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OCR for page 359
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 6.9.1.4 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.
