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G Radiological Effluent Technical Specifications (RETS) The U.S. Nuclear Regulatory Commission (USNRC) requires that op- erators of nuclear plants and fuel-cycle facilities monitor and report on releases of radioactive effluents. For nuclear plants, he monitoring and reporting system is specified in the Radiological Effluent Technical Specifi- cations (RETS). RETS requires the licensee to monitor effluent releases at every signifi- cant release point at the facility. Effluent monitoring consists of continuous measurements of some effluent streams; periodic measurement of radioac- tive particles trapped on filters, and measurement of samples from effluents released in batches. Detailed information about the RETS program for a given plant is contained in the licensee’s Offsite Dose Calculational Manual (ODCM), which is part of an operator’s application for a USNRC license. The USNRC also requires that the licensee participate in an Interlaboratory Comparison Program to ensure the accuracy and precision of the licensee’s data and also to carry out computational checks, data validation activities, and audits by USNRC personnel. Methods for estimating airborne and liquid effluent dispersions from nuclear plants are described in Regulatory Guides 1.111 (Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors) (USNRC, 1977a) and Regulatory Guide 1.113 (Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for The Purpose of Implementing Appendix I) (USNRC, 1977b), whereas methods used to derive the radionu- clide concentrations in foodstuffs from the air and water concentrations are described in Regulatory Guide 1.109 (Calculation of Annual Doses to Man 359
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360 APPENDIX G from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I) (USNRC, 1977c). Guidance to calculate the annual doses to humans from effluent releases from nuclear plants is also included in Regulatory Guide 1.109. Regulatory Guide 4.16 (Monitoring and Reporting Radioactive Mate- rials in Liquid and Gaseous Effluents from Nuclear Fuel-Cycle Facilities) indicates that estimates of exposures resulting from effluent releases from nuclear fuel-cycle facilities also should be calculated consistent with the ap- plicable guidance in Regulatory Guide 1.109. Alternatively, nuclear facility licensees can use Guide 4.20 (Constraint on Releases of Airborne Radioac- tive Material to the Environment for Licensees Other than Power Reactors) for estimating exposures from airborne releases. Of course, the nuclides of interest for exposures from nuclear fuel-cycle facilities differ from those for nuclear plants (see Chapter 2). The use of U.S. Environmental Protection Agency-approved codes (e.g., COMPLY) is accepted by the USNRC and these codes are generally used by fuel-cycle facilities to demonstrate com- pliance with exposure limits. These codes are generally conservative and overestimate exposures. Since external exposures from fuel-cycle facilities are essentially negligible compared to internal exposures, current models available in the literature are entirely sufficient. Similarly, current models are also sufficient for direct radiation exposure from stored waste, tailings piles, and depleted-uranium canisters. G.1 EFFLUENT MONITORING AT NUCLEAR PLANTS Regulatory Guide 1.21 (Measuring, Evaluating, and Reporting Radio- active Material in Liquid and Gaseous Effluents and Solid Waste) provides regulatory guidance for sampling and analysis of effluents from USNRC- licensed nuclear plants. Guidance to plant licensees on sampling and analy- sis methods and frequencies are provided in NUREG-1301 for Pressurized Water Reactors and NUREG-1302 for Boiling Water Reactors. These docu- ments contain guidance on: • Effluent monitoring instrumentation: Locations of monitoring instrumentation with respect to plant effluent systems, mini- mum number of operable channels, and surveillance (inspection) requirements. • Effluent monitoring: Sampling and analysis frequency, type of anal- ysis, and detection limits. Site-specific monitoring programs can deviate from the guidance in these NUREGs with appropriate justifications and approvals.
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361 APPENDIX G Regulatory Guide 1.21 recommends that licensees monitor all locations at the plant at which >1 percent of activity is discharged as: • liquid effluent, • noble gases into the atmosphere, or • anything else into the atmosphere. Title 10, Part 50 of the Code of Federal Regulations (10 CFR 50.36(a)(2)) requires licensees to report the principal radionuclides in effluent releases. These locations are referred to as significant release points and include vents and stacks for airborne effluents and liquid waste discharge points for liquid effluents. Releases are assessed using a combination of sample analyses, radiation monitoring, and flow, tank level, and system pressure indications, as appropriate, to ensure that the amount of radioactive mate- rial is not underestimated. Licensees are also required to monitor unplanned leaks and spills. If such leaks and spills result in offsite releases, then the magnitude of the releases must be estimated and reported to the USNRC along with the releases from routine operations. If the leak or spill occurs onsite, then a bounding analysis can be used to assess the potential offsite hazard. Continuous effluent releases are typically monitored by measuring gross radioactivity with a continuously indicating radiation monitoring system such as a sodium iodide detector. These gross measurements can be used to activate alarms and terminate effluent releases if radioactivity levels ex- ceed allowable limits. These continuous measurements are combined with analyses of physical samples (e.g., particulate materials trapped on filters or air samples) from the effluent stream to obtain quantitative estimates of the radionuclide concentrations in the effluent stream. Such samples are usually taken at specified frequencies, the value of which depends on the expected variability of radioactivity in the effluent stream. Batch effluent releases are sampled prior to purging or venting. Certain radionuclides, referred to as “hard-to-detect” radionuclides (e.g., iron-55, strontium-89, and strontium-90), may be analyzed after the release takes place. “Continuously indicating” radiation monitoring equipment may be used during the release to verify the representativeness of the grab sample or to more fully characterize the release. Table G.1 summarizes the guidance on sampling and analyzing air- borne and liquid waste. The guidance specifies analyses type, minimum sampling frequencies, and lower limits of detection for each type of release. The guidance for pressurized-water reactors in NUREG-1301 are similar, but some of the specified sampling points are different owing to the differ- ent design of these plants. Table G.1 footnotes list the principal radionu- clides that should be measured by the monitoring program.
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362 APPENDIX G TABLE G.1 Radioactive Airborne Waste Sampling and Analysis Program Minimum Type of Lower Limit of Detectiona Sampling Analysis Activity Release Type Frequency Frequency Analysis (µCi/ml) 1 × 10–4 Airborne Offgas Monthly Monthly Principal treatment Grab sample gamma emittersb system 1 × 10–4 Containment Prior to each Prior to each Principal 1 × 10–6 purgec purgec purge or gamma emittersb vent Grab sample Monthly Tritium (oxide) 1 × 10–4 Monthlyc,d Monthlyc Other Principal 1 × 10–6 airborne Grab sample gamma emittersb release points Tritium (oxide) 1 × 10–12 Continuouse Weeklyf All release Iodine-131 types listed Charcoal above sample 1 × 10–11 Continuouse Weeklyf Principal Particulate gamma emittersa sample 1 × 10–11 Continuouse Monthly Gross alpha Composite particulate analysis 1 × 10–11 Continuouse Quarterly Strontium-89 Composite Strontium-90 particulate sample 1 × 10–6 Continuouse Noble gas Noble gases monitor Gross beta or gamma 5 × 10–7 Batch Wasteg Liquid Each batch— Each batch— Principal completed completed prior gamma Release emittersh prior to each to each release Tanks release 1 × 10–6 I-131 1 × 10–5 a. Each batch— At least one per Dissolved completed 31 days and entrained prior to each gases (gamma release; at emitters) least one per 31 days
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363 APPENDIX G TABLE G.1 Continued Minimum Type of Lower Limit of Detectiona Sampling Analysis Activity Release Type Frequency Frequency Analysis (µCi/ml) 1 × 10–5 Compositei—at b. Each batch— H-3 completed least one per 31 prior to each days release 1 × 10–7 Gross alpha 5 × 10–8 c. Each batch— Composite—at Sr-89; Sr-90 completed least one per 92 prior to each days release 1 × 10–6 Fe-55 5 × 10–7 Continuousj Composite—at Principal Continuous least one per 7 gamma days emitters 1 × 10–6 I-131 1 × 10–5 a. Grab At least one per Dissolved sample—at 31 days and entrained least one per gases (gamma 31 days emitters) 1 × 10–5 b. Continuous Composite—at H-3 least one per 31 days 1 × 10–7 Gross alpha 5 × 10–8 c. Continuous Composite—at Sr-89, Sr-90 least one per 92 days 1 × 10–6 Fe-55 aThe LLD is defined, for purposes of these controls, as the smallest concentration of radio- active material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a “real” signal. bIncludes Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases; Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases; other gamma peaks that are identifiable must also be analyzed and reported. cSampling and analysis shall also be performed following shutdown, startup, or a thermal power change exceeding 15 percent of rated thermal power within a 1-hour period. dTritium grab samples shall be taken at least once every 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. continued
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364 APPENDIX G TABLE G.1 Continued eGuidance concerning the sample flow rate. See Table 4.11-2 footnotes in NUREG-1302 for details. fDetailed guidance concerning sampling. See Table 4.11-2 footnotes in NUREG-1302 for details. gA batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling. hThe principal gamma emitters for which the Lower Limit Detection (LLD) control applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 × 10–6. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Control 126.96.36.199 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. iA composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. jA continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. To be representa- tive of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release. SOURCE: NUREG-1302, Table 4.11-2. G.2 EFFLUENT MONITORING AT FUEL-CYCLE FACILITIES Requirements for monitoring effluent releases from front-end nuclear fuel-cycle facilities are contained in the following regulations: • 10 CFR 40.65 (Effluent Monitoring Reporting Requirements) ap- plies to “Part 40” fuel-cycle facilities. These include in situ leaching facilities, milling facilities, and uranium conversion and deconver- sion1 facilities. • 10 CFR 70.59 (Effluent Monitoring Reporting Requirements) ap- plies to “Part 70” fuel-cycle facilities. These include nuclear fuel fabrication plants as well as laser enrichment and centrifuge enrich- ment plants. • 10 CFR 76.35(g) (Contents of an Application) applies to “Part 1 A new uranium deconversion and fluorine extraction processing facility is planned for con- struction near Hobbs, New Mexico. This facility will deconvert depleted uranium hexafluoride tails from the enrichment process into a uranium oxide waste product for eventual disposal and will recover fluorine for commercial resale.
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365 APPENDIX G 76” fuel-cycle facilities. These are the Paducah and Portsmouth Gaseous Diffusion Plants. Because the plants are owned by the U.S. Department of Energy,2 they are subject to the regulations promulgated by the U.S. Environmental Protection Agency in 40 CFR 61 (National Emission Standards for Hazardous Air Pollut- ants), Subpart H (National Emission Standards for Emissions of Radionuclides Other Than Radon from Department of Energy Facilities) and Subpart Q (National Emission Standard for Radon Emissions from Department of Energy Facilities). G.2.1 Milling Facilities Guidance specifically for milling facility effluent monitoring is pro- vided in Regulatory Guide 4.14. This guide recommends that a program of soil, water, air, vegetation, food, and fish sampling and direct radiation monitoring be initiated at least 12 months prior to the construction of the milling facility. The guide also recommends that an operational monitoring program be conducted during construction and after the commencement of milling operations. The recommended operational monitoring program includes the following elements: • Sampling and analysis for natural uranium, thorium-230, radium- 226, and lead-210 particulates from facility stacks. • Sampling and analysis for natural uranium, thorium-230, radium- 226, and lead-210 particulates in air from three locations at or near the site boundaries in sectors that are expected to have the highest concentrations of airborne particulates; from one or more locations at the closest residence(s) or occupy-able structure(s); and from one control location. • Sampling and analysis for radon gas at five or more locations that were used for air particulate sampling. • Measurement of direct radiation at five or more locations that were used for air particulate sampling. G.2.2 Other Fuel-Cycle Facilities Guidance for monitoring programs at other front-end facilities (e.g., conversion, enrichment, fuel fabrication) is provided in Regulatory Guide 4.16. This guide recommends that licensees: • Establish a sampling program that is sufficient to determine quanti- 2 These U.S. government-owned plants are leased to USEC, a private corporation.
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366 APPENDIX G ties and average concentrations of radioactive material discharges from the facility. • Use continuous monitoring methods for determining releases of airborne effluents from process systems that have particulate or airborne materials that can be easily dispersed. • Use grab-sampling methods to confirm releases at points that are continuously monitored. Guidance for uranium recovery monitoring programs can be found in Table 2 of Regulatory Guide 4.14. This guide recommends that licensees perform: • Soil sampling and analysis at five or more locations that were used for air particulate sampling. • Surface water and groundwater sampling and analysis. • Periodic fish, food, and vegetation sampling and analysis, if available. • Sediment sampling and analysis. Requirements for conducting an effluent monitoring program at the U.S. Department of Energy-owned gaseous diffusion plants are provided in 40 CFR 61, Subpart H. This subpart requires radionuclide emission mea- surements to be made at all release points that have a potential to discharge radionuclides into the air in quantities that could cause an effective dose equivalent in excess of 0.1 mrem per year to any member of the public. Confirmatory measurements are required for other release points that have a potential to release radionuclides into the air. The subpart also contains specific requirements for measurement and analysis procedures using ap- proved methods and for quality assurance. REFERENCES USNRC (U.S. Nuclear Regulatory Commission) (1977a). Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors. Revision 1. USNRC (1977b). Regulatory Guide 1.113. Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I. USNRC (1977c). Regulatory Guide 1.109. Calculation of Doses to Man from Routine Re- leases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I. Revision 1. October 1977.