3F0887-20, Semiannual Radioactive Effluent Release Rept,Jan-June 1987
| ML20237H098 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/30/1987 |
| From: | Johnson S, Eric Simpson FLORIDA POWER CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 3F0887-20, 3F887-20, NUDOCS 8709030117 | |
| Download: ML20237H098 (46) | |
Text
{{#Wiki_filter:SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 1/01/87 - 6/30/87 FLORIDA POWER CORPORATION CRYSTAL RIVER - UNIT 3 FACILITY OPERATING LICENSE NO, DPR-72 DOCKET No. 50-302 AUGUST, 1987
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l Approved by: dd O . Manager, Sife $ lear Services Date: % \,11 9> ~} 8709030117 870630 72 DR ADOCK 0500 TE 45 _ _ _ = _ _ _ _ _ _ _ _ - _ . - _ _ _ _
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TABLE OF CONTENTS 2ADI I Introduction 1 II Releases of and Doses from Gaseous Effluents 5 III Releases of and Doses from Liquid Effluents 16 IV Solid Waste Shipments 27 V Meteorological Data 31 VI Technical Specification Reports 32 VII ODCM and PCP Changes 33
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LIST OF TABLES PADI I-1. : Annual Doses to Individuals.from all Releases 2. LII-1 Gaseous Effluents - Summation of All Releases 9 II-2 Gaseous Effluents - Ground Level Releases 10 II-3 Doses to Individuals from Continuous Gaseous 12 Effluent' Releases .) II-4 Doses to Individuals from Batch Gaseous Effluent Releases 13
.II-5 Doses'to.the Population from Continuous Gaseous Effluent Releases 14 II-6 Doses toLthe Population from Batch Gaseous Effluent Releases 15 III-1 Liquid Effluents - Summation of All Releases 21 III-2 Summation of Nuclides in Liquid Effluent Releases 22 III-3 Doses to-Individuals from Liquid Effluent Releases -
First and Second Quarters 25 III-4 Doses to the Population from Liquid Effluent Releases - First and Second Quarters 26
.IV-1 Summation of Solid Waste Irradiated Fuel Shipments 28 IV-2 Effluent and Waste Disposal Report Shipment Summary 29 ii
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. LIST OF I,.GURES PAGE I-1 Whole Body Doses to the Individual From Gaseous Effluent Releases 3 I Whole Pody, Doses to the Individual From Liquid Effluent Releases 4 II-1 Gaseous. Releases-Fission and Activation Gases and Tritium 11 III-1 Methods of Meeting 10 CFR 20, Appendix B, Table II, Column 2, MPC Limits .20 i III-2 Liquid Releases-Fission and Activation' products
'(Exclusing H 3, Gases, Alpha). 24 IV-1 Volume of Solid Waste Shipped off-Site 30 IV-2 Ratio of Content to Volume of Waste Shipped Off-Site 30 i
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I. INTRODUCTION This Effluent and Waste Disposal Report is submitted as required by Techni-cal Specification 6.9.1.5.d to the Crystal River Unit 3 Facility Operating License No. DpR-72. The data in this report covers the period from January 1 to June 30, 1987. There have been no changes to the Technical Specification Requirements for effluents and waste disposal of the Facility Operating License during the period of this report. Crystal River Unit 3 has had no significant measurable radiological impact on the surrounding environment during the reporting period. This is based on the Radiological Environmental Monitoring program data and the doses calculated for individuals and the population due to effluent releases i being significantly below the levels required by 10 CFR 50, Appendix I. The summations of gaseous and liquid effluents and solid waste shipments are in accordance with the tables in Regulatory Guide 1.21 (Rev. 1, 6/74) Appendix B. 4 j The individual and population doses were calculated using GASpAR (for gas-eous ef fluents) and LADTAp (for liquid effluents) computer codes obtained from the Nuclear Regulatory Commission and revised to include site specific data wherever possible. With respect to doses to individual members of the public, a study was j performed to compare the potential dose to members of the public at the , site boundary to that of the worst case individual inside the site i g boundary. The evaluation considered only doses due to gaseous effluents: i-exposure via liquid effluents was not considered since drinking water l pathways are not available, and the seafood ingestion pathway is not site boundary dependent. Meteorological conditions, building wake effects, I receptor distance, and receptor residence times were considered, f
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The study demonstrated that a member of the public at the site boundary ) receives a dose that is several times higher than the worst case member of the public inside the site boundary. Therefore, the dose summaries in this report are conservative estimates for all members of the public. The values reported for the activity of nuclides released are the actual measured activities. The totals of activity released is a total of only the nuclides that had measured activity. l
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.i II. RELEASES OF AND DOSES FROM GASEOUS EFFLUENTS
- 1. Reculatory Limits The Technical Specification limits for gaseous effluent releases are as follows:
l A, Specification 3.1L_L1 The dose rate at or beyond the SITE BOUNDARY, due to radioactive-materials released in gaseous effluents, shall be limited as follows:
- a. Noble gases: less than or equal to 500 mrem / year total body and less than or equal to 3000 mrem / year to the skin.
- b. Iodine-131, Tritium, and radioactive particulate with half-lives of greater than 8 days: less than or equal to 1500 mres/ year to any organ.
B. Specification 3.11.2.2 The air dose at or beyond the SITE BOUNDARY, due to radioactive noble gases released in gaseous effluents shall be limited to.
'a. During any calendar quarter: less than or equal to 5 mrad gamma and less than or equal to 10 mrad beta radiation, and'
- b. During any calendar year: less than or equal to 10 mrad gamma and less than or equal to 20 mrad beta radiation.
C. Specification 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Tritium, radioactive particulate with half-lives greater than 8 days in gaseous effluents released from the site to areas at or beyond the SITE BOUNDARY shall be limited as follows:
- a. During any calenda c mter: less than or equal to 15 mrem to any organ.
D. The maximum activity to be contained in one waste gas storage tank shall not exceed 39,000 Curies (considered as Xe-133). ; 2 Maximum Permissible Concentrations The maximum permissible concentrations of nuclides in gaseous releases is based on the resultant doses at the site boundary as determined , from the concentrations of nuclides at the release point. The OFFSITE l DOSE CALCULATION MANUAL provides the equations and dose factors that relate to the gaseous activity to be released to doses at the site 5
L 1 boundary, and restrictions are placed on quarterly and yearly release rates. The gaseous releases do not exceed the concentration limits specified in 10 CFR 20 and are as low as reasonably achievable in accordance with the requirements of 10 CFR 50. The total dose and dose rate calculations are derived from the methodology in NUREG-0133 and the dose factors in Reg. Guide 1.109.
- 3. Measurements and Approximations of Total Radioactivity The gaseous effluent release via the Auxiliary Building Exhaust is treated as a continuous release subdivided into discrete periods of filter changes and the radioactivity measured as follows:
A. Fission and Activation Gases - The total activity released is based on the total vent flow and the activity of the individual nuclides obtained from an isotopic analysis of a grab sample taken at least weekly. B. Iodines - The activity released as Iodine-131, 133, and 135 is based on the charcoal cartridge activities (RMA-2I), the particulate filters activities (RMA-2p) and the total vent flow. C. particulate - The activity released via particulate with half-lives greater than eight days is determined by isotopic analysis of particu-late filters (RMA-2p) and the total vent flow. The radioactivity released by batch releases of the Waste Gas Decay Tanks via the Auxiliary Building Exhaust is measured as follows: A. Fission and Activation Gases - The activity released is based on the volume released and the activity of the individual nuclides obtained from an isotopic analysis of a grab sample taken prior to the release. B. Iodines - The iodines from batch releases are included in the iodine determination from the continuous Auxiliary Building releases. C. particulate - The particulate from batch releases are included in the particulate determination from the continuous Auxiliary Building release. D. Tritium - The activity released as tritium is based on a grab sample analysis of each batch and the batch volume. The radioactivity released by purge releases of the Reactor Building through the Reactor Building vent is measured as follows: A. Fission and Activation Gases - The total activity released is based on the total vent flow and the activity of the individual nuclides obtained from an isotopic analysis of a grab sample taken prior to the beginning of the Reactor Building purge and at least weekly during continuous ventilation. 6 I
B. Iodines - The total curies released as iodine-131, 133 and 135 were determined from the charcoal cartridge activities (RMA-1I) and the particulate filter activities (RMA-1p). C. particulate - The total curies released via particulate with half-lives greater than eight days are determined by isotopic analysis of each purge particulate filter (RMA-1p). D. Tritium - The total curies released as tritium are based on grab samples taken for each purge (or the average if more than one grab - sample was taken). Estimated errors are based on errors in counting equipment calibration, counting statistics, vent flow rates, vent sample flow rates, nonsteady re-lease rates, chemical yield factors and sample losses for such items as charcoal cartridges. A. Fission and Activation Gas Total Release as calculated from process monitor readings and grab sample isotopics. Monitor Statistical Error 30% Monitor Error in Calibration 50% Vent Flow Rate 10% Non-Steady Release Rate _211 110% B. I-131 Total Release as calculated from charcoal and particulate filter activity. Statistical Error 60% Counting Equipment Calibration 10% Vent Flow Rate 10% Vent Sample Flow Rate 10% Non-Steady Release Rate 10% Losses from Charcoal Cartridge 104 110%
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C. Particulate with half-lives greater.than eight days release as calcu-lated from particulate filter activities. Statistical Error 60% Counting Equipment Calibration 10% Vent Flow Rate 10% Vent Sample Flow Rate 10% Non-Steady Release Rate _191 100t D. Total Tritium release as calculated from periodic grab sample analy-ses. Water Vapor in Sample Stream Determination 20% Vent Flow Rate 10% Counting Calibration and Statistics 10% Non-Steady Release Rate _1Q1
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- 4. Batch and Unolanned Releases The batch gaseous effluent releases may be summarized as follows:
OUARTER 1 QUARTER 2 l Number of Batch Releases 1.70E+01 5.00E+00 Total time for al:t releases (minutes) 1.34E+04 4.14E+03 Maximum time for any one release (ainutes) 1.30E+03 1.10E+03 Average time for all releases (ainutes) 7.89E+02 8.29E+02 Minimum time for any one release (minutes) 1.39E+02 4.97E+02 Number of Unplanned Releases 3.00E+00 1.00E+00 Total Unplanned Activity Released (Curies) 9.98E-01 3.52E-01 The summation of gaseous effluent releases is in Table II-1 and the summation of nuclides in gaseous ~ effluent ground level releases is in Table II-2. A description of unplanned releases follows: QLIE NCORf DESCRIPTION 2-25-87 87-44 Release from Intermediate Building due to operation of Post-Accident Sampling System. 2-26-87 87-44 Release from Intermediate Building due to operation of Post-Accident Sampling System. 3-13-87 87-48 Release from Intermediate Building due to operation of Post-Accident Sampling System. 6-02-87 87-89 Release from Intermediate Building due to operation of post-Accident Sampling System. l l l' l ! 8 r _-_ . - _ _ _ - -
TABLE II-1 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT - 1/1/87 - 6/30/87 GASEOUS EFFLUENTS - SUMMATION OF,ALL RELEASES Unit Quarter Quarter Est. Total 1 2 Error % A. Fission and Activation Gases
- 1. Total Release Ci 2.40E+02 4.08E+02 1.10E+02
- 2. Average Release Rate for Period uCi/sec 3.09E+01 5.19E+01 ;
- 3. Percent of Technical Specification Limit \ 2.20E-01 2.05E-01 B. Iodines
- 1. Total Iodine - 131 Ci 1.41E-04 1.28E-04 1.10E+02
- l. Average Release Eate for Period uCi/sec 1.81E-05 1.63E-05
- 3. Percent of Technical Specification Limit % 1.00E+01 7.00E-01 C. Particulate
- 1. Particulate with. half-lives > 8 days Ci 7.15E-08 1.49E-07 1.00E+02 l.
- 2. Average Release Rate for Period uci/sec 9.20E-09 1.90E-08
- 3. Percent of Technical Specification Limit \ 1.00E+01 7.00E-01
- 4. Gross Alpha Radioactivity Ci <4.23E-06 (4.32E-07 i
D. Tritium
- 1. Total Release C1 1.53E+00 1.22E+00 9.00E+01 l
- 2. Average Release Rate for Period uCi/sec 1.97E-01 1.55E-01
- 3. Percent of Technical Specification Limit 4 1.00E+01 7.00E-01 i
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TABLE II-2 , EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT - 1/1/87 - 6/30/87 GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES CONTINUOUS MODE BATCH MODE Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 ~1
- 1. ~F ission gases krypton-85 Ci , 2.71E+00 3.41E+00 krvoton-85m Ci 1.36E-01 3.93E-02 6.61E-03 krypton-87 C1 1.07E-01 6.34E-03 krypton Ci xenon-133 Ci' 7.78E+01 2.08E+02 1.40E+02 1.77E+02 xenon-133m 1Ci 9.38E-01 1.55E+00 xenon-135 C1 1.58E+01 1.48E+01 4.78E-01 9.82E-01 xenon-135m Ci 1.42E-03 xenon-131m Ci 1.83E+00 3.28E+00 xenon-138 Ci unidentified Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Total for Period Ci 9.38E+01 2.22E+02 1.46E+02 1.86E+02 fodines iodine-131 Ci 1.33E-04 7.86E-05 7.86E-06 4.90E-05 iodine-135 Ci 8.30E-06 lodine-133 Ci 9.77E-06 4.84E-05 1.40E-05 Total for Period C1 1.43E-04 1.27E-04 7.86E-06 7.13E-05
- 3. Particulate co1balt-58 Ci cesium-134 Ci cesium-137 Ci mancanese-54 Ci iron-59 Ci cobalt-60 Ci zine Ci molybdenum-99 Ci cerium-141 Ci cerium-144 Ci strontium-89 Ci 3.12E-08 strontium-90 Ci 4.03E-08 1.49E-07 i te11urium-132 Ci 6.80E-05 unidentified C1 1.77E-05 8.83E-06 3.25E-09 0.00E+00 Total Ci 1.77E-05 8.89E-06 6.80E-05 0.00E+00 10
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- - - - - - - - U mm es - - - - - - -
Et se EEEEEEE E O se EEEEEEE O23 or 4736932 7 23 or 3476201 L a 00 Da 65461 08 5 D00 D a 0632933 BB T - - f N - - A SEE e - 25P 452 3 OEE e f 321 1 1 21 Tm R93 s C99 s o I 35 o E51 o r F D S . D f 1 4 p 1 5 p y u y s == d o tt t == d u Q l o r dnnd n o r tt a ee B G nl aal a ee B G dnnt u ss - - eiffi f ss nl aal d oo e e n - - eiff u i DD l c eh nnh TCIIC oo e e eh nnd v o I DD l c TCIIA A o A i rr h rr h d n ii W ii w I AA AA dn d n aa eno aa eno EEEEEEE o t m cai t m NNNNNNN t em n t NNNNNNN N em cai n t NNNNNNN s B a a) c B a a) c e G t .e 0000000 0 G t .e 0000000 sir 0000000 0 s iMi sir iMi 0000000 o D( D 1 1 1 1 1 1 1 1 D( D 1 1 1 1 1 1 1 D n n n n n ono n ono o i oi o ioi i tit i tit t pt p t pt p a mpmn a mpmn nn oi umuo nn umuo susi oi susi i m nsnt i m nsnt sa onop sa onop rt CoCm rt CoCm enn C u enn C u mooe k s mooe ks y mCilkl n mCilkl n I tbli o y I tbli o a d aaiMC a daaiMC w enl t M l w enl t M h t muae uoh gwaa tt a h muae tt t t uoh gwaa a p l rneooe o a l rneooe pGI VCGM T P pGI VCGM Ili ,I I
TABLE II-5 Doses to the population from Continuous Gaseous Effluent Releases FIRST OUARTER Whole Body Dose . Organ Dose Pathway (Man-Rem) Orcan Dose (Man-Rem) plume Immersion 1.13E-03 Skin 3.77E-03. Ground contamination 1.25E-07 Skin 1.52E-07 Inhalation 9.71E-05 Thyroid 1.66E-04 Vegetable Consumption 6.04E-05 Thyroid 4.49E-04 Milk Consumption 1.30E-05 Thyroid 3.55E-04 Meat Consumption 1.21E-05 Thyroid 4.11E-05 Total 1.32E-03 . skin 3.96E-03 SECOND OUARTER Whole Body Dose Organ Dose Pathway (Man-Rem) Orcan Dose (Man-Rem) plume Immersion 1.19E-03 Skin 4.02E-03 Ground Contamination 6.74E-08 Skin 8.19E-08 Inhalation 4.42E-05 Thyroid 7.00E-05 Vegetable Consumption 1.78E-05 Thyroid 2.00E-04 Milk Consumption 6.66E-06 Thyroid 1.71E-04 Meat Consumption 4.26E-06 Thyroid 1.80E-05 Total 1.27E-03 Skin 4.09E-03 i 14
.F. TABLE II-6 Doses to the Population from Batch Gaseous Effluent Releases FIRST OUARTER f Whole Body Dose Organ Dose ' Pathway (Man-Rem) Orcan Dose (Man-Rem) 1 Plume Immersion 9.90E-04 Skin 3.61E-03 Ground Contamination 6.56E-09 Skin 7.97E-09 Inhalation 4.16E-08 ' Kidney. 5.45E-08 Vegetable Consumption 6.15E-08 Thyroid 2.31E-05 Milk Consumption 3.24E-08 Thyroid 1.60E-05 Meat consumption 7.15E-09 Thyroid 1.61E-06 , Total 9.90E-04 Skin 3.62E-03 SECOND OUARTER Whole Body Dose organ Dose. Pathway (Man-Rem) 010an Dose (Man-Rem) Plume Immersion 1.01E-03 Skin. 3.69E-03 Ground Contamination 5.04E-08 skin 6.12E-08 Inhalation 4.58E-08 Thyroid 1.96E-05 Vegetable consumption 1.64E Thyroid 9.29E-05 Milk Consumption 3.14E-07 Thyroid 1.79E-04 i Meat Consumption 1.74E-08 Thyroid 9.39E-06 ' Total 1.01E-03 Skin 3.C9E-03 i I l I J 15
III. RELEASES OF AND DOSES FROM LIQUID EFFLUENTS There are four' sources of liquid effluents released to the discharge canal:
- 1) the Laundry and Shower Sump Tanks, 2) the Evaporator Condensate Storage Tanks, 3) the Regeneration Waste Neutralization Tank, and 4) the condenser
<Hotwell. The Laundry Tanks and Evaporator Condensate Storage Tanks are batch type releases made through the plant liquid release monitor RML-2.
The Regeneration Waste Tank discharges are batch type, the Condenser Hot-well discharges are continuous types, both of which are made through plant liquid release monitor RML-7.
- 1. Reaulatory Limits The Technical Specification limits for liquid effluent releases are as follows: 1 Specification 3.11.1.1 The concentration of radioactive ' material released to UNRESTRICTED AREAS"shal? be less than or equal to the concentrations specified in 10 CFR_20, Appendix.B,. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, 'the ~ concentration shall be less than or equal to 2x 10-4 microcuries/ml total activity.
Specification 3.11.1.2 The dose or dose ccamitment to a MEMBER OF THE PUBLIC from radioactive material is in liquid effluents released to UNRESTRICTED AREAS shall be limited as follows:
- a. During any calendar quarter to less than or equal to 1.5' arem to the total body and less than or equal to 5 mrem to any organ.
- b. During any calendar year to less than or equal to 2 mrem to the total body and to less than or equal to 10 mrem to any organ.
l 16
- 2. R/idionuclide Concentrations The maximum permissible concentration values used in determining allowable liquid release concentrations are taken from 10 CFR 20, Appendix B, Table II, Column 2. Release rate and dilution ratio for each batch are determined by a mixed nuclide Mpc calculation performed before the release of the batch.
The concentration of each of the gamma emitting nuclides specifically noted in Figure III-1 is measured individually due to the requirements of the Technical Specification. Individual measurements are made on proportional composite liquid radwaste samples to determine the Fe-55, Sr-89 and Sr-90 concentration to be applied to individual release calculations. A distillation and liquid scintillation counting technique is used to measure the tritium concentration on each batch release. The measured and calculated concentration values for each batch are used to calculate the dilution ratio, release rate, and dilution rate, and expected doses prior to the release. The release data is then updated with actual release conditions and stored on a computer disc file. The disc file data is used to assure that quarterly and annual release limits are not exceeded. ) A. Fission and activation products - The total release values (not including tritium, gases, gross alpha) are comprised of the sum of the individual radionuclides activities released to the discharge canal for tne respective quarter. These values represent the activity known to be present in the liquid radwaste effluent. percent of applicable limit is determined by dividing the calculated total body or organ dose by the applicable Technical Specification limit and then multi-plying the result by 100. The most restrictive percent of limit is tnen used. B. Tritium - The measured tritium concentration is used to calculate the total release and average diluted concentration during each period. Average diluted concentration divided by the MpC limit, 3 x 10-3 uci/ml, is converted to give the percent of applicable limit. I l l l l l l l 17 { I
C. Dissolved and entrained gases - Concentrations of dissolved and ' en-trained gases in liquid effluents are measured by HpGe spectroscopy on a sample from each liquid release. Dissolved and entrained gases for which measured concentrations are determined include noble gases with half lives. greater than 8 hours. Iodine radionuclides in any form are determined during the isotopic analysis for each release, therefore a separate analysis for possible gaseous forms is not performed. The average diluted concentration of the dissolved and entrained gases is divided by the Mpc limit, 2 x 10-4 uCi/ml, and converted to give the percent of applicable limit.
- 3. Measurements and Approximations of Total Radioactivity Details of the analytical procedures for liquid radwaste analysis are as follows:
Measurement Frecuency Method
- 1. Gamma Isotopic Each Release HpGe spectrometry with on-line computer
- 2. Sr-89 Monthly composite Chemical separation and gas flow proportional counting
- 3. Sr-90 Monthly composite Chemical sepe. ration and gas flow proportional counting
- 4. Tritium Each Release Distillation and liquid scintillation counting
- 5. Alpha Monthly Composite Alpha scintillation
- 6. Dissolved Gases Each Release HpGe spectrometry with on-line computer
- 7. Fe-55 Monthly Composite Chemical separation and liquid scintillation counting Estimated errors are based on errors in counting equipment calibration, counting statistics, nonsteady release flow rate, chemical yield factors, sampling and mixing losses, and volume determinations.
A. Fission and Activation products Total Release as calculated for each batch. Statistical Error at MDA 60% Waste Volume 10% Counting Equipment Calibration 10% Sampling and Mixing _2Q1 100% l 18
B. Total Tritium Release as calculated from a monthly composite. Waste Volume 10% Counting Equipment Calibration 10% Sampling and Mixing _2Q1 40% C. Dissolved release. and Entrained Gasce Total Release as calculated from each Statistical Error at MDA 60% Waste Volume 10% Counting Equipment Calibration 10% Sampling and Mixing _2Q1 100t D. Total Gross Alpha Radioactivity Release as calculated from a monthly composite. L Statistical Error at MDA 60% l Waste Volume 10% i Counting Equipment Calibration 10% Sampling and Mixing _2Q1 100%
- 4. Batch and Unclanned Releases The batch liquid effluent releases may be summarized as follows:
QUARTER 1 QUARTER 2 Number of Batch Releases 1.20E+02 1.12E+02 Total Time for all Releases (ainutes) 1.47E+04 1.59E+04 Maximus Time for any one Release (ainutes) 2.67E+02 3.92E+02 I Average Time for all Releases (ainutes) 1.22E+02 1.42E+02 Miniaua time for any one Release (minutes) 2.30E+01 4.70E+01 Average dilution flow of Units 1, 2, and 3 during all Releases (liters / minute) 4.78E+06 4.14E+06 Number of Unplanned Releases 0.00E+00 0.00E+00 Total Unplanned Activity Releases (Curie) 0.00E+00 0.00E+00 t {
]
The summation of liquid effluent releases is in Table III-1 and the suasa- . tion of nuclides in liquid effluent releases is in Table III-2. These re- j leases are based on the dilution of the radioactive liquid effluent by the ~ nuclear services sea water of Unit 3. The doses to individuals from liquid effluent releases are in Table III-3 and the doses to the population from liquid effluent releases are in Table III-4. These doses are based on the dilution of the radioactive liquid effluents by the condenser cooling water of Units 1, 2, and 3. i l 1 19 ! j l
--________=___ _ _ _ _ _ - _ _ k
Figure III-1 . METHODS OF MEETING 10 CFR 20, APPENDIX B, TABLE II, COLUMN 2 MPC LIMITS r MPC RANGE GAMMA-RAY BETA ALPHA fuci/ml) EMITTERS FMITTERS EMITTERS I-131. I-132. I-133 Sr-89, Sr-90 I-135. Cs-134 (Separation and Beta (9 x 10-6 Scintillation Counting) , HpGe Gamma-Ray Fe-55 . Spectroscopy) (Separation and All Liquid Scintil- (Alpha lation Counting) Counting Sensitivity 10-7 uCi/ml as Pu-239) Ba-La-140. Na-24. Cu-64 Tritium Co-60, Fe-59. 2n-15 (Distillation and Liquid Scintillation 29 x 10-6 Ao-110m Mn-54, Co-SS Counting 10-5 uci/ml) Zr-Nb-95. Cs-Ba-137 As-76 F-18 Cr-51 No-239, Ce-141 Mo-Tc-99, Ce-Pr-144 l HpGe Gamma-Ray l Spectroscopy) l l 20 l l
i i' .) > TABLE III-1 ,
' 1 1
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT - 1/1/87 - 6/30/37 l LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES ' Unit Quarter Quarter Est. Total 1 2 1 Error t
- t. . Fission and Activation Products s
- 1. Total Release (not including tritium, s 1
gases, alpha) Ci 6.82E-02 4.60E JP 1.00E+02 f 2. Averakedilutedconcentrationduring perio uCi/ml 1.57E-08 '1.f1E-08)
- 3. Percent of applicable limit t 1.65E-01 2.?OE-01 B. Tritium '
i [~ ~
- 1. Total Release Ci 9 86E+01 6.15E+01 4.00E+01
- 2. Average diluted concentration during i per1M uCi/ml 1.48E-05 8.78E-06
- 3. Percent of applicable limit t 4.932-0+, 2.93E-01
~
C. Dissolved and entrained gases '
- 1. Total release C'i ' 3.$6E+00 1.17E+01 1.00E+02
\ ^ r. -
j
- 2. Averakedilutedconcentrationduring 4 i peric uCi/ml 5.57E-07 7.34E-06 !
.r.
- 3. Percent of applicable limit \ h570E-01 2;79E C%
_ . ~ _ D. Gross alpha radioactivity I *
,/
l
- 1. Total release ci <C. AJE-05 '
1.50E-05 1.00E+02 l l 1 J i j m-_. . ' . I { E. Volume of Waste released (prior to dilution) ' { c (.-- ;
- 1. Batch and Continuous Modes Liters 1.59E+C7 0.02E+06 1.00E+01 i t l
{ F. Volume of dilution water used during period !
+
l
- 1. Batch and Continuous Modes Liters 2.4GE+16 2.26E+10 1.00E+G1 l#
i
/
x 21 L
T o,_. s)" ,,.
?, <
4
.v ' / r, 1 / ; ,i 1
s- . A . g s I ' TABLE'III-2
- g. , .
. , EFFLUE~iT M4 ( bSTE DISPOSA$. SEh (NUAL REPORT - 1/1/87 - 6/30/8
, ..a / "' LI. QUID EFFLUENTS i, . , s CONTINUGUS MODE BATCH MODE sj ,, +- 4 tiuclides Released Unit Quarte; f_ ' 4:$rter 2
' Quarter 1 Quarter 2
.__,a._= '
andium-24 Ci ,_ _ L o O;- 3 , , , chromium .} 1..a.__ Ci J .3 E-03 ' 1.46E-03
, ED9P as;ie-54 ' ,,,;. _ Ci.,__ _ 1.17E-04 6.62E-04 1.UIL-55-- _
. _31. . , , _ _
' _ J.167E-03 1.31E-03 17.m d L r.E .
srb.g l,t-58 Ci T
<i.11E_02 6.53E-03
.copalt-60 Ci__h' _ ,;. '
6.58E-03 1.83E-02
)
une-65 _ _ _ ct_ *a , _ . , ,,,, Imd;11;n-88 . Ci '
<f 4_, 7.95E-04 .
Stronti'im-09 , , _ , ci 1.23E-05 1.30E-03 3.41E-04 strontrm-90 ,,, _ _ Ci 1.34E-04 5trontium-92 C1 5.28E-04 8.91E-04 Iirrqal.um- 95.. ,- %_ Ci 3 . 5 7 E-05.___ L _ 11710nium-37 j_ ,_ . ,,.Ci__ ,,_ , _ _ 7.39E 8.81E-05 , nicblum-SS _
<Ci _ , . ,
.,,_. 2.16E-04 1.26E-04 AQl'/bde;wm-99 _,_J1_._,_, .. 2 . ,,_. ___
hEhnCli.ltm- 99m .,_ Ci .m 14,,,_,, __.j_.02E-05 4.61E-04 tjtEllDi;, tium-i01 _._ Ci g_ _ ' y' n'?Agn,ium-103
' ?
ci _ i
~1 86E-05
>$ily,st-110m Ci 1.50E-03 2.56E-03
,@,llurium-132 _ _ _ , . Ci - _ , , 1.05E-04 6.92E-06 ip, dine-131 Ci 1.24E-03 1.85E-02 lodine-132 . .;,_ Ci ,_,.
iodine-133 ..,,_ C i __ _ _ 6.21E-05 1.65E-04 rggium-13J._
~
, .L_C1 .,_ ' l
_u 3.35E-03 6.67E-04
%!sium-13f., __ l Ci .b gesium-137 C1 1.17E-04 4.12E-03 1.79E-03 ces ing .,11P. - , , _ 2m 23E-05 Barium-139 Ci 1.24E-04 lanthanum-140 Ci 1.25E-04 cerium-141 , , _ ,
Ci Cerium-144 ,, 4__-
- l_.. __
. k- l .l unidentified ,
Ci 1.91E-03 0.00Es00 3.13E-02 8.74E-03 I" 1 Total for period (above) Ci 2.04E-03 0.00E-00 6.62E-02 4.61E-02 22
l TABLE III-2 (Continued) EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT - 1/1/87 - 6/30/87 LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE Dissolved & Entrained Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 Gases xenon-133 Ci 3.77E+00 1.14E+01 xenon-135 Ci 2.21E-02 7.55E-03 xenon-133m Ci 3.01E-02 5.18E-02 xenon-131m Ci 3.12E-02 1.84E-01 krvoton-85m Ci 3.04E-04 l krvoton-85 Ci 1.88E-02 i krypton-88 Ci 4.22E-05 i f tritium Ci 3.31E-02 9.86E+01 6.15E+01 l l NOTE: There were no continuous releases during the second quarter. l 23
Z.ZZ:2J~ Z-~ ._- - FIGURE 3IT-252 Z:.Z-~;;r Z ~- ~ ~ ' t LIQUIIFRELEASES-FISSIONIAND ACTIVATION-PRODUCTS- - - - - - (EXCLt1DINGrit37 GASES 7 ALPHA) q..
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1 82 f983 1984 1985 1986 1987 1988 24
TABLE III-3 Doses to Individuals from Liquid Effluent Releases FIRST OUARTER - CONTINUOUS RELEAMES Whole Body Dose' Oraan Dose Dose Dose Pathway Ace Group 13xgg) Ace Grouc Organ (area) Fish Adult 2.81E-06 Teen Liver 4.28E-06 Invertebrates Adult 4.26E-07 Child Bone 8.38E-07 Shoreline Use- Teen 2.05E-06 Teen Skin 2.39E-06 i l Total Teen 3.80E-06 Teen Liver 6.97E-06 SECOND OUARTER - CONTINUOUS RELEASES Whole Body Dose Oraan Dose Dose Dose Pathway Ace Group faren)_ Ace Grouc Organ (mren) Fish Invertebrates Shoreline Use No second Quarter continuous releases Total FIRST GUARTER - BATCH RELEASES Whole Body Dose Oraan Dose Dose Dose Pathway Ace Group (aren) Ace Grouc Oraan (mrem) Fish Adult 4.05E-04 Adult GI-LLI 4.61E-03 Invertebrates Adult 2.28E-04 Adult GI-LLI 3.60E-03 Shoreline Use Teen 3.72E-04 Teen Skin 4.36E-04 Total Teen 8.27E-04 Adult GI-LLI 8.27E-03 SECOND OUARTER - BATCH RELEASES Whole Body Dose Orcan Dose Dose Dose Pathway Ace Group (arem) Ace Group l Organ farem) l Fish Adult 2.73E-04 Adult GI-LLI 7.00E-03 Invertebrates Adult 2.84E-04 Adult GI-LLI 3.88E-03 Shoreline Use Teen 8.48E-04 Teen Skin 9.97E-04 t Total Teen 1.32E-03 Adult GI-LLI 1.10E-02 l j 25
TABLE III-4 Doses to the Population from Liquid Effluent Releases FIRST OUARTER - CONTINUOUS RELEASES Whole Body Dose Organ Dose Pathway (Man-Rem) Organ Dose (Man-Rem) Sport Fish 3.99E-04 Liver 7.61E-04 Commercial Fish 4.12E-07 Liver 7.85E-07 Sport Invertebrate 1.70E-05 Bone 3.27E-05 Commercial Invertebrate 1.22E-07 Bone 2.34E-07 Shoreline Use 9.19E-06 Skin 1.07E-05 Swimming 1.39E-08 Thyroid 1.39E-08 Boating 1.39E-08 Thyroid 1.39E-08 Total 4.26E-04 Liver 7.94E-04 SECOND OUARTER - CONTINUOUS RELEASES Whole Body Dose Organ Dose Pathway (Man-Rem) Organ Dose (Man-Rem) Sport Fish Commercial Fish Sport Invertebrate Commercial Invertebrate NO SECOND QUARTER CONTINUOUS RELEASES Shoreline Use Swimming ' Boating Total FIRST OUARTER - BATCH RELEASES Whole Body Dose Organ Dose Pathway (Man-Rem) Organ Dose (Man-Rem) Sport Fish 6.06E-02 GI-LLI 6.63E-01 Commercial Fish 6.23E-05 GI-LLI 6.71E-04 Sport Invertebrate 9.78E-03 GI-LLI 1.10E-01 Commercial Invertebrate 6.88E-05 GI-LLI 7.37E-04 Shoreline Use 1.66E-03 Skin 1.95E-03 Swimming 8.99E-06 Thyroid 8.99E-06 Boating 8.99E-06 Thyroid 8.99E-06 Total 7.21E-02 GI-LLI 7.74E-01 SECOND OUARTER - BATCH RELEASES Whole Body Dose Organ Dose Pathway (Man-Rem) Organ Doce (Man-Rem) Sport Fish 4.37E-02 GI-LLI 1.02E-00 l Commercial Fish 4.49E-05 GI-LLI 1.04E-03 Sport Invertebrate 1.40E-02 GI-LLI 1.56E-01 Commercial Invertebrate 1.00E-04 GI-LLI 1.11E-03 Shoreline Use 3.80E-03 Skin 4.46E-03 Swimming 1.63E-05 Thyroid 1.63E-05 Boating 1.63E-05 Thyroid 1.63E-05 Total 6.16E-02 GI-LLI 1.18E+00 l 26 l
IV. SOLID WASTE SHIPMENTS Solid waste shipments from the plant may include solidified liquid waste, dry compressed waste, spent resins, irradiated components and spent fuel.
- 1. Reaulatory Limits The Technical- Specification for solid waste shipment reporting is as follows:
Specification 6.9.1.5(d) The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:
- 1. container volume,
- 2. total curie quantity (specify whether determined by measurement or estimate),
- 3. principle radionuclides (specify whether determined by measure-ment or estimate),
- 4. type of waste (e.g., spent resin, compacted dry waste, evaporator bottons),
- 5. type of container (e.g., LSA, Type A, Type B, large quantity),
and
- 6. solidification agent (e.g. . cement).
The summation of solid waste and irradiated fuel shipments is presented in
. Table It'-1 and Shipment Summaries in Table IV-2.
27
TABLE IV-1 EFFLUENT AND WASTE DISPOSAL SEMfANNUAL rep 0RT - 1/1/87 - 6/30/87 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Non irradiated fuel) First
- 1. Type of waste Unit 6-month Est. Total period Error, %
- a. Spent resins, filter sludges, evaporator m3 7.25E+01 bottoms, etc. Ci 1.98E+02 2.00E+01
- b. Dry compressible waste, contaminated m3 6.44E+01 equip, etc. Ci 2.20E+00 5.00E+01
- c. Irradiated components, control m3 E+00 rods, etc. Ci E+00 5.00E+01
- d. Other (describe) Solidified liquid m3 4.67E+00 Ci 3.48E-03 2.00E+01
- 2. Estimate of major nuclide composition (by type of waste)
- a. Co-58 3.55E+01% Cr-51 6.34E+00% E *a Cs-137 1.94E+01% Ni-63 4.88E+00% E *s co-60 1.43E+01% Sb-122 2.68E+00% E t Cs-134 1.04E+01% Mn-54 1.77E+00% E %_.
- b. Co-60 5.24E+01% Co-58 1.45E+00% E %
Ni-63 3.78E+01% Ma-54 1.40E+00% E % Fe-55 2.74E+00% E % E Cs-137 2.46E+00% E % E '6
- c. E % E *.
E % E *. E '6 E S.
- d. Co-60 2.67E+01% C-14 1.53E+01% Cs-134 2.22E+00%
Fe-55 2.17E+01% H-3 7.41E+00%
]
Ni-63 1.58E+00%
]
Cs-137 2.08E+01% Sb-125 3.40E+00% E S. l 1
- 3. Solid Waste Disposition j Number of Shioments Mode of Transportation Destination l 12 Transport Truck - Chem-Nuclear Systems, Inc.
Exclusive Use Vehicle Barnwell, SC l B. IRRADIATED FUEL SHIPMENTS (Disposition) Number of Shioments Mode of Transportation Destination None NA NA l l l 28 j
- TABLE IV EFFLUENT AND WASTE O!SPOSAL SEMIANNUAL REPORT 1/1/87 - 6/30/87 SHIPMENT
SUMMARY
DATE AND CONTAINER TOTAL WASTE CONTAINERS SOLIDIF. SHIPHENT # VOLUME CURIES PRINCIPLE RADIONUCLIDES TYPE TYPE AGENT 1-06-87 87-1 402 3.7E-4 Cs-134, Cs-137, I-131, H-3, Sb-122 SC LSA N.A. 1-20-87 Cs-134, Cs-137, Co-58, Co-60, I-131 87-4 119 31,5 N1-63, Hn-54, Cr-51 F LSA N.A. 1-22-87 10 0 92 87-7 2 9 107 1.478 Cs-137, Co-60, Ni-63, Hn-54, Fe-55 NW LSA N.A. < 1-29-87 87-8 402 3.81E-4 Cs-134, Cs-137, I-131, H-3, Sb-122 SC LSA N.A. 2-05-87 Cs-134, Cs-137, Co-58, Co-60 Hn-54, 87-10 119 92.8 Sb-122, Cr-51, Nb-95, 2r-95, Ag-110m F LSA N.A. 2-20-87 Cs-134, Cs-137, CO-66 N1-63, H-3 87+12 193 1.81 Sb-122 SR LSA N.A. 2-26-87 87-14 402 1.70E-3 Cs-134, Cs-137, I-131 H-3 SC LSA N.A. 3-5-S7 Cs-137, Co-58, Co-60, Ni-63, Hn-54, 87-15 86 9 7.5 6,49E-1 Fe-55 CW LSA N.A. 4-16-87 87-18 402 6.28E-4 Cs-134, Cs-137, H-3 SC LSA N.A. 6-03-87 Cs-134, Cs-137, Co-58, Co-60, Ni-63, 87-20 22 Q 7.5 3.48E-3 Hn-54 Fe-55 SL LSA Cement 6-03-87 Cs-134 Cs-137, Co-60, N1-63, C-14, 87-20 64 0 7.5 7,56E-2 H-3, Fe-55, Sb-125 CW LSA N.A. 6-19-87 87-21 402 6.72E-4 Cs-134 Cs-137, I-131 H-3 SC LSA N.A. 6-23-87 Cs-134 Cs-137, Co-58, Co-60, Ni-63, 87-22 119 71.7 Hn-54 Sr-89 SR LSA N.A. i t.*ASTE TYPE: SR - Spent Resin NW - Non-Compacted Waste CE - Contaminated Equipment SC - Secondary Resin CW - Compacted Waste IC - Irradiated Components F - Filters EB - Evaporator Bottoms SL - Solidified Liquid l l l l l i 29 i
i L - FIGURE IV l. VOLUE OF SOLID WASTE SHIPPED OFF-SITE 800 600-Waste V mglume, 408
-"~'
400-254 250 289 280
- 223 218
- 204 200 188 160 137 A B A B A B A B A B A 1982 1983 1984 1985 1986 1987
- FIGURE IV RATIO 0F CURIE COMEN TO VOLUE OF WASTE SHIPPED OFF-SITE
- 9. 15.76 8.0 "
7.5-hh 7.0 -- 6.5-6.0 " 5.64
~
- 5. 5 -
5.0 - 4.5 Ci/m3 4.0 ' 3.5 - 3.0 - 2.5 " 2.0 1*57 1.5 1.46 1.23 1.0 0.56
- 0. 5 , 0.21 t i l J
t A B A B A B A B A B A l 1982 1983 1984 1985 1986 1987 l 1 A - First Half B - Second Half I 30 I
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V METEOROLOGICAL DATA ~ L- The meteorological data at 33 feet (10 meters) will be summarized in Tables V-1, V-2, V-3, and V-4 in the year end report. l The classification of atmosphere stability is as follows: Stability Pasquill Temperature change classification catecories with heiahtf*c/100m) Extremely unstable A <-1.9 Moderately unstable B 2-1.9 to-<-1.7 Slightly unstable C 1-1.7 to <-1.5 Neutral D 2-1.5 to <-0.5 Slightly stable- E 2-0.5 to < 1.5 Moderately stable F 2.1.5 to < 4.0
- Extremely stable .G 1 4.0 The data recovery rates are:
First Quarter ...... 98% Second Quarter ..... 82% 31
VI TECHNICAL SPECIFICATION REPORTS Technical Specification Sections 3.3.3.8, 3.3.3.9, and 3.12.1.2.b require reporting out of specification conditions in the Semiannual Radioactive Effluent Release Report. There were no reports as required by the above specifications for the period of this report.
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VII ODCM AND PCP CHANGES Technical Specifications Section 6.9.1.5.d requires reporting of changes to the offsite Dose calculation Manual (ODCM) and the Process Control Program (PCP) in the Sealannual Radioactive Effluent Release Report. Revisions to the ODCM and PCP for the period of this report are included in this section. The changes are marked for ease of review. I 33
Flow Rates (Variable - based on setpoint needs, nominal or maximum values listed below.) '
- 1) Reactor Building Purge Exhaust Duct = 50,000 cfa =
2.4 x 107 cc/sec ,
- 2) Auxiliary Building and Fuel Handling Area Exhaust Duct = 156,000 cfm = 7.4 x 307 cc/sec l
- 3) Waste Gas Decay Tank Release Line = 50 cfm max = 2.4 i x 104 cc/sec (X/Q) = 2.5 x 10-6 sec/m3 For all vent releases. The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary.
In order for a gaseous release to be within the limits of specification 1.1-1, the following equations must be satisfied: Eq. 4 Dose Rate + 500 1 1 (Total Body) Eq. 5 Dose Rate + 3000 1 1 (Total Body) l Eq. 6 Dose Rate + 1500 1 1 (I, T, P) where: 500 = The allowable total body dose rate due to noble gas gamma emissions in area /yr. i 3000 = The allowable skin dose rate due to noble gas beta emissions in aren/yr. 1500 = The allowable organ dose rate in aren/yr. Equations 1, 2, and 3 are solved for each release type and release point currently releasing or awaiting release. If equations 4, 5, and 6 are satisfied, the release can be made under the assumed flow rates. If one or more of equations 4, 5, and 6 are not satisfied, action must be taken to reduce the radionuclides release rate prior to initiating a release (or to reduce the radionuclides release rate already in progress). l The following actions are available to reduce the release rates at the three l release points. l
- 1) Waste Gas Decay Tanks a) Release Valve may be throttled b) Tank contents may be diluted c) Release may be delayed for longer decay time.
l 1 o_ _ _
I i Setpoint Calculation 1.4-1 Reactor Building Purge Exhaust Duct Monitor (RM-A1) (Batch Type Releases) INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if the nuclide concentration or release rate exceed the limits determined. METHODOLOGY Reactor Building atmosphere is circulated through radiation monitor RM-A6 (contain-cont atmosphere noble gas monitor) and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM-A1 (Reactor Building purge exhaust duct monitor), and factors are applied to account for background radi-ation and statistical counting variations. The obtained value is used as the ale,rm/ trip setpoint and RM-L: is adjusted to this value prior to initiating the release. CALCULATION Monitor Setpoint (CPM) = Net CPM x Vent Fraction Vac RM-A6 Conc. RM-A6/CPN + Bkg 3.3/Bkg PDRR x Admin. Limit Vac RM-A1 Conc. RM-A1/ CPM wheret Net CPM = The observed monitor count rate, in counts per minute. Vent Fraction = The portion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1. The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1. PDRR = The noble gas gamma emission Projected Dose Rate Ratio (PORR) calculated in accordance with Section 1.3. This-ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1 equation 4. See Note. Admin. Limit = Administrative Limit to reduce setpoint to 10% of allowable limit. Admin. Limit = 10. Vac RM-A6 = The actual vacuus reading at RM-A6 at the time of sampling. Vac RM-A1 = The actual or average vacuun reading at RM-A1 during normal operation. Conc. RM-A6/ CPM = The ratio of uci/cc to cpu at RM-A6 based on an actual sample or derived from the calibration curve. l Conc. RM-A1/ CPM = The ratio of uci/cc to cps at RM-A1 based on an actual sample or derived from the calibration curve. BKG = The background count rate at RM-A1 in counts per minute (cpa).
L Cetpoint Calculation 1.4-1A Reactor Building Purge Exhaust Duct Monitor (RM-A1) ' (Special Release for Functional Testing of the Reactor Building Purge System) l INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alars and pathway isolation occur if the nuclide concentration or release rate exceed the' limits determined. METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere is continuously passed through radiation monitor RM-A2 and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM-A1, and factors are applied to account for background radiation and statistical counting variations. The obtained value is used as the alara/ trio setpoint and RM-A1 is adjusted to this valve prior to initiating the release. CALCULATION Monitor Setpoint (cpm) = Net CPM x Vent Fraction Vac RM-A6 Conc. RM-A6/ CPM PDRR x Admin. Limit Vac RM-A1 Conc. RM-A1/ CPM
+ Bkg + 3.3 @
where: Net cpm = The observed monitor count rate, in counts per minute. Vent Fraction = The portion of the total plant gaseous release associated with this vent and discharge type. Value can be set.to a number between 0 and 1. The summation of the vent fractions of'RM-A1 and RM-A2 cannot exceed 1. PDRR = The noble gas ganna emission Project Dose Rate Ratio (PDRR) calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1 equation
- 4. See Note.
Admin. Limit = Administrative Limit to reduce setpoint to 10% of allowable limit. Admin. Limit = 10. Vac RH-A6 = The actual vacuun reading at RM-A6 at the time of sampling. Vac RM-A1 = The actual or average vacuum reading at RM-A1 during normal operation. Conc. RM-A2/ CPM = The ratio of uci/cc to cpa at RM-A2 based on an actual , sample or derived from the calibration curve. j Conc. RM-A1/ CPM = The ratio of uci/cc to cpa at RM-A1 based on an actual sample or derived from the calibration curva.
-18a-
Setpoint Calculation 1.4-2 Reactor Building Purge Exhaust Duct Noaitor (RM-A1) (Continuuus Type Releases) INTRODUCTION Following completion of the analyses required by Section 1,2-1 and ' determination .
.of release ' rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway . isolation occur if the nuclide concentration or release' rate exceed the limits determined.
NETaonoLOGY Reactor - Building atmosphere is passing through radiation monitor RM-A1 during a
' continuous type release. Factors are applied to the observed count rate to account 'for background radiation and statistical counting-variations. The obtained value
- is used as .the alara/ trip setpoint and RM-A1 is adjusted to this value weekly
'during continuous releases.
1 CALCUL& TION Monitor Setpoint (CPM) = Net CPM x Vent Fraction PDRP. x Admin Limit + Bkg + 3.3 (Bkg where: Net cpm = The observed monitor count rate, in counts per minute. Vent Fraction = The portion of.the total plant gaseous release associ- } ated with this vent and discharge type. Value can-be set to a number between 0 and 1. The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1. PDRR = The noble gas gamma emission Projected Dose Rate Ratio j (PDRR) calculated in accordance with Section 1.3. This i ratio is the actual projected dose rate divided by the ! allowable dose rate referenced in Section 1.3-1 equa- l tion 4. See note. ' Admin. Limit = Administrative Limit to reduce setpoint to 10% of allowable lizit. Admin. Limit = 10. BKG = The background count rate at RM-A1 in counts per minute (cpa). l , 3.3 (~BKg = A statistical spread on the background count rate which 3 represents a 99.95% confidence level on monitor j counting. This factor is included to prevent inadver- l tent high/ trip alaras due to randos counts on the moni- { tor. Only the positive (+) sida of the spread is i applied, l p NOTE: If there are no isotopes identified in the sample, the LLDs for Xe-133 and Kr-85 may be used as actual values for the purpose of the setpoint calculation.
-1S-(
Setpoint Calculation 1.4-3 Auxiliary Building & Fuel Handling Area Exhaust Monitor (RM-A2 ) (Continuous Type Releases) INTRODUCTION Following completion of the analyses required by Section 1.2-2 and determination of release rates and concentration limits in accordance with Section 1,3-1, the r monitor setpoint requires adjustaent to assure that alarm and pathway isolation ' occur if the nuclide concentration exceeds the limits determined. METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere' is continuously passing through radiation monitor RM-A2. Factors are applied to the observed count rate to account for background radiation and statistical counting variations. The obtained value is used as the alara/ trip setpoint and RM-A2 is adjusted to this value weekly. CALCULATIVE Monitor Setpoints = Net cpm x Vent Fraction (CPM) PDRR x Admin Limit + Bkg + 3.3 VBkg where: Net CPM = The observed monitor count rate, in counts per minute. Vent Traction = The portion of the total plant gaseous release associ-ated with this vent and discharge type. Value can be set to a number between 0 and 1. The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1. PDRR = The noble gas gamma emission Projected Dose Rate Ratio I (PDRR) calculated in accordance with Section 1.3. This l ratio is the actual projected dose rate divided by the J allowable dose rate referenced in Section 1.3-1 equa-tion 4. See note. Admin. Limit = Administrative Limit to reduce setpoint to 10% of allowable limit. Admin. Limit = 10. BKG = The background count rate at RM-A2 in counts per minute (cpa). l 3.3 VBKg = A statistical spread on the background count rate which represents a 99.95% confidence level on monitor counting. This factor is included to prevent inadver- I tent high/ trip alarms due to randon counts on the moni- l tor. Only the positive (+) side of the spread is I applied. NOTE: If there are no isotopes identified in the sample, the LLDs for Xe-133 and Kr-85 may be used as actual values for the purpose of the setpoint calculation. l l
Representative Sampling Method No. 3.1-1 (Evaporator Condensate Storage Tanks, Laundry & Shower Sump Tanks, Secondary Drain Tank) To obtain representative samples from these tanks, the contents of the tank to be sampled will be recirculated through two contained volumes and a grab sample will be collected upon completion. No additions of liquid waste will be made to this tank until completion of the release. Representative Sampling Method No. 3.1-2 (Secondary Drain Tank and/or Plant Condensate) A representative sample may be obtained via grab sample of the Turbine Building Sump or the Secondary Drain Tank, Plant Condensate, or from the release compositor. Representative Sampling Method No. 3.1-3 (Waste Gas Decay Tank) Represent'ative gas, iodine, and particulate samples are drawn irom the waste gas decay tank sample lines. . No' additions of waste gas is allowed into a tank following sampling until the release has been completed. Representative Sampling Method No. 3.1-4 (Reactor Building & Auxiliary Building & Fuel Handling Area Exhaust) Representative gas, iodine, particulate and tritium samples arc taken f rom these ducts at tne location of the radiation monitors. The sample for the Reactor Building purge Duct is taken from radiation monitor RM-A6 prior to a purge and is drawn from radiation monitor RM-A1 during a purge. The sample for the Auxiliary Building and Fuel Handling Area Exhaust Duct is drawn from RM-A2 during venting since this is a continuous release pathway. If samples cannot be obtained from the ducts of the Reactor or Auxiliary Building, samples can be obtained from areas of these buildings that are considered to be representative of the radionuclides concentrations present throughout the respective buildings. Sampling times end volumes should be established to assure the LLD Limits of Sections 1.2 and 4.2 for the radionuclides can be met. I l 4.0 ADMINISTRATIVE CONTROLS 4.1 Responsibility / Revisions Changes to the Process Control Program (PCP) are the responsibility of the Manager, Site Nuclear Services. Technical Specification 6.14 stipulates the required approvals necessary to modify the Process Control Program prior to implementing any changes to the process (See Section 3.2). 4.2 Reporting Changes to the PCP are submitted to the Nuclear Regulatory Commission in the next Semiannual Radioactive Effluent Release Report as per Techt.. cal Specifi-cation 6.9.1.5 (d). Reporting of nonconformances are to be done in accordan:e with the applicable Plant procedures. 4.3 Documentation All documentation associated with the verification of the Process Control Program is controlled in accordance with the appropriate implementing proce-dures. 4.4 Definitions Batch - (1) For sampling or processing, a batch is the largest homogeneous volume of waste that has been recirculated and controlled as per the PCP. (2) For solidification testing, a batch is taken to be a disposal
, container (i.e. , 55 gallon drums, etc. ) utilized in the solidi-fication of the waste.
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-s- LEVIS C N 1
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7 h syf.ag. ;c coee se c,m Florida Power CORPORATION August 26, 1987 3F0887-20 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Semiannual Radioactive Effluent Release Report
Dear Sir:
Pursuant to Title 10, Code of Federal Regulations, Part 50.36(a)(2) and Crystal River Unit 3 Technical Specification 6.9.1.5(d), Florida Power Corporation hereby submite the Crystal River Unit 3 Semiannual Radioactive Effluent Release Report for the period January 1, 1987 through June 30, 1987. If you have any questions concerning this matter please , contact this office. ! Sincerely, {
/ '
E.C. Simpson Director Nuclear Operations Site Support REF/dhd I xc: Dr. J. Nelson Grace Regional Administrator, Region II Mr. T.F. Stetka 77b Senior Resident Inspector
,[f Post Of fico Box 219
- Crystal River, Flonda 32629 + Telephone (904) 795-3802 l
A Flonda Progress Company
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- - - _ _ - _ - _ _ _ $}}