nv < 1 (3-10) where Qy-7 2 07y, - total noble gas release race, Ci/sec 07v, - noble gas release rate for nuclide i, Ci/sec f %
1 Wy = E One, N 1 (Chy, f i Ev " E Orv M i i
( Crv s Nt- gamma air dose factor for nuclide i, rad /yr per Ci/see Mi- beca air dose factor for nuclide i, rad /yr per Ci/sec l
'Since Kr-85 is the-only remaining noble gas, equations (3-9) and (3-10) can be solved. The noble gas release rate to demonstrate compliance with OCDM Control 3.2.2.2 is:
Q;v < 1.54E-3 C1/sec 3-5 Amendment 11 (December 1994)
. i 1
3.4 ODCM CONTROL 3.2.2.3 [
' This section will be used to demonstrate compliance with ODCM Control 3.2.2.3 at least once per 31' days.
3.4.1 METHOD 1 l
Method I utilizes the actual particulate (T1/2 > 8 days), and tritium l releases to determine compliance with ODCM Control 3.2.2.3 as follows:
10 0 Qy Eg < 1 (3 11) l where i
Rt - 1/Q y I Qy Rg f i Rg - dose factor for nuclide i, rem /yr per Ci/see from Table B-2 Qy -IQ - total cricium, and particulate release race, Ci/sec y
Qy - cumulative quarterly release rate of each, (l 1
particulate, and tricium nuclide 1 Ci/see 1
Nuclides which require analysis of monthly or quarterly composite samples (e.g., H-3. Sr 89, Sr-90) are not considered in the calculation required by ODCM Surveillan'ce' Requirement 4.2.2.3.1 every 31 days. When the results of these analyses are available, they will be used to confirm that those nuclides., averaged over the sample period, did not cause violation of ODCM Surveillance Requirement 4.2.2.3.1.
1 34 2 METH"OD 2 (Octional)
Should the dose limits of ODCM Control 3.2.2.3 be exceeded using Method 1, a more accurate dose calculation may be made using the methodology specified in Section 3.7 to demonstrate compliance.
l Amendment 11 3-6 l iDecember 1994) l
3.5 ODCM CONTROL 3.2.2.4 This section will be used to demonstrate compliance with ODCM Control 3.2.2.4 at least once per 31 days.
Flow diagrams of the ventilation exhaust treatment system, as applicable to ODCM Control 3.2.2.4, are shown in Figure 3-1.
3.5.1 NOBLE CASES The noble gas release rate limits for ODCM Control 3.2.2.4 will be datermined using the equations listed below. The allowable release rate is
! the lower of the two values calculated by Equations 3-12 and 3-13:
On < 100 Hy (3-12) i l
Q.rv < (3*13) 50 Hy i
I where all parameters have been previously defined.
Since Kr 85 is the only remaining noble gas, equations (3-12) and (3-13) can be solved. The release rate to demonstrate coinpliance with ODCM l Control 3.2.2.4 is: Q7 y < 7.69E 4 Ci/sec l
1 l
3 52 PARTICULATES and TRITIUM The particulace and tritium release race limits for ODCM Control 3.2.2.4 will be determined using the equation listed below. The allowable release-rate is calculated by Equation 3 14:
1 Qy < (3-14) 200 K i where all parameters have been previously defined, i
37 Amendment 11 l (December 1994)
3.6 ODCM CONTROL 3.2.2.5 - TOTAL DOSE [
This section describes the methods to be used to determine compliance with ODCM Control 3.2.2.5, which requires that the annual (calendar year) dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be less than or equal to 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.
3 6.1 SURVEILLANCE REOUIREMENTS ODCM Surveillance Requirement 4.2.2.5.1 requires that cumulative dose:
contributions from liquid effluents and from gaseous effluents shall be determined in accordance with ODCM Surveillance Requirements 4.2.1.2, 4.2.2.2. and 4.2.2.3, and in accordance with the methodology and parameters in the ODCM.
These calculations are to be performed in order to determine whether entry I
into the ACTION statement of ODCM Control 3.2.2.5(a) is required. The I ACTION statement is entered when the calculated doses from the releases of radioactive materials in liquid or gaseous effluents exceed twice the limits of ODCM Controls 3.2.1.2, 3.2.2.2, or 3.2.2.3.
_36 2 METHODOLOGY Dose calculations for the three effluent categories of ODCM controls 3.2.1.2, 3.2.2.2, and 3.2.2.3 are to be performed in accordance with the methodology of Sections 2.3, 3.3, and 3.4, respectively. If any one of these dose limits is exceeded by a factor of two or more, then a specific determination of the actual dose to the l
Amendment 10 3-8 (December 1993)
likely most exposed real member of the public shall be performed. This evaluation shall include a determination-of the total dose from all offluent pathways plus direct radiation contributions from the reactor unit, onsite storage tanks, radwaste, etc.
Should the above total dose determination be required, realistic estimates of-the specific receptor location and exposure pathways shall be developed
'in accordance with appropriate NRC guidance.
I i
39 Amendment 8 (July 1992)
i 3.7 ODCM REPORTING REOUIREMENT 5.1.2 This section describes the method that will be used to calculate doses from gaseous effluents, as required by ODCM Reporting Requirement 5.1.2 (annual Radioactive Effluent Release Report).
3.7.1 GENERAL METHODOLOGY The models of Regulatory Guide 1.109 (Rev. 1, 1977) will be utilized, incorporating site-specific modeling parameters, to compute doses from gaseous effluents for this ODCM Control.
3.7.2 PLANT / SITE-SPECIFIC ASSUMPTIONS l
Meteorological dispersion and deposition factors will be based on historical meteorological data from the Trojan meteorological monitoring system. Separate meteorological factors have been derived for batch and continuous releases, The meteorological model described in Appendix C will be used.
The methodology described in Appendix D will be used to assess the radiation doses from radioactive effluents to individuals due to their activities in UNRESTRICTED AREAS during the reporting period. The results will be reported in the annual Radioactive Effluent Release Report.
i l
Amendment 14 3-10 (April 1997)
I j
i 4
1
\
Plant Vent l fR I
Fuel and Auxiliary Building :
Ventilation Exhaust Vent Collection Header :
l (a)
Containment Purge '
R Effluent Radiation Monitor i
l f
)
Figure 3-1 Ventilation Exhaust Treatment System 1
Amendment 11 (December 1994)
4.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 4.1 LIOUID EFFLUENT MONITORS This section will be used to ensure compliance with ODCM control 3.2.1.1, l Alarm / trip setpoints (High Alarm) will be used for effluent radiation monitors as described below.
t 4.1.1 DELRT ED 4.1.2 DELETED l
l l
l t
i 4-1 Amendment 14 I (April 1997) l 1
T- .
d
(
4.1.3 LIOUID RADWASTE DISCHARGE MONITOR (PRM-9)
The setpoint for liquid radwaste discharge is dependent on the actual' values for radwaste discharge flow rate, Plant dilution flow rate, and isotopic composition of the effluents; thus, a variable radiation monitor setpoint must be utilized.
l Prior to discharge, an isotopic analysis of the batch release will be made ;
L l for principal gamma emitters to determine the required dilution ratio (DR) as follows: i I A t DR = E (4-3) l 1 ECV 1 where 3 I
Ai = concentration of nuclide i in batch release, pCi/ml ECVi = 10 CFR 20 effluent concentration value for nuclide i, pCi/ml.
The maximum tank discharge flow rate (F4) is then determined as follows:
i 0.7 x V "
l Fd = (4-4) l DR l
where Fd = maximum tank discharge flow rate, gpm V m = minimum expected Plant dilution flow rate, gpm 0.7 = conservatism factor Amendment 14 4-2 (April 1997) l
l 1
' The radiation monitor setpoint will then be calculated as follows:
E 1
A i xe i High Setpoint = (4-5) 0.7 where High Setpoint = radiab:.on monitor High Alarm setpoint, net epm A1 = concentration of nuclide i in batch release,
- pCi/ml 1
i-ei - radiation monitor efficiency for nuclide i, cpm per pCi/ml.
l For batch releases of very low activity, the calculated radiation monitor catpoint may be statistically insignificant when compared to PRM back-ground radiation fluctuations. In these cases', the radiation monitor l setpoint may be calculated as follows:
l I
High Setpoint = 3.0 x'A PRM (4-6) where A PRM = PRM background reading fluctuation.
l In addition, during low flow conditions, PRM-9 readings increase during the discharge due to-increasing background radiation. If this effect causes the PRM to alarm, the High Setpoint can be reset.
I I
I I
4-3 Amendment 14 j J
(April 1997)
f 4.2 GRSEOUS EFFLUENT MONITORS This section will be used to ensure compliance with ODCM Control 3.2.2.1.
4.2.1 SETPOINT CALCULATIONS FOR NOBLE GAS EFFLLTNT CHANNELS (PRM 20)
Maximum setpoint values for noble gas effluent channels will be based on Kr-85. The limiting release rate for Kr-85 is the minimum of the release f rates calculated by Equations 4-7 and 4-8.
1 l
t Q1 s (4-7) j 2K i 1
Q* s 0.33(L + 1.1 N g) (4-8) .
! j i
f where Qi = limiting release rate for nuclide i, in Ci/see i
l Ki = total body dose factor for nuclide i, in rem /yr 1
l per Ci/sec l
Li = beta skin dose factor for nuclide i, in rem /yr per Ci/see Ni = gamms air dose factor for nuclide, in rad /yr per Ci/sec (NOTE: these are " air" rads, not " tissue" rads) .
1.1 = constant, the average ratio of tissue to air energy absorption coefficients with the units of rem /" air" rad.
For Kr-85, Qi s 0.176 Ci/sec.
I I
Amendment 14 4-4 !
(April 1997) l
Radiation monitor setpoints are calculated using Equation 4-9.
Q 1 x e x 60 x 10 8 Setpoint 5 (4-9)
F x 2.83E4 where Setpoint . High alarm setpoint value, net epm 60 = constant, sec/ min 108 = constant, pCi/Ci 2.83E4 = constant, cc/ft3 F = Fuel and Aux Bldg Vent Exhaust effluent flow rate, cfm
= 105,000 cfm Qg = release rate of Kr-85, Ci/sec
= 0.176 Ci/sec e = detector efficiency for Kr-85, cpm per pCi/cc
= 6.8E+7 cpm per pCi/cc .
The setpoint for PRM 2C is 2.4E+5 cpm.
4.2.2 SETPOINT CALCITLATIONS FOR PARTICULATE CMANNELS fPRM 1A. 2A)
Limiting release rates for particulates are calculated using Equation 4-11.
1-Q,,R,,
OP A (4 - 11) 0.67 R m 4-5 Amendment 14 (April 1997)
where op - limiting release rate for particulates (excluding Sr-90) , Ci/sec Qsg = maximum quarterly release rate of Sr-90, ci/sec from Table 4-1 Rsa = Particulate dose factor for Sr-90, rem /yr per Ci/see rum = largest particulate dose factor (excluding Sr-90) rem /yr per Ci/sec.
The limiting release rate from Equation 4-11 will then be applie'd to Equation 4-9 with the appropriate detector efficiencies and effluent flow rates to obtain the particulate channel setpoints.
Normally, the setpoints would be based on the nuclide with the most restrictive dose factor (Sr-90). However, the dose factor for Sr-90 is so much higher than the dose factors for other particulates in relation to actual releases that to assume that all particulates are Sr-90 for setpoint calculations would be unnecessarily restrictive. The contribution of Sr-90 was weighted by_the maximum quarterly release yet observed. Basing the setpoints on the next most restrictive nuclide, Cs-134, Equation 4-11 can be solved and the limiting release rate applied to Equation 4-9 to obtain the setpoints for PRM 1A and PRM 2A.
For Cs-134, Op s 2.1 E-7 Ci/sec Amendment'14 4-6 (April 1997)
L________________________ .
1 l
Radiation monitos setpoints for PRM 1A an.d PRM 2A can be calculated by
-solving Equation 4-9, where l
F = effluent flow rate, cfm
= for PRM 1A, .50,000cfm (Cor. cal:aaent purge)
= For TRM 2A,.105,000 cfm (Puel and Aux. Bldg. vent exhaust) l l
Qi = release rate of Cs-134, Cf/sec
= 2.1E-7 Ci/sec E = drtector a efficiency for Cs-134, cpm per sci /cc 4 1.10E+12 cpm per yCi/cc
- cnd all other terms have been previously dt. fined.
1 i
The setpoint for PRM 1A is 9.8E+3 cpm.
The setpoint for PRM 2A is 4.7E+3 cpm. ]
4.2.3 CONDENSATE DEMINERALI2ER BUILDING EFPLUENT FONITORING I
To ensure compliance with ODCM Control 3.2.2.1, samples will be analyzed 1
monthly. Due to the limited effluent volume discharged f: om the building, j it is not necessary to set release limits for this pathway I'
4-7 Amendment 11 (December 1994)
T_ABLE 4-1 HISTORICAL ? ARTICULATE RELEASES Curies Released ~
i Xg,gr Quarter II .12 Total 1977 1 5.2E-4 1.3E-2 2 8.2E-5 1.3E-2 3 2.9E-5 1.2E-3 4 2.8E-5 3.7E-4 1978 1 2.7E-4 3.6E 3 2 7.9E-5 2.0E 3 3 7.8E-5 5.4E-4 4 5.2E-5 3.8E-4 1979 1 1.1E-5 5.3E-3 2 3.8E-5 4.4E-3 3 5.7E-5 8.4E-4 4 1.2E-4 9.3E-3 l
1980 1 5.1E-6 1.4E-3 2 5.4E-5 1.1E-2 3 5.8E-5 6.9E-4 4 4.6E-6 8.1E-4 ;
1981 1 9.0E-6 1.5E-2 l 2 1.4E-5 2.1E-2 l 3 5.2E 6 9.7E-4 l 4 1.1E-5 2.0E 3 j 1982 1 2.6E-5 1.8E 3 l 2 4.5E 4 4.3E 3 3 3.6E-5 8.7E-4 4 3.7E 4 2.1E-3 ,
1983 1 1.1E-4 2.4E-3 j 2 3.6E-4 9.4E-4 l 3 5.5E-5 4.6E-4 i l
4 5.5E-5 5.0E 4 1984 1 8.2E-5 2.1E-3
. 2 8.7E 5 2.2E 3 f
3 7.4E-5 5.2E-4 4 5.1E-5 6.1E 4 .
1985 1 6.6E 5 2.0E-4 2 1.0E-8 2.5E 5 3 5.6E-5 2.0E 4 4 4.9E.7 1.9E-6 Amendment 11
{
(December 1994) r
4 l
1 l TABLE 4-2 [
EFFLUENT PATHWAY FLOW RATES USED FOR METHOD 1 SETPOINT CALCUIATIONS Flow-Rate Effluent Monitor (cfm3
- PPM 1: Containment Purge Mode 50,000 PRM 1: Containment Pressure 140 Relief Mode PRM 2: Fuel and Auxiliary Bldg 105,000 Vent Exhaust li l
Amendment 11 (December 1994) l I
+.
TABLE 4.'3 PARTICULATE CHANNEL DETECTOR EFFF;IENCIES (PRMs 1A and 2A) 1 Detector Efficiency Nuclide (com/uci/ce)
Co 60 8.20E+11 Sr-89 2.40E+12 Sr-90 1.40E+12 Cs-134 1.10E+12 Cs 137 1.50E+12 i
l' I
l Amendment 11 (December 1994) f
t l
l 5.0 ENVIRONMENTAL MONITORING l-
! In accordance with ODCM Control 3.3.1, the radiological environmental i
monitoring stations are listed in Table 5-1 with the radial distance presented
- in macers. The location of these stations with respect to the Trojan Nuclear Plant is shown in Figure 5-1.
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' ANT I I ff -
Amendment 13 -
(December 1995)
,0 CATIONS I I ' '
E .
6.0 TROJAN PROCESS CONTROL PROGRAM i FOR SOLID RADIOACTIVE WASTE I
This chapter will be used to ensure compliance with ODCM Control 3.2.3.1 and the waste form requirements of 10 CFR 61.56, 6.1 PURPOSE To vorify that processed radioactive wastes to be shipped offsite for burial i noot the shipping and burial ground requirements for solidification and dzwstoring.
6.2 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REOUIRING SOLIDIFICATION 6.2.1 SCOPE l This section pertains to radioactive vaste containing a total specific cecivity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A waste as defined in l 10 CFR 61. These wastes must be stabilized by solidification and contain no freestanding liquids prior to shipment offsite for burial, or else be packaged in a high integrity container in accordance with Section 6.3.
6.2.2 PROCRAM ELEMENTS For the disposal of radioactive waste requiring solidification, PCE shall impicment the following steps:
(1) An NRC-approved contract vendor solidification service shall be utilized. The contract vendor solidification service may consist of solidification by the contractor or supply of l
materials, procedures, and process control program for PGE solidification.
61 Amendment 10 (December 1993)
9 i
(2) This vendor service shall include transmittal to PCE of copies of their solidification procedure and process control program prior to performing the solidification.
(3) The process parameters included in the process control program may include, but are not limited to, waste type, vaste pH, waste / liquid / solidification agent / catalyst ratios, vaste oil
! concent, waste principal chemical constituents, and mixing and curing times.
(4) The vendor solidification procedure and process control program shall be incorporated into a temporary Plant Operating Manual procedure that will be effective during the solidification process. .This procedure will identify all Plant interfaces with the vendor's equipment (e.g., flush water, fire protection shielding requirements, etc.), as well as identify the actions to be taken if excess free liquids are observed. This procedure shall require at least one representative test specimen from at la least every tench batch of waste processed to ensure solidification. The procedure should also include the actions to be taken if the test specimen fails to solidify.
l l
(5) This temporary procedure shall be reviewed per plant procedures for adequacy in meeting applicable State, Federal, Department of Transportation, and burial ground regulatory requirements and approved by the Plant Ceneral Manager or desi 5 nee prior to its implementation. This review shall ensure that the stability requirements of 10 CFR 61.56(b) for wastes exceeding Class A I concentrations are met by the vendor solidification program.
I Amendment 11 62 I (December 1994) )
i l
6.3 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH-INTEGRITY CONTAINERS
'6.3.1 SCOPE This section pertains to radioactive waste containing a specific activity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A waste as defined in 10 CFR 61. These wnstes must be stabilized by packaging in dewatered form in a high-integrity container which meets burial ground and regulatory requirements, or else be colidified in accordance with Section 6.2.
6.3.2 PROGRAM ELEMENTS For disposal of radioactive waste requiring a high-integrity container, PGE ehnll implement the following steps:
(1) A contract vendor high-integrity container shall be used.
1 (2) The container shall be demonstrated to have been authorized by the NRC Division of Waste Management or the State of Washington prior to acceptance for use by PGE. This shall include provision by the vendor to PGE of documentation reflecting this authorization.
(3) The material placed in the high-integrity container shall meet all applicable burial ground and regulatory waste form requirements for waste which is packaged in this manner.
(4) The above criteria shall be met by following Plant procedures which will be reviewed'and approved by the Plant General Manager or designee in accordance with Plant administrative procedures prior to implementation at the time of packaging and disposal.
6-3 Amendment 11 (December 1994)
I i
)
l I
6.4 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHER WET WASTES 6.4.1 SCOPE This section pertains to bead-type spent radioactive demineralizer resin and other wet wastes, such as absorbed oils, which contain a total specific activity less than the burial ground criteria for solidification, and which does not exceed the concentration limits for Class A waste as defined in 10 CFR 61.
6.4.2 PROGRAM RT.RMENTS (1) The dewatered resin or wet wastes must meet the requirements of 10 CFR 61.56 or those of the burial ground (whichever is more restrictive) for freestanding, noncorrosive liquid.
(2) For bead resins, the preceding criterion will be met by following approved Plant Operating Manual procedures for dewatering resin.
(3) Liquid waste other than oil must be solidified or packaged in sufficient absorbent material to absorb twice the volume of liquid. Oil must be solidified.
6.5 SUPPORTING DOCUMENTS The_following procedures are used in support of the process control program.
Vendor procedures listed are retained and maintained by the Radiation Protection Specialist:
(1) Local and Vendor Procedures:
RP 303 (Radioactive Waste Shipment Procedure)
RP 320 (RPMP 2) (Packaging of Radioactive Waste)
RP 324 (RPMP 2-8) (Radwaste Drumming - Absorbed Oil)
RP 310 (RPMP 4) (Waste Classification)
Amendment 13 6-4 (December 1995)
l .
l l RP 312 (RPMP 5) (10 CFR 61 Sampling Program)
OI 11-7 Series (Resin Transfers) i NuPac Manual OM-42 . (Resin Drying System Manual / Topical Report)
RP-300 (Radioactive Material Shipment)
RP-321 (Drumming and Compacting)
RP-323 (Packaging in Metal Boxes)
RP-325 (Packaging in C-Vans)
RP-326 (Packaging in EL-50 HIC)
RP-327 (Packaging Resin in EL-50 HIC) l RPMP T-3 (Liner Dewatering Test) l (2) Vendor Procedures Incorporated into Trojan Procedures:
Vectra H-24 (Enviralloy HIC Handling)
Vectra H-18 (Poly HIC Handling)
Vectra OM-108 NS (Resin Drying System)
Vectra OM-35 (HIC Remote Closure)
Vectra OM-32 (EA-50 Wedge Closure)
Vectra H-24 (Enviralloy HIC Handling) 1 1
(3) Other:
10 CFR 61 Sample Resultsl13 l
! SHIPCALC Waste Classification Computer ProgramIII l
SCAL-FAC 10 CFR 61 Scaling Factor Computer Programill l
(1) These items are listed since they are procedures / tools used in
( classifying waste for burial. They have no effect on the physical !
processing of waste for disposal. l 1
6-5 Amendment 14 i (April 1997)
{
6.6 PROGRAM CHANGES Changes to the PCP shall be documented and records of re: views performed shall be retained. This documentation shall contain sufficient information to support the change (s) and appropriate analyses or evaluations justifying the change (s) ; and a determination that the change (s) maintain the overall conformance of the solidified waste product to the existing requirements of Federal, State, or other applicable regulations.
Changes to the PCP shall be effective after review and approval by an Independent Satety Reviewer and the approval of the plant General Manager or designee.
f Amendment 14 6-6 l (April 1997) t
APPENDIX A i
DERIVATION OF NOBLE CAS DOSE FACTORS (K, L, M. N) ;
l Tha noble gas dose factors were derived using the maximum annual average site boundary x/Q for batch releases as follows:
Ki - DFBi x x/Q x 10 12 x 10~3 (A-1)
L, - DFSi x g/Q x 1012 x 10~3 (A-2)
Mi 12 x go-s (3,3)
= DFfi x x/Q x 10 (A-4)
Ni = DF{i x X/O x 10 12 x 10-3 where K, - gamma total body dose factor for nuclide i, rem /yr per Ci/see Li - beca skin dose factor for nuclide i, rem /yr per Ci/sec M, - beta air dose factor for nuclide i, rad /yr per Ci/sec N, - gamma air dose factor for nuclide i, rad /yr per Ci/sec DFB, = Regulatory Guide 1.109 (Rev.1,10/77) total body dose factor, mrem /yr per pCi/m 8 DFS, - Regulatory Guide 1.109 (Rev. 1, 10/77.) skin dose factor, l mrem /yr per pCi/m 3 A-1 Amendment 3 (December 1985) 1
DFf - Regulatory Guide 1.109 (Rev. 1, 10/77) beca air dose factor, mrad /yr per pCi/m 3 DF} - Regulatory Guide 1.109 (Rev. 1, 10/77) gamma air dose factor, mrad /yr per pCi/m X/Q - 1. 3 x 10-s ,,cfas (historical average maximum continuous site boundary X/Q. see Appendix C) 1012 - constant, pci/ci 10 3 - constant, rem / mrem or rad / mrad.
The values of K , L , M , N , and DFB , DFS , DFf , DF{ are given in 3 3 3 i 3 i Table A-1.
Amendment 11 A2 (December 1994)
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APPENDIX B DERIVATION OF PARTICULATE DOSE FACTORS DOSE FACTOR R, Ths term R, is based on the combination of: (a) inhalation, ground plane, vegetable ingestion, meat ingestion and milk ingestion pathways which are present at the location of maximum potential dose (ie, the controlling oxposure location), (b) annual average continuous release meteorology at the controlling exposure location. (c) the most restrictive age group (child), and (d) the critical organ for each nuclide.
Determination of the Site Boundarv as the Contro11ine Exnosure Location The controlling exposure location is that offsite location where the combination of existin5 Pathways and annual average meteorology would indicate tha maximum potential dose. That is, the controlling exposure individual is assumed to breath the air at the nearest residence with the highesc x/Q value, to reside at the nearest residence with the highest D/Q value, and to obtain all the individual's vegetables, mest, and milk from the production locations with the highest D/Q values. To be conservative, it is assumed that the controlling exposure location for all pathways is the site boundary.
The meteorology at the site boundary is discussed in Appendix C and is listed in Tables C.1 and C 2.
i The following general equation is used to calculate Rg values:
R ,
- 10 '3 [(RfxI/Q,) *(RfxD/Q,) +(R[xD/C) *(RfxD/0,) + (Rf x D/Q,)] (B 1) .
j where R, - total dose factor for nuclide i, rem /yr per Ci/see B.1 Amendment 11 (December 1994)
4 R{- inhalation pathway dose factor for nuclide i, mrem /yr per i Ci/m 3 Ri - ground plane pathway dose factor for nuclide i, mrem /yr per Ci/m2 -see RY- vegetable ingestion pathway dose factor for nuclide i, mrem /yr per Ci/m 2 .,,e Rf - meat ingestion pathway dose factor for nuclide i, mrem /yr perCi/m2-see cow or goat milk ingestion pathway dose factor for nuclide i, Rf -
mrem /yr per Ci/m2 .3,e x/Q, - atmospheric dispersion factor for continuous releases at the site boundary, sec/m 3' D/Q, - atmospheric deposition factor for continuous releases at the h i site boundary, m-2 10'3 - constant, rem / mrem.
The dose factors, R[. R?, RY R7, Ri were derived as follows and are listed in Table B-1.
Inhalation Pathway Dose Factor Rf
- 1012(BR) (DFA3 ) (B-2)
Rf where 1012 - constant, pCi/Ci Amendment 11 B-2 (December 1994)
(BR) - breathing rate of the receptor of child age group - 3700 m 3 /yr (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5)
( DFA ) 3
- maximum organ inhalation dose factor for the receptor for
'nuclide i, in mrem /pCi (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-9). The total body is considered as an organ in the selection of the DFAg.
G Ground Plane Pathway Factor Rf RG- (1012)(8760)(SF)(DFG )[(1-e**it)/g 1 3 il (B-3) where 10:2 - constant, pCi/Ci 8760 - constant, hr/yr 13 - decay constant for nuclide i, sec -1 e - exposure cima, 4.73 x los see (15 yr)
( DFC, ) - ground plane total body dose conversion factor for nuclide i, mrem /hr per pCi/m 2(Regulatory Guide 1.109, Rev. 1, 10/77, Table E-6)
SF - shielding factor for residential structures, 0.7.
Vegetation Pathway Factor Rf Man is considered to consume two types of vegetation, fresh leafy vegetables and produce. The vegetation dose factor combines these two pathways using the following equation:
B-3 Amendment 11 (December 1994) r
12 e (DFL1 ) '***L + Uf f,e '*** (B-4)
Rf = 10 Ukf te where Uh
- consumption rate of fresh leafy vegetation by the child receptor 26 kg/yr (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5)
Uf - consumption rate of produce and stored vegetation by the child receptor, 520 kg/yr (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5) ft- fraction of the annual intake of fresh leafy vegetation grown locally, 1.0 (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15) f, - fraction of the annual intake of produce and stored vegetation grown locally, 0.76 (Re5ulatory Guide 1.109, (i Rev. 1, 10/77. Table E-15) et - average time between harvest of leafy vegetation and its consumption, 8.6 x 10' see (1 day) (Regulatory Guide 1.109, Rev. 1. 10/77, Table E 15) en- average time between harvest of stored vegetation and its consumption, 5.2 x 10' see (60 day) (Regulatory Guide 1.109, Rev. 1. 10/77. Table E-15)
Y, - vegetation area density, 2.0 kg/m 2(Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15)
Amendment 11 B-4 (December 1994)
r- fraction of deposited activity retained on vegetation. 0.2 for particulates (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15)
(- decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 x 10-7 sec ~1 (corresponding to a 14-day half life) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15)
(DFL ) -
i maximum organ ingestion dose factor for nuclide i, in mrem /pci (Regulatory Guide 1.109. Rev. 1, 10/77, Table E-13) 1012 - constant, pCi/Ci and all other terms have been defined previcasly.
The concentration of critium in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Rf for tritium is based on X/Q:
Rg _3 = (109 (109 Uat+ a s, f I*i) [0.75(0.5/H)] (B-5) where 103 - constant, gm/kg H- absolute humidity of the atmosphere, 8 gm/m3 (Troj an Appendix I Evaluatinn, 5/76) s.5 Amendment 11
0.75 - fraction of total plant Lass that is water (
0.5 - ratio of the specific activity of the plant mass water to the atmospheric water 1012 - constant, pCi/Ci and all other terms have been defined previously.
l Crass-Cow Meat Pathway Factor Rf RD = 10 ' A0'(U'P) f'f' + (1-f'f') e -*'** (B-6)
+A w (F r) (r) (f.) (DFL t) 1 e "*'* 8 1
2 Yp Y, ,
where Fr - stable element transfer coefficient for meat, in days /kg (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-1) lI U,- child recepter's meat consumption race, 41 kg/yr (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5) er- transport time from pasture to receptor,1.73 x 10' see (20 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15) en- transport time from crop field to receptor, 7.78 x 10' see (90 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E 15)
Y, - agricultural productivity by unit area (stored food), 2.0 kg/m2 (Regular.ory Guide 1.109, Rev. 1, 10/77, Table E-15) f, - fraction of year that cow is on pasture, 0.5 (Regulatory Guide 1.109, Rev. O, Page 1.109-26) l Ame ridme n t 11 B6 ibecember 1994)
f, - fraction of cow feed that is pasture grass while cow is on pasture, 1.0 Qr - cows' consumption rate of feed, 50 kg/ day (Regulatory Guide 1.109, Rev. 1, 10/77. Table E-3) e Y, - agricultural productivity by unit area (pasture), 0.7 kg/m2 (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15).
1012 - constant, pCi/Ci and all other terms have been defined previously.
The concentration of critiwn in meat is based on its airborne concentration rather than the deposition. Therefore, the Rf for tritium is based on x/Q:
(B-7)
R _3 = (109 (10 ) F gOrU ,p (DFL3 ) [0.7 5 (0. 5/H) ]
3 where all terms have been defined previously.
Crass Coat-Milk Pathway Factor Rf R9
= 10" A + A, (F,) (r) (f.) (DFL ) Y, 1
+ e *
~
(B-8) 1 1
Y, ,
where Qr - goat's consumption rate of feed, 6 kg/ day (Regulatory Guide 1.109, Rev. 1, 10/77. Table E-3)
U., - child receptor's milk consuinpcion rate , 330 1/yr (Regulatory Culde 1.109, Rev. 1, 10/77, Table E-5)
B7 Amendment 11 (December 1994)
S Y, - agricultural productivity by unit area of stored feed. 2.0 kg/m 2 i (Regulatory Guide 1.109, Rev. 1, 10/77, Table E 15)
F, - stable element transfer coefficient for milk, in days /1 (Regulatory Guide 1.109, Rev. 1, 10/77, Tables E-1 and E-2) r- fraction of deposited activity retained on goat's feed grass.
0.2 for particulates (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15) er- transport time from pasture to goat, to milk, to receptor, 1.73 x 105 see (2 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E 15) en- transport time from pasture, to harvest, to goat, 7.78 x 10' see (90 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15)
I 1012 - constant, pCi/Ci and all other terms have been previously defined.
Crass Cow-Milk Pathway Factor Rf e~ '*'
- '+ (B-9)
R9 = 10 0 F
(F*) (r) (f.) (DFL )
t Y,
1 A i + A, ,,Y where Qr - cow's consumption race of feed, 50 kg/ day (Regulatory cuide 1.109, Rev. 1, 10/77, Table E 3)
U.,- child receptor's milk consumption race, 330 1/yr (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-5)
Amendment 11 B-8 (December 1996)
Y, - agricultural productivity by unit area of stored feed, 2.0 kg/m2 (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15)
F, - stable element transfer coefficient for milk, in days /1 (Regulatory Guide 1.109. Rev. 1, 10/77, Table E-1) r- fraction of deposited activity retained on cow's feed grass.
0.2 for particulates (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15) er- transport time from pasture to cow, to milk, to receptor, 1.73 x 105 see (2 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E-15) ta - transport time from pasture, to harvest, to cow,
- 7. 78 x 10' sec (90 days) (Regulatory Guide 1.109, Rev. 1, 10/77, Table E 15) .
1012 - constant, pCi/Ci and all other terms have been previously defined.
The concentration of tritium in milk is based on the airborne concentration .
rather than the deposition. Therefore, 'the Rf for tritium is based on x/Q:
as.3.u0=>a0=>ro,0.o wry m.Ts m.smo 1 (3 10) where all parameters have been defined previously.
B-9 Amendment 11 (December 1994)
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= f .* 3 3 4 5 A.:: ? d d a 0 8 Amendment 11 (December 1994)
l TABLE B-2 k PARTICULATE DOSE FACTORS Ri Composite Dose Factor At Site Boundary (rem /vr)
(Ci/seel Nuclide H-3 1.07E+02 Mn-54 1.50E+05 Fe-55 7.31E+04 Co-57 3.66E+04 Co-60 1.64E+06 Zn 65 2.98E+05 Sr-89 2.97E+06 Sr-90 9.26E+07 Zr-95 1.25E+05 Ag-110m 5.74E+05
((
Sb-124 2.32E+05 Cs-134 7.02E+06 Cs-137 6.53E+06 Ce 144 8.39E+05 Amendment 11 (December 1996
APPENDIX C METEOROLOGY QUARTERLY AVERAGE METEOROLOGY Meteorology data required for the compilation of the radioactive release reports in ODCM Control 5.1.2 was calculated at the end of each calendar quarter from 1976 until 1993 using the NRC Computer code XOQDOQ and the methodology of Regulatory Guide 1.111 (Rev. 1, 7/77).
The maximum site boundary X/Q and D/Q values for continuous and batch releases for each quarter from 1976 until 1993 are presented in Table C-1 and C.2. Tables C-1 and C-2 include the' calculated averaSe X/Q and D/Q values for the entire time period. These average values are to be used for the compilation of radioactive release reports in ODCM Control 5.1.2.
I C-1 Amendment 11 (December 1994)
TABLE C-1 Sheet 1 of 3 HISTORICAL METEOROLOGY DATA BATCH RELEASES QUARTER YEAR DIRECTION X/Q D/Q PDF 1 1976 N 5.60E-05 4.70E-07 0.85 2 1976 N 4.10E-05 2.00E-07 0.85 3 1976 NNE 3.80E-05 2.70E-07 0.85 4 1976 NNE 6.00E-05 1.30E-07 0.85 1 1977 N 7.10E-05 3.60E 07 0.85 2 1977 N 7.00E-05 2.60E-07 0.85 3 1977 NNV 7.10E-06 5.40E-08 0.92 4 1977 NNE 1.00E-05 7.00E-08 0.91 1 1978 N 1.20E-05 5.00E-08 0.92 2 1978 N 1.20E-05 3.90E-08 0.92 3 1978 N 5.90E-06 2.80E-08 0.92 4 1978 N 1.20E-05 7.30E-08 0.92 1 1979 NNW 1.10E-05 7.50E-08 0.92 2 1979 SE 8.60E-06 3.20E-08 0.90 3 1979 ESE 5.20E-06 2.40E-08 0.91 4 1979 NNW 1.30E-05 6.40E-08 0.92 1 1980 NNV 1.60E-05 8.30E 08 0.92 2 1980 N 1.00E 05 2.80E-08 0.92 3 1980 N 5.20E 06 1.50E-08 0.92 4 1980 N 1.10E-05 7.30E-08 0.92 1 1981 NNW 1.20E 05 7.30E-08 0.92 2 1981 N 9.00E-06 3.20E-08 0.92 3 1981 NNE 2.80E-06 1.40E 08 0.92 I
4 1981 N 9.30E-06 6.70E-08 0.92 p
Ammendment 11 (December 1994)
TABLE C-1 Sheet 2 of 3 HISTORICAL METEOROLOGY DATA BATCH RELEASES QUARTER YEAR DIRECTION X/Q D/Q PDF 1 1982 N 8.20E-06 6.00E 08 0.92 2 1982 NNW 6.00E-06 3.80E-08 0.92 3 1982 NNE 5.50E-06 1.20E-08 0.92 4 1982 NNU 1.30E-05 8.30E-08 0.92 1 1983 N 1.30E-05 7.00E-08 0.92 2 1983 N 6.70E-06 2.50E-08 0.92 3 1983 N 7.60E-06 2.00E-08 0.92 4 1983 N 1.20E-05 7.10E-08 0.92 1 1984 N 1.30E-05 6.60E-08 0.92 2 1984 N 1.00E-05 3.90E-08 0.92 3 1984 ENE 4.40E-06 7.10E-09 0.92 4 1984 N 1.60E-05 9 .10 E- 0 ?, 0.92 1 1985 N 1.10E-05 6.30E-08 0.92 2 1985 N 6.00E-06 1.80E-08 0.92 3 1985 E 8.10E-06 1.10E-08 0.92 4 1985 N 1.30E-05 7.10E-08 0.92 1 1986 N 1.20E-05 6.50E-08 0.92 2 1986 N 1.10E 05 3.60E 08 0.92 3 1986 NE 5.90E-06 9.30E 08 0.91 4 1986 N 1.10E-05 5.00E-08 0.92 1 1987 N 1.50E-05 6.30E 08 0.92 2 1987 N 6.10E-06 2.20E-08 0.92 3 1987 ESE 4.60E-06 1.60E-08 0.91 4 1987 N 1.30E-05 5.20E-08 0.92 Amendment 11 (December 1994)
TABLE C-1 Sheet 3 of 3 HISTORICAL METEOROLOGY DATA [
BATCH RELEASES YEAR DIRECTION X/Q D/Q PDF QUARTER 1 1988 N 1.30E-05 7.30E-08 0.92 2 1988 N 9.10E 06 2.20E-08 0.92 3 1988 ESE 1.20E-05 2.50E-08 0.91 4 1988 N 1.90E-05 7.80E-08 0.92 1 1989 NW l.50E-05 1.10E-07 0.92 2 1989 E 1.00E-05 1.50E-08 0.92 3 1989 ESE 8.50E-06 3.10E-08 0.91 4 1989 NW 9.50E-06 6.50E-08 0.92 1 1990 N 8.60E-06 4.20E-08 0.92 2 1990 N 1.10E-05 3.40E-08 0.92 3 1990 ESE 5.60E-06 2.00E-08 0.91 4 1990 N 1.40E-05 8.90E 08 0.92 1 1991 NW 8.40E-06 4.30E-08 0.92 g 2 1991 N 7.20E 06 3.00E-08 0.92 3 1991 ESEE 4.60E-06 2.10E-08 0.91 4 1991 NW 8.40E-06 4.30E-08 0.91 1 1992 NW l.70E 05 5.90E-08 0.92 2 1992 N 8.80E 06 1.40E-08 0.92 3 1992 ESE 6.60E-06 2.40E-08 0.91 4 1992 N 1.40E 05 9.20E 08 0.92 1 1993 NNW 1.60E-05 7.00E-08 0.92 2 1993 NW 8.90E-06 3.80E 08 0.92 3 1993 ESE 5.30E 06 2.10E-08 0.91 4 1993 W l.20E-06 5.90E 08 0.92 AVERAGE 1.38E 05 6.73E-08 0.92 Aine ndme n: 11 (December 199:-)
l . _ _ _ _ .
(
, 1 TABLE C-2 Sheet 1 of 3 HISTORICAL METEORLOGICAL DATA CONTINUOUS RELEASE -1 QUARTER YEAR DIRECTION X/Q D/Q PDF 1 1976 N 6.10E-05 5.70E-07 0.85 2 1976 N 3.60E-05 2.20E-07 0.85 3 1976 N 1.80E-05 1.00E-07 0.85 4 1976 N 3.70E-05 2.00E-07 0.85 1 1977 N 5.40E-05 3.00E-07 0.85 2 1977 N 3.80E-05 1.50E-07 0.85 3 1977 N 6.20E 06 2.90E-08 0.92 4 1977 N 8.60E-06 6.00E-08 0.92 1 1978 NNW 1.20E-05 6.50E-08 0.92 .
2 1978 N 1.10E-05 3.30E-08 0.92 3 1978 N 6.70E-06 3.10E-08 0.93 4 1978 NNW 1.30E-05 5.10E-08 0.92 1 1979 NtN 1.10E-05 7.00E-08 0.92 2 1979 N 5.90E-06 2.00E-08 0.92 3 1979 ESE 4.80E-06 2.10E-08 0.91 4 1979 N 1.40E-05 8.10E-08 0.92 1 1980 NtN 1.20E-05 6.10E-08 0.92 2 1980 N 8.50E-06 2.60E-08 0.92 3 1980 ESE 4.50E-06 1.90E-08 0.91 4 1980 N 1.40E-05 6.90E-08 0.92 1 1981 N 1.40E-05 5.30E-08 0.92 2 1981 N 9.60E 3.60E-08 0.92 3 1981 N 2.90E-06 1.80E 0P 0.92 4 1981 N 8.80E-06 7.20E-08 0.92 Amendment 11 (December 1994)
TABLE C-2 Sheet 2 of 3
. HISTORICAL METEORLOGICAL DATA CONTINUOUS RELEASE YEAR DIRECTION X/Q D/Q PDF QUARTER 1 1982 N 9.20E-06 7.10E 08 0.92 2 1982 NNW 5.70E-06 3.60E-08 0.92 3 1982 ESE 6.30E-06 2.60E-08 0.91 4 1982 N 1.10E 05 6.00E-08 0.92 i
1 1983 N 1.40E-05 7.90E-08 0.92 ,
2 1983 N 6.50E-06 2.30E-08 0.92 3 1983 N 7.00E-06 2.20E-08 0.92 4 1983 N. 1.30E 05 7.30E-08 0.92 1 1984 NNW 1.20E-05 8.00E-08 0.92 2 1984 N 1.10E 05 4.10E-08 0.92 3 1984 N 6.10E-06 1.50E-08 0.92 4 1984 N 1.50E-05 8.20E-08 0.92 (
1 1985 NNW 1.20E-05 6.40E-08 0.92 2 1985 N 6.00E-06 1.90E-08 0.92 3 1985 ENE 7.20E-06 9.30E-08 0.92 4 1985 N 1.40E-05 7.00E-08 0.92 1 1986 N 1.30E-05 5.50E-08 0.92 2 1986 N 9.00E-06 2.80E-08 0.92 3 1986 ENE 7.30E 06 1.00E-08 0.92 4 1986 N 1.10E-05 5.60E-08 0.92 1 1987 N 1.40E 05 6.20E-08 0.92 2 1987 N 5.80E-06 2.00E-08 0.92 1987 ESE 4.80E 06 1.50E-08 0.91 3
b 4 1987 N 1.40E 05 5.10E-08 0.92 Amendment 11 (Oecember 199t)
TABLE C-2 Sheet 3 of 3 HISTORICAL METEORLOGICAL DATA CONTINUOUS RELEASE QUARTER YEAR DIRECTION X/Q D/Q PDF 1 1988 N 1.30E-05 6.10E-08 0.92 2 1988 N 9.00E-06 2.50E-08 0.92 3 1988 ESE 1.10E-05 2.90E-08 0.91 4 1988 N 1.90E-05 6.50E-08 0.92 1 1989 NNW 1.50E-05 8.80E-08 0.92 2 1989 E 9.10E-06 1.60E-08 0.92 3 1989 ESE 7.80E-06 2.70E-08 0.91 4 1989 N 1.30E-05 6.10E-08 0.92 1 1990 N 1.20E-05 6.00E-08 0.92 2 1990 N 9.40E-06 3.10E-08 0.92 3 1990 ESE 5.10E-06 2.10E-08 0.91 4 1990 N 1.60E-05 8.50E-08 0.92 1 1991 NNW 1.20E-05 7.70E-08 0.92 2 1991 N 6.50E-06 2.60E-08 0.92 s 3 1991 ESE 4.80E-06 2.30E-08 0.91 4 1991 NNW 1.20E-05 7.70E-08 0.92 1 1992 NNU 1.40E-05 7.10E-08 0.92 2 1992 E 6.30E-06 2.00E Od 0.92 3 1992 ESE 4.50E-06 2.00E-08 0.91 4 1992 N 1.30E-05 8.00E-08 0.92 1 1993 NNW 1.50E-05 6.80E 08 0.92 2 1993 NNW 1.10E-05 4.50E-08 0.92 3 1993 ESE 6.70E-06 2.20E-08 0.91 4 1993 NNW 9.50E-06 3.40E-08 0.91 AVERACE 1.25E-05 6.44E-08 0.91 Amendment 11 (December 1994)
APPENDIX D METHODOLOGY FOR DETERMINING DOSES TO PERSONS UTILIZING UNRESTRICTED AREAS VITHIN THE SITE EXCLUSION AREA BOUNDARY Noble gas doses are directly proportional to the atmospheric dispersion factor (x/Q, sec/m 3). The methodology contained in this Appendix assumes quarterly average meteorology and ground-level release sector average x/Q (Meteorology and Acomic Energy Equation 3.144).
The following equation is used to calculate the adjusted X/Q values for specific locations within the site boundary:
X/Q A (sec/m )
3 = 1 (0.01) (O
[OF) Equation (D-1) o,pI(2 n 2$)
n ,
where f - wind frequency, percent
- mean wind speed, meters /see x - downwind distance , meters n - number of cardinal compass sectors - 16 o, - vertical dispersion parameter, Pasquill Class E hours of annual occupancy OF - occupancy factor -
8760 NOTE: Occupancy factors for both recreational and occupational cases should be considered.
D1 Amendment 8 (July 1992)
J l
To obtain doses at these locations, multiply the calculated doses at the (
site boundary by the ratio of the adjusted atmospheric dispersion factor (x/Q) at the location of interest to the X/Q at the site boundary:
Equation (D-2) pA. (x/0)^ [D 88]
,(%/Q)sa .
where D^ - dose at the location of interest (mrem)
(x/Q)^ - adjusted x/Q at the location of interest (from Equation D-1) (sec/m8)
(x/Q)sa - X/Q at the site boundary (sec/m3)
Dsn - dose at site boundary (arem).
The methodology described in this Appendix was used to determine doses to persons utilizing unrestricted areas within the site exclusion area boundary beginning in the third quarter of 1992 and continuing through the fourth quarter of 1993. Table D-1 summarizes the correction factors calculated during this time period. Based upon the historical data, a batch release correction factor of 3.8 will be used and a continuous release correction factor of 4.0 will be used.
Amendment 11 D-2 (December 1994)
TABLE D-1 CORRECTION FACTOR FOR PERSONS UTILIZING UNRESTRICTED AREAS WITHIN THE SITE EXCLUSION AREA BOUNDARY BATCH CONTINUOUS QUARTER YEAR . (X/0) # (X/0)#
(X/0)38 (X/0)38 3 1992 2.5 2.8 4 1992 2.1 1.9 1 1993 1.5 1.5 2 1993 1.7 2.4 3 1993 3.8 4.0 4 1993
- 1.6 MAXIMUM 3.8 4.0 MINIMUM 1.5 1.5 AVERACE 2.3 2.4 -
- No batch releases during the quarter.
Amendment 11 (December 1994)
o l APPENDIX E Deleted i
E-1 Amendment 11 (December 1994)
APPENDIX F BASIS FOR CURIE RELEASE VALUES UTILIZED IN LIQUID EFFLUENT SURVEILIANCE REQUIREMENTS This appendix demonstrates that ODCM Surveillance Requirement 4.2.1.2.1, Radioactive Effluents (Liquid) Dose, based on the total curies released (excluding tritium and dissolved gases), results in offsite doses significantly below the ODCM Control limits of 1.5 mrem total body / calendar quarter and 2.5 mrem to maximum organ / calendar quarter.
Table F-1 presents the maximum calculated dose due to liquid effluents from Trojan for the period 1976-1985 (reference PGE-1015 dated March 1977 and PCE-1015 for 1978 through 1985).
Columns 4 and 5 of Table F-1 show the offsite dose which would have resulted if Trojan had released 2.5 Ci in each of the quarters during 1976-1985. It can be seen that in no case would 2.5 mrem to maximum organ or 1.5 mrem total body have been exceeded. In fact, during the quarter with the highest mrem /Ci released factor, the offsite doses were
<10 percent of the ODCM Control limits of 1.5 mrem total body and 2.5 mrem maximum organ.
F1 Amendment 10 (December 1993)
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& j APPENDIX G l
l QUALITY' ASSURANCE REQUIREMENTS FOR THE ENVIRONMENTAL AND EFFLUENT MONITORING PROGRAM
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Environmental and effluent monitoring is a quality-related activity. The Trojen Quality Assurance Program is applicable to the following areas:
$ Organization.
+ QA program and training.
+ Design control.
+ Procurement.
+ Instructions, procedures and drawings.
l $ Document control.
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! + Purchased material, equipment and services.
- Identification of materials.
+ Testing.
+ M & TE.
l + Handling, storage and shipping.
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+ Inspection, test and operating status.
+ Nonconforming materials.
l l + Corrective actions.
l l + QA records.
I l + Audits.
I l
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1 l
i l
l i C-1 Amendment 11 l
(December 1994) i L