U.S. NUCLEAR REGULATORY COMMISSION October 2008 

OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1

DRAFT REGULATORY GUIDE

Contact: S. Garry 

(301) 415-2766 

DRAFT REGULATORY GUIDE DG-1186

(Proposed Revision 2 of Regulatory Guide 1.21) 

MEASURING, EVALUATING, AND REPORTING RADIOACTIVE 

MATERIAL IN LIQUID AND GASEOUS EFFLUENTS AND 

SOLID WASTE 

A. INTRODUCTION

This guide describes a method the staff of the U.S. Nuclear Regulatory Commission (NRC) considers 

acceptable for use in measuring, evaluating, and reporting radioactivity in effluents and solid radioactive 

waste shipments. The regulatory guide also provides guidance on determining and reporting the public 

dose from nuclear power plant operations. The regulatory basis for the radiological effluent control 

program is established in: 

1. Title 10, Section 20.1501, “Surveys,” of the Code of Federal Regulations (10 CFR 20.1501) 

(Ref. 1); 

2. 10 CFR 50.36a, “Technical Specifications on Effluents from Nuclear Power Reactors” (Ref. 

2); 

3. 10 CFR 20.1302, “Compliance with Dose Limits for Individual Members of the Public,” and 

4. 10 CFR 72.44(d), “License Conditions.” 

10 CFR 20.1501 requires surveys that may be necessary and are reasonable to evaluate the magnitude and 

extent of potential radiological hazards. 10 CFR 20 defines “survey” as an evaluation of the radiological 

conditions and potential hazards related to radioactive material or other sources of radiation, including 

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory 

position in this area. It has not received final staff review or approval and does not represent an official NRC final staff position. 

Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory 

analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may 

be submitted to the Rulemaking, Directives, and Editing Branch, Office of Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555-0001; emailed to nrcrep.resource@nrc.gov; submitted through the NRC’s interactive 

rulemaking Web page at http://www.nrc.gov; faxed to (301) 415-5144; or hand-delivered to Rulemaking, Directives, and Editing 

Branch, Office of Administration, US NRC, 11555 Rockville Pike, Rockville, MD 20852, between 7:30 a.m. and 4:15 p.m. on 

Federal workdays. Copies of comments received may be examined at the NRC’s Public Document Room, 11555 Rockville Pike, 

Rockville, MD. Comments will be most helpful if received by December 30, 2008. 

Electronic copies of this draft regulatory guide are available through the NRC’s interactive rulemaking Web page (see above); the 

NRC’s public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the NRC’s Electronic 

Reading Room at http://www.nrc.gov/reading-rm/doc-collections/; and the NRC’s Agencywide Documents Access and 

Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML080660617. 
DG-1186, Page 2

(1) a physical survey of the location of radioactive material and (2) measurements or calculations or levels 

of radiation or concentrations or quantities of radioactive material present. 10 CFR 50.36a requires the use 

of plant technical specifications and operating procedures to keep radioactive effluents as low as 

reasonably achievable (ALARA) per the numerical design guidance of 10 CFR 50 Appendix I. 10 CFR 

50.36a also requires reporting (1) the quantity of each of the principal radionuclides released to 

unrestricted areas in liquid and gaseous effluents and (2) other information used to estimate the maximum 

potential annual radiation doses to the public from radioactive effluents. 10 CFR 20.1302, establishes 

requirements for surveys in the unrestricted and controlled areas and for radioactive materials in effluents 

released to unrestricted and controlled areas. The purpose of these surveys is to demonstrate compliance 

with (1) the (TEDE) dose limits of 10 CFR 20.1301 (e.g., 100 mrem/year) for individual members the 

public. Although 10 CFR 20.1302(b)(2) provides a second method of demonstrating compliance with 

dose limits for individual members of the public, nuclear power plant Technical Specifications essentially 

require use of method 10 CFR 20.1302 (b)(1) to determine the TEDE to the individual likely to receive the 

highest dose. This requirement is based on actual, credible exposure pathways to a real individual that has 

been located during a land use census. (See also Regulatory guide 1.109 and Attachment 6 of SECY-03-

0069) In addition, 10 CFR 72.44(d) establishes requirements for the independent spent fuel storage 

installation (ISFSI), high level nuclear waste, and reactor-related greater than class “C” waste. This 

regulatory guide describes methods for implementing these requirements for all fuel cycle activities (i.e., 

fuel storage and plant operation) at a commercial nuclear power plant. 

The reports required under (1) Subpart M, “Reports,” of 10 CFR Part 20, “Standards for Protection 

Against Radiation,” (2) 10 CFR 50.72, “Immediate Notification Requirements for Operating Power 

Reactors,” and (3) 10 CFR 50.73, “Licensee Event Report System,” or other licensee requirements must be 

made in accordance with the applicable regulations. In addition, effluent discharges and radioactive 

material losses reported under those regulatory provisions should also be reported in the ARERR described 

in this regulatory guide.

The NRC issues regulatory guides to describe methods that the staff considers acceptable for use in 

implementing specific parts of the agency’s regulations, to explain techniques that the staff uses in 

evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory 

guides are not substitutes for regulations, and compliance with them is not required. The methods and 

practices outlined in regulatory guides are one acceptable method for implementing the regulations. 

Licensees may use other methods, but in those cases, additional written justifications or evaluations may be 

necessary to demonstrate compliance with the regulations. Because of this, the use of terms such as 

“shall,” “must,” and “should” in this regulatory guide should be viewed in the context of emphasis rather 

than regulatory compliance. 

This regulatory guide contains information collections covered by 10 CFR Part 50 that the Office of 

Management and Budget (OMB) approved under OMB control number 3150-0011. The NRC may 

neither conduct nor sponsor, and a person is not required to respond to, an information collection request 

or requirement unless the requesting document displays a currently valid OMB control number. 
DG-1186, Page 3

B. DISCUSSION

1. Regulatory Guidance 

Five basic documents contain the regulatory guidance for implementing the 10 CFR Part 20 regulatory 

requirements and plant technical specifications related to monitoring and reporting of radioactive material 

in effluents and environmental media, solid radioactive waste disposal, and resultant public dose: 

(1) Regulatory Guide 1.21, “Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and 

Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled 

Nuclear Power Plants” (Ref. 3) 

(2) Regulatory Guide 4.1, “Programs for Monitoring Radioactivity in the Environs of Nuclear Power 

Plants” (Ref. 4) 

(3) Regulatory Guide 4.15, “Quality Assurance for Radiological Monitoring Programs (Inception 

Through Normal Operations to License Termination)—Effluent Streams and the Environment,” 

(Ref. 5) 

(4) NUREG-1301, “Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 

Controls for Pressurized Water Reactors,” (Ref. 6) 

(5) NUREG-1302, “Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 

Controls for Boiling Water Reactors,” (Ref. 7) 

These five documents, when used in an integrated manner, provide the basic principles and 

implementation details for developing and maintaining effluent and environmental monitoring programs at 

nuclear power plants. The three regulatory guides specify the principles of radiological monitoring, and 

the two NUREGs provide the specific implementation guidance for baseline monitoring programs. 

Regulatory Guide 1.21 addresses the measuring, evaluating, and reporting of effluent releases, solid 

radioactive waste, and public dose from nuclear power plants. It describes the important concepts in 

planning and implementing an effluent and solid radioactive waste program. Concepts covered include 

meteorology, release points, monitoring methods, identification of principal radionuclides, unrestricted 

area boundaries, continuous and batch release methods, representative sampling, composite sampling, 

radioactivity measurements, decay corrections, quality assurance (QA), solid radioactive waste shipments, 

and public dose assessments. 

Regulatory Guide 4.1 addresses the environmental monitoring program. It discusses principles and 

concepts important to environmental monitoring at nuclear power plants. The regulatory guide addresses 

the need for preoperational and background characterization of radioactivity. It also addresses onsite and 

offsite monitoring, including the principal exposure pathways and the significant exposure pathways. The 

guide defines the principal exposure pathways, the program scope of sampling media and sampling 

frequency, and the methods of comparing environmental measurements to effluent releases in the annual 

environmental report. 

Regulatory Guide 4.15 provides the basic principles of QA in all types of radiological monitoring 

programs. It does not specifically address nuclear power plants but covers all types of licenses and 

licensees. It provides the principles for structuring organizational lines of communication and 

responsibility, using qualified personnel, implementing standard operating procedures, defining data 

quality objectives, performing quality control (QC) checking for sampling and analysis, auditing the 

process, and taking corrective actions. 
DG-1186, Page 4

NUREG-1301 and NUREG-1302 provide the detailed implementation guidance by describing baseline 

effluent and environmental monitoring programs. The NUREGs specify effluent monitoring and 

environmental sampling requirements, surveillance requirements for effluent monitors, types of monitors 

and samplers, sampling and analysis frequencies, types of analysis and radionuclides analyzed, lower limits 

of detection (LLDs), specific environmental media to be sampled, and reporting and program evaluation 

and revision. 

2. Objectives of the Radiological Effluent Control Program 

The requirements for the radiological effluent control program (RECP) are in 10 CFR Part 20 and 10 CFR 

Part 50, Appendix I, “Numerical Guides for Design Objectives and Limiting Conditions for Operation to 

Meet the Criterion ‘As Low as Is Reasonably Achievable’ for Radioactive Material in Light-Water-Cooled 

Nuclear Power Reactor Effluents.” Specific requirements described in a facility’s technical specifications 

(TS), in conjunction with the regulatory positions provided in this guide, can be used as a basis for 

establishing the radiological effluent control program. The radiological effluent control program for a 

nuclear power plant has the following six basic objectives: 

(1) Ensure that effluent instrumentation has the functional capability to measure and evaluate effluent 

releases 

(2) Ensure that effluent treatment systems are used to reduce effluent releases to as low as is 

reasonably achievable (ALARA) levels 

(3) Establish instantaneous release rate limitations on the concentrations of radioactive material 

(4) Limit the annual and quarterly doses or dose commitment to members of the public in liquid and 

gaseous effluents to unrestricted areas 

(5) Measure, evaluate, and report the quantities of radioactivity in gaseous effluents, liquid effluents, 

and solid radioactive waste 

(6) Evaluate the dose to members of the public 

The radioactive effluent release report, submitted annually prior to May 1 (unless a licensing basis exists 

for a different submittal date), in conjunction with the annual radiological environmental operating report 

submitted annually by May 15, are used to demonstrate compliance with TS for the radioactive effluent 

control program. The reports demonstrate the following: 

(1) Effectiveness of effluent controls and measurement of the environmental impact of radioactive 

materials 

(2) Compliance with the design objectives and limiting conditions for operation required to meet the 

ALARA criteria in Appendix I to 10 CFR Part 50 

(3) Relationship between quantities of radioactive material released in effluents and resultant radiation 

dose to individuals from principal pathways of exposure 

(4) Compliance with the radiation dose limits to members of the public established by the NRC and 

the U.S. Environmental Protection Agency (EPA) 

(5) Compliance with the effluent reporting requirements of 10 CFR 36a 

Licensees may also – if they so choose – use the format specified in this regulatory guide for 10 CFR 

72.44(d) ISFSI effluent reports. However, the ISFSI effluent reporting requirement of 10 CFR 72.44(d) is 

not normally satisfied by inclusion as part of the ARERR since the reporting dates may conflict. If the 

dates are coincident, or can be met with a single report, the 10 CFR 72.44(d) reporting requirements may 

be fulfilled using the ARERR. 
DG-1186, Page 5

C. REGULATORY POSITION

1. Effluent Monitoring 

Guidance for Effluent Monitoring – Monitoring programs should be established to identify and quantify 

specific radionuclides in effluents. NUREG-1301 (for pressurized-water reactors (PWRs)) and NUREG1302 (for boiling-water reactors (BWRs)) specify the generic controls and surveillance requirements, 

including the frequency, duration, and methods of measurement. Site-specific requirements may differ 

from the generic NUREG-1301/1302 guidance provided there is either a documented evaluation or 

justification for such deviations as part of an ODCM authorized change, or if submitted as part of the 

original offsite dose calculation manual (ODCM) in accordance with Generic Letter 89-01 and approved 

by the NRC. These NUREGs provide specifications for LLDs, requirements for batch releases and 

continuous releases, sampling frequencies, analysis frequencies and timelines, and composite sample 

requirements. 

Release Points for Effluent Monitoring – The ODCM should identify the facility’s release points, that 

are used to quantify the liquid and gaseous effluent releases to the onsite and offsite environs. For those 

release points containing contributions from two or more inputs (or systems), it is preferable to monitor 

each major input (or system) individually to avoid dilution effects which may impede or prevent 

radionuclide identification. NUREG-1301/1302 contains detailed guidance for the content and format of a 

licensee’s ODCM. For purposes of effluent monitoring the ODCM should contain and/or describe the 

following: 

(a) Significant release points (see definition in glossary) (e.g., stacks, vents, liquid radioactive waste 

discharge points) 

(b) Less significant or intermittent release points (see definition in glossary) (e.g., main steam-line 

safety valves, steam-driven feed water pumps, turbine building sources, storm drains, evaporative 

releases, leachate seepage and/or evaporative releases from ponds in the restricted or controlled 

areas before discharge under 10 CFR 20.2001, “General Requirements”) 

(c) Site environs map showing the following: 

i. significant release points 

ii. less significant or intermittent release points 

iii. boundaries of the restricted area and the controlled area (per 10 CFR Part 20 definitions) 

iv. boundary of the unrestricted area for liquid effluents (e.g., at the end of the pipe or 

entrance to a public waterway) 

v. boundary of the unrestricted area for gaseous effluents (e.g., the site boundary) 

(d) Dose calculation methodologies for all significant exposure pathways that contribute in excess of 

10% of the public dose from all effluent pathways combined. 

The unrestricted area may be defined differently (and separately) for each of the following: (1) liquid 

effluents, (2) gaseous effluents, and (3) if appropriate, for other radiological controls such as direct 

radiation. 
DG-1186, Page 6

Monitoring a Significant Release Point – A release point is considered a significant release point if 

either (1) the dose is greater than or equal to 1% of the total from gaseous release points or liquid release 

points (considered separately), or (2) the activity (e.g., curies) is greater than or equal to 1% of the total 

effluent activity from gaseous release points or liquid release points (considered separately). Additionally, 

nuclides are totaled separately by the categories as listed in regulatory position 9.1.1 (e.g., fission and 

activation gases, iodines/halogens, particulates, tritium, and gross alpha) and 9.2.1 (fission and activation 

products, tritium, dissolved and entrained noble gases, and gross alpha). Lastly, the summing intervals for 

dose and activity calculations are either quarterly, annually, each fuel cycle, or longer period as specified 

by the licensee. 

Significant release points should be listed in the ODCM. Releases from a significant release point should 

be assessed based on an appropriate combination of actual sample analysis results, radiation monitor 

responses, flow rate indications, tank level indications, and system pressure indications as necessary to 

ensure the amount of radioactive material released and the corresponding doses are not substantially 

underestimated (see 10 CFR 50, Appendix I, Section III, “Implementation”). If activity is detected when 

monitoring a significant release point, it must be reported in the effluent totals (i.e., Table A-1 or Table A2) in the ARERR (provided the amount released is significant to the 3 significant figures required for the 

ARERR). 

Monitoring a Less-significant Release Point – A release point is considered a less-significant release 

point if both (1) the dose is less than 1% of the total from gaseous release points or liquid release points 

(considered separately), and (2) the activity (e.g., curies) is less than 1% of the total effluent activity from 

gaseous release points or liquid release points (considered separately). Additionally, nuclides are totaled 

separately by the categories as listed in regulatory position 9.1.1 (e.g., fission and activation gases, 

iodines/halogens, particulates, tritium, and gross alpha) and 9.2.1 (fission and activation products, tritium, 

dissolved and entrained noble gases, and gross alpha). Lastly, the summing intervals for dose and activity 

calculations are either quarterly, annually, each fuel cycle, or longer period as specified by the licensee. 

Less-significant release points are, to the extent reasonable, required to be listed in the ODCM. Releases 

from a less-significant or intermittent release point may be assessed to the extent reasonable using 

assumptions and bounding calculations (in lieu of, or in addition to, sampling and analysis). Sampling and 

analysis should occur, however, if changing plant conditions may reasonably affect the status of a lesssignificant release point (e.g., significant change is primary-to-secondary leakage in PWRs or substantial 

cross-contamination between systems). The guidance in this regulatory guide regarding monitoring lesssignificant release points for purposes of accountability (via the ARERR) does not replace, supersede, or 

otherwise modify any responsibility for monitoring systems, normally not contaminated, as outlined in 

NRC IE Bulletin 80-10 (Ref. 36). If monitoring a less-significant release point using actual samples (e.g., 

grab samples or composite sampling), principal radionuclides should be monitored in accordance with the 

predetermined and appropriate analytical sensitivity (e.g., LLD). If activity is detected when monitoring a 

less-significant release point, it must be reported in the effluent totals (i.e., Table A-1 or Table A-2) in the 

ARERR (provided the amount released is significant to the 3 significant figures required for the ARERR). 

Monitoring Leaks and Spills -- Areas where leaks or spills have occurred should be identified as an 

“impacted area” for decommissioning purposes in accordance with NUREG-1757, “Consolidated 

Decommissioning Guidance,” issued September 2006 (Ref. 10). A leak or spill should be assessed to 

estimate the parameters listed in regulatory position 9.5 (Supplemental Information). Sampling and 

analyses of the source of the leak (undiluted effluent) and sampling of soil and/or contaminated ground 

water should be performed if – and as soon as – practical. In some instances sampling, particularly soil 

sampling, may not be practical if the leak occurred in inaccessible areas, or if there are extenuating 

considerations. In this respect, groundwater monitoring may be used as a surrogate for soil sampling. If 
DG-1186, Page 7

sampling is not practical, this information (about the inaccessible areas or safety considerations) should be 

included in the 10 CFR 50.75(g) records. The leak/spill location should be estimated to identify the extent 

of the impacted area and predicted size or extent and movement of the contaminated plume. If a spill is 

promptly remediated (e.g., within 48 hours), and if surveys of the remediated area indicate no detectable 

residual contamination remaining in the soil or groundwater, and the excavated material is stored, disposed 

(or otherwise dispositioned) as solid radioactive waste, then – by definition – there was no liquid release to 

the unrestricted area, and the spill need not be reported in the ARERR for purposes of liquid release 

accountability. However, the 10 CFR 50.75(g) decommissioning file should be updated to include the 

appropriate information. Additionally, the licensee may choose to report any or all portions of the 

decommissioning file (e.g., “spills or leaks”) in the ARERR for purposes of transparency of disclosure 

regarding decommissioning records. 

Prompt response and timely actions should be taken to the extent reasonable, including isolating the leak or 

spill at the source, preventing the spread of the leak or spill, and remediation of the leak or spill. The dose 

to members of the public from the leak or spill should be evaluated using realistic exposure scenarios (see 

Attachment 6 to SECY-03-0069, “Results of the License Termination Rule Analysis,” dated May 2, 2003, 

(Ref. 11) for more information). Local authorities should be notified of the leak or spill when the leak or 

spill exceeds the threshold established with local authorities. When notifying local authorities, licensees 

should ensure effective communication using the guidance provided in NUREG/BR-0308, “Effective Risk 

Communication,” issued January 2004 (Ref. 12) especially with respect to ensuring the risk is described in 

the appropriate context. In addition, an evaluation should be made as to whether to notify the NRC of the 

event in accordance with 10 CFR 50.72 and NUREG-1022 (Ref. 35). In general, notifications to the NRC 

should be made for large leaks and spills or when significant public concern is raised. 

The licensee should use its problem identification and resolution program (corrective action program) in 

the evaluation of the event, and appropriate documentation should be placed in or cross-referenced to the 

10 CFR 50.75(g) files. Remediation should be evaluated and implemented as appropriate, based on factors 

such as the location and accessibility of the contamination, the concentrations of and extent of the 

contamination, rates of monitored natural attenuation, and the decommissioning criteria described in 

NUREG-1757. 

Monitoring Continuous Releases -- For continuous releases, gross radioactivity measurements are often 

the only practical means of continuous monitoring. These gross radioactivity measurements are typically 

used to actuate alarms and terminate (trip) effluent releases, but by themselves, they are generally not 

acceptable for demonstrating compliance with effluent release limits. Grab samples should be collected at 

scheduled frequencies (see NUREG-1301 and NUREG-1302 or as approved in GL 89-01 submittals) to 

quantify specific radionuclide concentrations and release rates. The frequency of sample collection and 

radionuclide analyses should be based on the degree of variance in (1) the magnitude of the release and (2) 

the relative radionuclide composition from an established norm. Where the magnitude of the release and 

the relative nuclide composition of a continuous release vary significantly over the course of the discharge 

period, a combination of grab samples and continuous monitor readings can assist in accurately estimating 

the release. Continuous monitoring data (e.g., chart recorder data), as well as grab sample data, should be 

reviewed periodically and used to identify this variance from the established norm. Periodic evaluations 

should be made between gross radioactivity measurements and grab sample analyses of specific 

radionuclides. These evaluations should be used to establish the conversion factors between monitor 

readings and concentrations of radionuclides when determining monitor alarm setpoints and effluent trip 

setpoints. 

Monitoring Batch Releases -- For batch releases, measurements should be performed to identify specific 

radionuclides before a release to the environment. In those cases in which an analysis of specific 
DG-1186, Page 8

radionuclides (such as strontium-89/90 and iron-55) cannot be done before release (see NUREG-1301 and 

NUREG-1302), representative samples should be collected for the purpose of subsequent composite 

analysis. The composite samples should be analyzed at the scheduled frequencies specified in NUREG1301 and NUREG-1302. 

Principal Radionuclides for Effluent Monitoring – NUREG 1301/1302 provides a list of “principal 

gamma emitters” for which the LLD control applies. In this revised regulatory guide, the term “principal 

radionuclide” is introduced from a risk-informed perspective. In this context, the term “principal 

radionuclide” has special significance with respect to the (1) monitoring methods and the (2) required 

sensitivity levels (e.g., LLDs) for an analytical method. From a risk-informed perspective, a list of 

“principal radionuclides” may be evaluated by the licensee and identified for use at a particular site. 

Although the LLD list from NUREG-1301/1302 may be used as a starting point for any determination of 

“principal radionuclides,” the list from NUREG-1301/1302 is not a required list of principal radionuclides. 

 Instead, the “principal radionuclides” at a site will be dependent on site-specific factors (e.g., amount of 

failed fuel, extent of system leakage, sophistication of radwaste processing equipment, and level of 

expertise in operating radwaste processing systems). As a result, the “principal radionuclides” may vary 

from site to site depending on the list contained in the licensee’s ODCM. 

One method of determining the “principal radionuclides” is to classify all nuclides as “principal 

radionuclides,” and apply the LLDs directly from NUREG-1301/1302. This method is simple to 

implement, but in certain cases it may entail inappropriately long count times or alternate methods of 

analysis for low-activity nuclides with no—or extremely low—dose significance. 

Alternatively, a risk-informed approach may be used based on an extension of the guidance provided in 

Revision 1 of Regulatory Guide 1.109 (Ref. 23), section C (see discussion of “significant exposure 

pathways”). In a risk-informed approach, the “principal radionuclides” may be determined based on their 

relative contribution to (1) the public dose or (2) the amount of activity discharged. Under this concept, 

radionuclides that have either a significant activity or a significant dose contribution should be monitored 

in accordance with a predetermined and appropriate analytical sensitivity level (LLD). In the context of 

required sensitivity levels for an analytical method (i.e., LLD), a radionuclide is a “principal radionuclide” 

if it contributes more than 1% of the public dose or constitutes more than 1% of the total activity [in a 

particular reporting interval, and for a particular radionuclide category listed in Appendix A, Table A-1 

(for gas releases) or Table A-2 (for liquid releases)]. 

If a risk-informed approach is used, “principal radionuclides” should be determined based on an evaluation 

over a time period that includes a refueling outage (e.g., one fuel cycle). A periodic reevaluation should be 

performed to determine whether the radionuclide mix has changed and/or to identify new “principal 

radionuclides.” If a risk-informed approach is applied to the determination of “principal radionuclides,” 

the ODCM specifies the list of “principal radionuclides.” The list in NUREG-1301/1302 is not a 

regulatory requirement; however, the list in the ODCM is the requirement. Licensees are allowed to revise 

the ODCM in accordance with the ODCM change process as described in the plant’s technical 

specifications (which includes documented evaluations of such changes). Once a nuclide is identified as a 

“principal radionuclide,” its release from “significant release points” must be monitored in accordance with 

the LLDs specified in the ODCM. 

For radionuclides that are not “principal radionuclides,” licensee discretion may be applied to (1) the 

sensitivity of analysis and (2) the monitoring methods for those nuclides. If analytical sensitivities or 

monitoring methods (e.g., sample frequencies, analysis frequencies) are chosen which are different from 

NUREG-1301/1302, the basis for the deviations should be documented. For example, data quality 
DG-1186, Page 9

objectives (DQOs) and other concepts from Regulatory Guide 4.15, Revision 2, (Ref. #5) may be useful 

for determining risk-informed sensitivity levels for an analytical method. 

However, the concept of “principal radionuclides” does not reduce the burden for reporting nuclides 

detected in effluents. In addition to “principal radionuclides,” other radionuclides detected during routine 

monitoring of release points must be reported in the radioactive effluent release report and included in dose 

assessments to members of the public. 

The production (and the release) of Carbon-14 from commercial nuclear power plants in the United States 

is relatively constant and is typically proportional to the power rating of the reactor. The quantity of 

carbon-14 discharged can be determined by measurement or by use of a normalized C-14 source term and 

scaling factors based on power generation; see National Council on Radiation Protection and 

Measurements Report No. 81,”Carbon-14 in the Environment” (Ref. 9). Because the production of C-14 

is expected to be relatively constant, if sampling is performed for C-14, the sampling frequency may be 

adjusted to that interval that allows adequate reporting of effluents.

2. Effluent Sampling 

Representative Sampling – The baseline schedule for sample collection and analyses appears in NUREG1301 and NUREG-1302 or as approved by the NRC and put in the original ODCM approvals submitted as 

part of Generic Letter 89-01. Additional samples should be obtained as needed to characterize abnormal 

releases or significant operational evolutions. Samples should be representative of the overall effluent in 

the bulk stream, collection tank, or container. Representative samples should be obtained from well-mixed 

streams or volumes of effluent at sampling points by using proper equipment and sampling procedures. 

Sampling Liquid Radwaste – Before sampling, large volumes of liquid waste should be mixed to ensure 

that sediments or particulate solids are distributed uniformly in the waste mixture. For example, a large 

tank may be mixed using a sparger system or re-circulated by up to three volumes to ensure that a 

representative sample can be obtained, as recommended by American Society for Testing and Materials 

(ASTM) D 3370-07, “Standard Practices for Sampling Water from Closed Conduits” (Ref. 13). If tank 

mixing practices deviate from industry standards (i.e., re-circulation or otherwise), a technical evaluation or 

other justification should be provided. Sample points should be located where there is a minimum of 

disturbance of flow caused by fittings and other physical characteristics of the equipment and components. 

 Sample nozzles should be inserted into the flow or liquid volume to ensure sampling of the bulk volume 

of pipes and tanks. Sample lines should be flushed for a sufficient period of time before sample extraction 

to remove sediment deposits and air and gas pockets. Generally, three line volumes should be purged (see 

ASTM D 3370-07) before withdrawing a sample, unless a technical evaluation or other justification is 

provided. Periodically, a series of samples should be taken during the interval of discharge to determine 

whether any differences exist as a function of time and to ensure individual samples are indeed 

representative of the effluent mixture. In some instances, this may be accomplished by collecting one or 

more samples (either by “grab” or composite sampler) during the discharge and comparing with one or 

more samples taken prior to the discharge. If a series of samples are collected, these samples can be used 

to assess the amount of measurement uncertainty in obtaining representative samples. 

Sampling Gaseous Radwaste – Regulatory Guide 4.15, American National Standards Institute (ANSI) 

N42.18-2004, “Specification and Performance of On-Site Instrumentation for Continuously Monitoring 

Radioactivity in Effluents” (Ref. 14), and ANSI/Health Physics Society (HPS) N13.1-1999, “Sampling 

and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear 

Facilities” (Ref. 15), contain the general principles for designing and conducting monitoring programs for 

airborne effluents, including obtaining valid samples of airborne radioactive material in effluents and the 

guidelines for sampling from ducts and stacks. In addition, evaluations should be performed to validate 
DG-1186, Page 10

representative samples (e.g., evaluate the potential for inaccurate sampling of a gaseous effluent stream 

that may bypass a particulate filter and collect on an iodine collection cartridge), as well as inaccurate 

sample analyses configurations or counting geometries. 

Sampling Bias – Sampling and storage techniques that could bias quantitative results for effluent 

measurements should be evaluated and corrections applied as necessary. These biases include inaccurate 

measurement of sample volumes resulting from pressure drops in long sample lines and loss of particulates 

or iodine in sample lines resulting from deposition or plate-out. Samplers for gaseous waste should be 

evaluated for particulate deposition using ANSI N13.1-1999 (Ref. 15). 

Composite Sampling – Composite samples should be representative of the average quantities and 

concentrations of radioactive materials released in liquid and gaseous effluents. Composite samples should 

be collected in proportion to the rate of flow of the effluent stream or in proportion to the volume of each 

batch of effluent releases. 

Sample Preparation and Preservation – Methods of sample preparation and/or sample storage should 

minimize the potential for loss of radioactive material (i.e., deposition of analyte on walls of the sample 

container or volatilization of analyte). Composite sample storage time should be as short as practical to 

preclude deposition on the storage container, or sample stabilization should be considered. Before 

quantitative radionuclide analyses for liquid effluent composites, samples should be mixed thoroughly so 

that the sample is representative of the released effluent. 

Procedures should be instituted for handling, packaging, and storing samples to ensure that losses of 

radioactive materials or other factors causing sample deterioration do not invalidate the analysis. For 

example, filters should be stored carefully so as to prevent loss of radioactive material from the filter paper. 

Short-lived Nuclides and Decay Corrections – In the analysis of short-lived radionuclides, 

measurements should be made as soon as practical after collection to minimize loss by radioactive decay 

(e.g., short-lived noble gases). In other cases, when needed to improve the detection of the longer lived 

nuclides, time should be allowed for the decay of short-lived, interfering nuclides. 

Radiological decay corrections should be made as though the effluent were released uniformly throughout 

the sampling period (unless it is shown that the majority of the effluent was released during a particularly 

short interval). The time intervals between sample collection and analysis should be recorded. To estimate 

radioactive decay during composite sample collection, the midpoint of the sampling period may be used as 

the sample time. If the midpoint of the compositing interval is chosen as the sample time, those nuclides 

whose half-lives are short relative to the compositing period may be underestimated (if those nuclides 

appeared only in the early part of the compositing period) or overestimated (if those nuclides appeared in 

the last part of the compositing period). In order to avoid such bias, an alternate method of determining the 

effective sample time may be employed for composite samples. One such alternate approach involves 

application of weighting factors to compensate for (1) the decay time of each individual composited 

portion, and (2) the quantity of each individual composited portion in relation to the total quantity of the 

final composite sample. Another alternate approach would be to decrease the composite sampling interval. 

3. Effluent Dispersion (Meteorology and Hydrology) 

Meteorological Data – Gaseous effluents discharged into the atmosphere are transported and diluted as a 

function of (1) the atmospheric conditions in the local environment, (2) the topography of the region, and 

(3) the characteristics of the effluents. Site-specific meteorological data should be measured in accordance 

with Regulatory Guide 1.23, “Meteorological Monitoring Programs for Nuclear Power Plants” (Ref. 16). 

The meteorological data does not need to be reported in the ARERR, but the data must be summarized and 
DG-1186, Page 11

maintained for inspection. An annual meteorological summary report that provides the joint frequency 

distributions of wind direction and wind speed by atmospheric stability class (see Regulatory Guide 1.23) 

should be prepared and maintained on-site for the life of the plant. In addition, hourly meteorological data 

should be recorded, and available, if needed for assessing abnormal gaseous releases. 

Atmospheric Transport and Diffusion – Site specific meteorological data collected should be analyzed 

and used to generate gaseous effluent dispersion factor(s) χ/Q and deposition factor(s) D/Q in accordance 

with Regulatory Guide 1.111, “Methods for Estimating Atmospheric Transport and Dispersion of Gaseous 

Effluents in Routine Releases from Light-Water-Cooled Reactors” (Ref. 17). The use of annual average 

meteorological conditions (χ/Q and D/Q) is appropriate for continuous releases. This practice may also be 

acceptable for calculating doses from intermittent releases if the releases occur randomly and with 

sufficient frequency to justify the use of annual average meteorological conditions (see Regulatory Guide 

1.111). When calculating long-term, annual average frequency distributions, the use of 5 (or more) years 

of data should be used. If long-term, annual average χ/Q and D/Q values are used, the values must be 

revalidated or updated periodically (e.g., every 3 to 5 years). 

Release Height – The release height affects the transport and dispersion of radioactive materials especially 

with respect to “downwash” and building wake effects. For facilities with both ground-level and elevated 

releases, an evaluation should be made to determine the proper location of the highest exposed individual. 

From a dispersion perspective, when determining the highest exposure location (submersion and/or 

deposition), the evaluation should consider the magnitude of release originating as an elevated release and 

the magnitude of release originating as a ground-level release. For example, a close-in, downwind location 

in one sector may have a higher χ/Q (i.e., less dispersion) for a ground-level release; however, the majority 

of the source term may be originating as an elevated release, causing a higher concentration (χ) at a more 

distant location, possibly in a different sector. 

Aquatic Dispersion (Surface Waters) – Liquid radioactive effluent may be discharged in accordance 

with 10 CFR 20.2001 into a variety of receiving surface water bodies, including non-tidal rivers, lakes, 

reservoirs, cooling ponds, estuaries, and open coastal waters. This effluent is dispersed by various 

mechanisms (i.e., turbulent mixing, stream flow in the water bodies, and internal circulation or flowthrough in lakes, reservoirs, and cooling ponds). Parameters influencing the dispersion patterns and 

concentrations near a site include the direction and speed of flow of currents, both natural and plantinduced, in the receiving water; the intensity of turbulent mixing; the size, geometry, and bottom 

topography of the receiving water; the location of effluent discharge in relation to the receiving water 

surface and shoreline; the amount of recirculation of previously discharged effluent; the characteristics of 

suspended and bottom sediments; and sediment sorption properties. Regulatory Guide 1.113, “Estimating 

Aquatic Dispersion of Effluents from Accidental and Routine Releases for the Purpose of Implementing 

Appendix I” (Ref. 18), describes calculational models for estimating aquatic dispersion to surface water 

bodies. However, the dispersion characteristics may be highly site dependent, and local characteristics 

should be considered when performing dispersion modeling and dose assessments.

Spills and Leaks to the Ground Surface – Liquid effluents discharged onto the land surface are 

transported and diluted as a function of hydrologic processes and properties of the effluent. The effluents 

may temporarily accumulate, pool, and/or run off to natural and/or engineered drainage systems. During 

this process, water may also be absorbed into the soil (addressed in the next paragraph). Regulatory Guide 

1.113 discusses the use of simple models to estimate transport of effluents through surface water bodies 

and considers water usage effects. Spills or leaks of radioactive material to the ground surface should 

initiate some spill or leak run-off characterization. These characterization activities should – at a minimum 

– satisfy (1) the requirements of 10 CFR 50.75(g) as well as (2) the effluent reporting requirements of 

NUREG-1301/1302 typically associated with planned effluents (e.g., sampling prior to discharge to 
DG-1186, Page 12

unrestricted areas). Refer to sections 9.5.1, 9.5.2, and 9.5.9 for guidance on the general format for 

reporting abnormal releases and effluents to unrestricted areas. 

Spills and Leaks to the Ground Water – Liquid radioactive leaks and spills may be inadvertently 

discharged to onsite (or offsite) ground water. In addition to surface water run-off (as described in the 

above paragraph), leaks and spills onto the ground surface can be absorbed into the soil. Once in the soil 

some of the material in the leak or spill may, depending on the local-soil properties and associated liquid 

flux of the release, eventually reach the water table. The dispersion of this material depends on the local 

subsurface geology and hydrogeology characteristics. Liquid releases into the subsurface will be 

transported as a function of ground water flow processes and conditions (e.g., hydraulic gradients, 

permeability, porosity, and geochemical processes). 

A ground water site conceptual model should be developed to predict the subsurface water flow parameters 

to include direction and rate and to be used as the basis for estimating the dispersion of inadvertent liquid 

effluent releases into ground water (see Regulatory Guide 4.1). References that can be used in developing 

an adequate ground water site conceptual model include: 

1. ANSI/American Nuclear Society (ANS) 2.17-2007, “Evaluation of Subsurface Radionuclide 

Transport at Commercial Nuclear Power Production Facilities” (Ref. 19); 

2. NUREG/CR-6948, “Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities 

and Sites,” issued November 2007 (Ref. 20); and 

3. various other industry documents. 

Simple analytical models or more rigorous numerical codes (i.e., simulations) may be used to evaluate 

subsurface transport following a release. These models and codes will depend on the release rate, depth of 

the release, ground water flow directions, ground water flow rates, geochemical conditions, and other 

geochemical processes (e.g., geochemical retardation). Additionally, water usage such as ground water 

pumping from wells may create local groundwater depression(s) that can alter the natural groundwater 

flow. 

In general, sites with low-level spills or leaks (e.g., resulting in residual contamination that is within 

approximately a factor of 10 to 100 above the laboratory LLD for the nuclide of concern) generally do not 

require extensive site characterization and monitoring. Conversely, sites with a highly contaminated 

subsurface, which are likely to exceed applicable radiation protection standards (e.g., 10 CFR 20.1402 or 

10 CFR 20.1403, also refer to Table A-8), require more extensive evaluation. Initial assessments should 

be conducted with very simple models using scoping surveys and/or bounding assumptions. The 

complexity of the models should increase only as (1) more knowledge is obtained about the system under 

evaluation and as (2) the dose estimates rise above significant residual radioactivity levels (see definition in 

glossary). Many industry standard documents (Ref. 8, 19, 32) contain details of various industry practices 

that have been used as part of a groundwater monitoring program. 

Spills or leaks of radioactive material to the ground water should initiate some scoping or characterization 

activities. These scoping or characterization activities may be direct (i.e., occurring at, or very near, the 

source of the leak) or indirect (i.e., occurring at some distance from the source of the leak) depending on 

the accessibility of the source of the spill or leak. For spills or leaks occurring below the soil surface in 

inaccessible locations, direct scoping and characterization may not be possible. In these cases, indirect 

monitoring techniques (e.g., groundwater monitoring wells in a down gradient direction) should be used to 

satisfy existing regulatory requirements. These scoping or characterization activities should – at a 
DG-1186, Page 13

minimum – satisfy (1) the requirements of 10 CFR 50.75(g) as well as (2) the effluent reporting 

requirements of NUREG-1301/1302 typically associated with routine, planned effluents (e.g., sampling 

prior to discharge to unrestricted areas). Refer to sections 9.5.1, 9.5.2, and 9.5.9 for guidance on the 

general format for reporting abnormal releases and effluents to unrestricted areas. 

Scoping or characterization surveys may include elements such as groundwater monitoring wells, 

groundwater sampling, or soil sampling. However, the need for scoping or characterization surveys are 

sometimes incorrectly interpreted as a need for additional monitoring (e.g., sampling and additional wells) 

beyond that constituting a reasonable and necessary survey. Judicious use of site conceptual modeling 

together with a strategic and carefully planned monitoring program can ensure necessary and reasonable 

surveys are performed. Scoping and characterization surveys beyond what are necessary and reasonable to 

meet the regulatory requirements (e.g., surveys that are part of a voluntary industry initiative) may be 

reported in the ARERR. Indeed, the inclusion of such information in the ARERR is encouraged if it 

assists in the assessment of the annual effluents from the site. Groundwater information that is not related 

to the current year’s effluents or dose assessments, as required by the ARERR, may be more appropriately 

reported in the AREOR as part of the ongoing radiological environmental monitoring program (REMP). 

Refer to Regulatory Guide 4.1 and NUREG-1301/1302 for additional information on the REMP. 

Although licensees may conduct a groundwater monitoring effort for many different reasons, for purposes 

of this regulatory guide, the surveys, characterization activities, conceptual models, and other components 

of any groundwater monitoring effort should be sufficient to: 

1. appropriately report, for purposes of accountability, effluents discharged to unrestricted areas, 

2. document information in a format consistent with Table A-7 and regulatory position 9.5 

(Abnormal Releases), 

3. provide advance indication of potential future effluents to unrestricted areas (to ensure releases are 

planned and monitored prior to discharge), 

4. sufficiently demonstrate significant residual radioactivity has not migrated offsite to an unrestricted 

area in the annual reporting interval, and to 

5. communicate pertinent information to the NRC. 

4. Quality Assurance 

Regulatory Guidance – A range of quality control checks and tests should be applied to the analytical 

process. Regulatory Guide 4.15, Revision 1 and Revision 2, describe the QA program activities for 

ensuring radioactive effluent monitoring systems and operational programs meet their intended purpose. 

The applicability of Revision 1 or Revision 2, as a licensing commitment, is contained in each licensee’s 

licensing basis. Licensees with programs in operation prior to the issuance of Regulatory Guide 4.15, 

Revision 2 may rely exclusively in Revision 1. Regulatory Guide 4.15, Revision 2, contains guidance on 

determining appropriate sensitivity levels for analytical instrumentation based on data quality objectives 

(DQO). The use of DQO may provide a better technical basis for determining sensitivity levels (LLDs) 

than the use of default values supplied in NUREG-1301/1302. A combination approach (using both 

Regulatory Guide 4.15 Revision 1 and Revision 2) can be used to determine appropriate sensitivity levels 

(LLDs) different (i.e., higher or numerically larger) than those listed in NUREG-1301/1302. 

QC Checks – QC checks of laboratory instrumentation should be conducted daily or before use, and 

background variations should be monitored at regular intervals to demonstrate that a given instrument is in 
DG-1186, Page 14

working condition and functioning properly. QC records should include results of routine tests/checks, 

background data, calibrations, and all routine maintenance and service. 

Functional Checks – Routine qualitative tests/checks (e.g., channel operational tests, channel checks, or 

source checks to demonstrate that a given instrument is in working condition and functioning properly) 

may be performed using radioactive sources that are not National Institute of Standards and Technology 

(NIST) traceable. The schedule for source checks, channel checks, channel calibrations, and channel 

operational tests should be in accordance with NUREG-1301 and NUREG-1302. 

Procedures – Individual written procedures should be used to establish specific methods of calibrating 

installed radiological monitoring systems and grab sampling equipment. Calibration practices used for 

ancillary equipment and systems (e.g., meteorological equipment, airflow measuring equipment, in-stack 

monitoring pitot tubes) should be documented in written procedures. Calibration procedures may be 

compilations of published standard practices or manufacturers’ instructions that accompany purchased 

equipment, or they may be specially written in house to include special methods or items of equipment not 

covered elsewhere. Calibration procedures should identify the specific equipment or group of instruments 

to which the procedures apply. 

Written procedures should be used for maintaining counting room instrument accuracy, including 

maintenance, storage, and use of radioactive reference standards; instrumentation calibration methods; and 

QC activities such as collection, reduction, evaluation, and reporting of QC data. 

Calibration of Lab Equipment and Radiation Monitors – Calibrations (e.g., laboratory equipment and 

continuous radiation monitoring systems) should be performed using reference standards certified by NIST 

or standards that have been calibrated against NIST-certified standards. Calibration standards should have 

the necessary accuracy, stability, and range required for their intended use. Continuous radioactivity 

monitoring systems should be calibrated against appropriate NIST standards. The relationship between 

concentrations and monitor readings should be determined over the full range of the readout device. 

Adequacy of the system should be judged on the basis of reproducibility, time stability, and sensitivity. 

Periodic in-service correlations, that relate monitor readings to the concentrations and/or release rates of 

radioactive material in the monitored release path, should be performed to validate the adequacy of the 

system. These correlations should be based on the results of analyses for specific radionuclides in grab 

samples from the release path. 

The use of NIST traceable sources combined with mathematical efficiency calibrations may be applied to 

instrumentation used for radiochemical analysis (e.g., gamma spectroscopy systems) if using a method 

provided by the instrument manufacturer (e.g., Canberra’s ISOCS). 

Calibration of M&TE – Measuring and test equipment (M&TE) should be calibrated using reference 

standards certified by NIST or standards that have been calibrated against standards certified by NIST. 

The calibration standards should be representative of the sample types analyzed and have the necessary 

accuracy, stability, and range required for their intended use. 

Calibration Frequency – Calibrations should generally be performed at regular intervals in accordance 

with the frequencies established in NUREG-1301 and NUREG-1302. A change in calibration frequency 

(increased or decreased) should be based on the reproducibility and time stability characteristics of the 

system. For example, an instrument system that gives a relatively wide range of readings when calibrated 

against a given standard should be recalibrated at more frequent intervals than one that gives 

measurements within a more narrow range. Any monitoring system or individual measuring equipment 

should be recalibrated or replaced whenever it is suspected of being out of adjustment, excessively worn, 

or otherwise damaged and not operating properly. 
DG-1186, Page 15

Measurement Uncertainty – An estimate should be made of the measurement uncertainty (formerly 

called measurement error) associated with the measurement of radioactive materials in effluents. Counting 

statistics can provide an estimate of the statistical counting uncertainty involved in radioactivity analyses. 

Normally, the statistical counting uncertainty decreases as the amount (concentration) of radioactivity 

increases. Thus, for the radioactive effluent release report, the statistical counting uncertainty is typically a 

small component of the total uncertainty. The sampling uncertainty is likely the largest component and 

includes uncertainties such as the uncertainty in volumetric and flow rate measurements and laboratory 

processing uncertainties. 

The total or expanded uncertainty associated with the effluent measurement will include the cumulative 

uncertainties resulting from the total operation of sampling and measurement. The objective should be to 

evaluate the entire measurement process and obtain an approximate measure of the uncertainty associated 

with reported results. For monitoring batch and continuous releases, uncertainties in the estimates of 

volume, flow rate, radionuclide concentration, and gross radioactivity should be quantified and collectively 

assessed. Table 1 summarizes various uncertainty components that should be assessed to quantify the 

expanded uncertainty in batch and continuous monitoring systems. For each system component, one or 

multiple methods can be chosen to quantify the uncertainty in each component. The total uncertainty can 

be calculated by taking the square root of the sum of squares of each system component uncertainty. 

NUREG-1576, “Multi-Agency Radiological Laboratory Analytical Protocols Manual,” issued July 2004 

(Ref. 21), and ANSI/HPS N13.1-1999 provide further details for quantifying uncertainty. 

Table 1. Calculating Total System Uncertainties 

Release Type System Component Uncertainty 

Bulk volume in 

storage/holdup tank 

• Calibration tolerance (± %) of volumetric 

measuring device (volume or pressure 

gauge) for storage and holdup tanks 

• % difference between two or more different 

methods of estimating bulk volume 

Flow rate • Calibration tolerance (± %) of flow meter 

• % variation over time of multiple 

measurements using same flow meter 

• % variation between two or more different 

flow meters (each meter should be properly 

calibrated) 

• % difference between two or more different 

methods of estimating flow rate 

Batch 

Radionuclide 

concentration 

• % variation between results of replicate grab 

samples 

• % variation between results of grab and 

composite samples 

Continuous Flow rate or velocity 

in effluent and 

sampling systems 

• Performance criterion (% tolerance) of initial 

check of systems 

• % variability over time in flow 

measurements 

• % tolerance of flow calibrations 
DG-1186, Page 16

• % difference between two or more different 

methods of estimating flow rate 

Radionuclide 

concentration 

• % variation between results of replicate grab 

samples 

• % variation between results of grab and 

composite samples 

Continuous radiation 

monitors for gross 

radioactivity 

• Performance criterion (% tolerance) of initial 

check of measurement system for gases and 

aerosols 

• % tolerance of periodic monitor calibration 

• % variability over time in continuous 

radioactivity measurements 

5. Dose Assessments for Members of the Public 

The regulations referenced in regulatory positions 5.5 through 5.7 contain both dose limits and design 

objectives that the licensee demonstrates compliance with through calculations. The fundamental 

parameters associated with the dose calculations are summarized in Table 2. Important concepts for these 

calculations are discussed in regulatory positions 5.8 through 5.10. Due to differences between these 

regulations, only demonstrating compliance with radiological effluent technical specifications (based on 

Appendix I to 10 CFR Part 50) does not necessarily ensure compliance with 10 CFR 20.1301(a) or 40 

CFR Part 190, particularly if there is a direct radiation component (e.g., from BWR shine, ISFSI, or 

radioactive materials storage). 

5.1 Bounding Dose Assessments may be used for purposes of demonstrating compliance with 

regulatory dose limits or criteria, particularly in those circumstances where compliance can be 

readily demonstrated using conservative assumptions. Dose assessments can be performed using 

conservative assumptions that result in a bounding, overestimation of the dose to the highest real 

individual. For example, instead of assessing dose to a real individual bounding assessments can 

be performed at the site boundary. If bounding assumptions are made, the radioactive effluent 

release report should state such and should annotate the assumptions. Additionally, if bounding 

assessments are used and are based on hypothetical exposure pathways (which may yield inflated 

doses), do not use the bounding assessment as the basis for determining “significant exposure 

pathways” as described in regulatory position “C” of Regulatory Guide 1.109. 

5.2 Occupational Workers are those individuals who, in the course of employment, are assigned 

duties involving exposure to radiation, or exposure to radioactive material. Training per 10 CFR 

19.12 is required for those individuals that may exceed 100 mrem/year. Each licensee is 

responsible for classifying individuals (by location) as either members of the public or as 

occupational workers. See definition of “occupational dose” in 10 CFR Part 20. For example a 

licensee may choose to classify all individuals in the controlled area, including a vending machine 

delivery person, or a coal plant worker as occupational workers since the assigned duties involve 

exposure to radiation at the adjacent facility. An occupational worker is a member of the public 

while in the unrestricted area. For more information, see requirements under 10 CFR 19.11. 

5.3 Members of the public reside in the unrestricted area, but at times may enter the controlled area 

of a commercial nuclear power plant. Each licensee is responsible for classifying individuals (by 
DG-1186, Page 17

location) as either members of the public or as occupational workers. See definition of “members 

of the public” in 10 CFR Part 20. An occupational worker is a member of the public while in the 

unrestricted area. 

If bounding assessments are not used, licensees should perform evaluations to determine the 

highest dose to an actual member of the public, regardless of whether the individual is in an 

unrestricted area or a controlled area. If no member of the public is allowed in the controlled area, 

the evaluation need only consider members of the public in the unrestricted area. For example a 

licensee may choose to classify all individuals in the controlled area, including a vending machine 

delivery person, as members of the public. A member of the public is a real individual in a 

designated location where there is a real exposure pathway (e.g., a real garden, real cow, real goat, 

or actual drinking water supply); not a fictitious fencepost resident or pathway via a virtual goat or 

cow. 

5.4 Occupancy factors for members of the public in the unrestricted area should be assumed to be 

100 percent unless the land-use census provides site-specific information indicating otherwise. 

Occupancy factors may be applied inside the controlled area based on site usage characteristics for 

members of the public or occupational workers. 

5.5 10 CFR Part 50, Appendix I, contains numerical guidance for design objectives and limiting 

conditions of operation for radwaste systems to ensure radioactive liquid and gaseous effluents to 

unrestricted areas are ALARA. This numerical guidance is listed in terms of annual air doses 

(gamma and beta), annual total body doses, and annual organ doses. Technical Specifications 

requires that exposure to liquid and gaseous effluents conform to the numerical guidance in 10 

CFR 50 Appendix I. Per 10 CFR 50.34a, these numerical guides for design objectives and 

limiting conditions of operation are not to be construed as radiation protection standards. For 

these dose calculations, the operative verbiage is: 

5.5.1 air doses (gamma and beta), total body doses, and organ doses (based on ICRP-2) 

5.5.2 effluent releases only (excludes direct radiation from the facility and ISFSIs) 

5.5.3 current annual period (excludes accumulated radioactivity from prior-year effluents) 

5.5.4 unrestricted area (excludes individuals in the restricted areas and controlled areas) 

5.6 10 CFR 20.1301(a) through (c) specifies dose limits for members of the public from all licensed 

and unlicensed sources associated with the licensed facility (see 10 CFR 20.1001, “Purpose”). 

This includes doses from (1) current-year effluents released to unrestricted areas, (2) current-year 

contributions from direct radiation (e.g., nitrogen-16, radioactive materials storage, radioactive 

waste storage, ISFSIs), and (3) radioactivity remaining in the unrestricted area from prior-year 

operations. The limits are applicable to members of the public in unrestricted areas and controlled 

areas. This limit is based on TEDE, which is distinct from total body dose and organ dose (based 

on ICRP-2). 

5.6.1 The “current-year” excludes doses received in prior years. 

5.6.2 “TEDE” is a ICRP-26 term (that is distinct from total body dose and organ dose which are 

ICRP-2 terms). 

5.6.3 “Locations” include the unrestricted area and controlled areas (if members of the public 

are present on-site). 
DG-1186, Page 18

5.6.4 “Sources that exist or remain in the area” includes doses from radioactivity remaining in 

the environment as a result of prior-years effluents (such as accumulated radioactivity in 

lake water from prior year releases). For example, based on the land-use census, if 

drinking water is an actual exposure pathway from a lake containing accumulated 

radioactivity, the dose assessment must include contributions from both the effluent 

release during the current year and the accumulated radioactivity from prior-year discharge 

5.7 10 CFR 20.1301(e) requires compliance with EPA's generally applicable environmental radiation 

standards promulgated in 40 CFR Part 190, “Environmental Radiation Protection Standards for 

Nuclear Power Operations,” (Ref. 22). Licensees must assess the highest cumulative (whole body 

and organ) doses from the uranium fuel cycle to a real individual outside the site boundary. The 

limits include (1) contributions from current-year effluents, (2) current-year direct radiation from 

the facility, and (3) accumulated radioactivity from prior-year effluents. These requirements 

include the following: 

5.7.1 “Whole body and organ doses” are ICRP-2 terms. 

5.7.2 “Any member of the public” means any individual (i.e., a real person), in an unrestricted 

area, who is not engaged in nuclear fuel cycle operations. 

5.7.3 The “unrestricted area” means in the general environment outside the (boundaries of) 

locations under the control of persons possessing or using radioactive material. This 

excludes the controlled area and the restricted area. See definition of “generally 

applicable environmental radiation standards” in 10 CFR 20.1003. 

5.7.4 “Current-year effluents” includes both normal and abnormal releases to the unrestricted 

area. 

5.7.5 “Current-year direct radiation” includes all direct radiation from the facility (e.g., 

radioactive waste storage, and ISFSIs), but excludes doses from radioactive waste 

shipments. 

5.7.6 “Cumulative” dose means the sum of (1) current year effluent dose, (2) current year direct 

radiation dose, and (3) dose from accumulated radioactivity. 

5.7.7 “Accumulated radioactivity” includes radioactive material from prior-year effluent 

releases that may have accumulated in lake water. 

5.7.8 The “uranium fuel cycle” excludes uranium mining, radioactive waste shipping (in the 

unrestricted area), operations at waste disposal sites, and reuse of non-uranium special 

nuclear materials (per 40 CFR 190). 

Table 2, Parameters Associated with Dose Calculations 

10 CFR 20.1301(a)-(c) 10 CFR 50, Appx. I 10 CFR 20.1301(e) 

Dose TEDE Whole Body, Max of 

Any Organ, Gamma 

Air, and Beta Air 

Whole Body, Thyroid, 

and Max of Any Organ 
DG-1186, Page 19

Basis ICRP-26 ICRP-2 EPA 40 CFR 190 

Where Controlled Area and 

Unrestricted Area 

Unrestricted Area Unrestricted Area 

Individual 

receptor 

Real person/pathway 

(nearest real residence, 

real garden, real 

dairy/meat animal) 

Real person/pathway 

(nearest real residence, 

real garden, real 

dairy/meat animal) 

Real person/pathway 

(nearest real residence, 

real garden, real 

dairy/meat animal) 

Sources Liquid and Gas Radwaste, 

Direct Radiation 

(e.g., Shine, N-16, ISFSI, 

Rad. Mat. Storage, 

Outside Tanks, etc...) 

Residual radioactivity 

(e.g., lake water, 

return/recycle) 

Liquid and Gas 

Radwaste, 

Liquid and Gas 

Radwaste, 

Direct Radiation 

(e.g., Shine, N-16, 

ISFSI, Rad. Mat. 

Storage, Outside Tanks, 

etc...) 

Residual radioactivity 

(e.g., lake water, 

return/recycle) 

Sources Licensed and Unlicensed Licensed Only Licensed and 

Unlicensed 

When Current year, plus 

residual radioactivity 

from previous years. 

Current year. Current year, plus 

residual radioactivity 

from previous years. 

5.8 Dose assessments to show compliance with 10 CFR Part 50, Appendix I design objectives, 

should include quarterly and annual doses using the considerations of regulatory position 5.5, and 

should be reported in a format similar to that listed in Table A-4. 

5.8.1 Doses from liquid effluents 

5.8.1.1 Total body dose, quarterly and annual 

5.8.1.2 Organ dose, quarterly and annual (maximum, any organ) 

5.8.1.3 Percent of limits for each of the above 

5.8.1.4 Sector and distance 

5.8.2 Doses from gaseous effluents: 

5.8.2.1 beta and gamma air doses, quarterly and annual 

5.8.2.2 organ dose commitment from iodine, tritium, and particulate releases with 

half-lives greater than 8 days, quarterly and annual 

5.8.2.3 Percent of limit for each of the above 

5.8.2.4 Sector and distance 

5.8.3 An evaluation of the local exposure pathways to determine the highest exposed member of 

the public should be performed. However, maximum doses from various exposure 

pathways should not be additive from different locations. For example, dose from a 

downstream drinking water pathway should not be added to the dose to an upstream 
DG-1186, Page 20

resident whose exposure is from gaseous effluents and direct radiation. 

The highest doses (often referred to as the “Maximum” doses) to real individuals are 

assessed as described in Regulatory Guide 1.109, “Calculation of Annual Doses to Man 

from Routine Releases of Reactor Effluents for the Purpose of Demonstrating Compliance 

with 10 CFR Part 50, Appendix I” (Ref. 23). The locations and exposure pathways are 

those where actual individuals are present and exposed. Maximum exposed individuals 

are characterized as “maximum” with regard to food consumption, occupancy, and other 

usage of the region in the vicinity of the plant site. For example, licensees should make 

“maximum” (hypothetical) assumptions for food consumption and occupancy factors at 

actual locations when assessing dose to the maximum actual exposed individual, unless 

they have determined and applied site-specific (actual) data. In this context, 

“hypothetical” applies to the various factors for a particular individual (e.g., consumption 

factors, occupancy factors, and age groups) and not the individual’s location. 

5.8.4 The objective of Appendix I is to provide numerical guides for design objectives and 

limiting conditions for operation to ensure that effluent control equipment is effective in 

reducing emissions to ALARA levels. The numerical guidance pertains to quarterly and 

annual dose criteria at or beyond the unrestricted area from current-year effluent releases. 

The calculations do not include dose from radioactivity in prior-year, accumulated, 

effluent releases (e.g., last year’s radioactivity remaining in lake water is excluded). 

5.8.5 Significant exposure pathways should be considered (and included) in all dose 

assessments. In addition to the exposure pathways identified in Regulatory Guide 1.109, 

other exposure pathways that may arise because of unique conditions at a specific site 

(such as inadvertent leaks to ground water) should be considered if they are likely to 

contribute significantly to the total dose. A pathway is considered significant if a realistic 

evaluation yields an additional dose increment equal to or more than 10% of the total from 

all pathways considered (see the regulatory position in Regulatory Guide 1.109). An 

evaluation of less significant exposure pathways (not included in dose assessments) should 

be performed and maintained for purposes of demonstrating compliance with this 

requirement. 

5.8.6 Real exposure pathways are identified based on the results of the Land Use Census. Dose 

calculations should be performed based on realistic exposure pathways. For example, if 

the land use census does not identify a cow (or goat) milk pathway, the licensee is not 

required to assess that pathway. Similarly, if a licensee discharges liquid effluent to a 

body of water that is not a drinking water pathway (e.g., no local intake from private or 

public supplies), a drinking water assessment is not required for that release point. 

5.9 Dose assessments to show compliance with 10 CFR 20.1301(a) thru (c) should be reported 

with consideration of regulatory position 5.6 and in a format similar to that listed in Table A-5. 

5.9.1 Report the following 

5.9.1.1 Liquid effluent contribution to TEDE dose 

5.9.1.2 Gaseous effluent contribution to TEDE dose 

5.9.1.3 Direct radiation contribution to TEDE dose 

5.9.1.4 TEDE dose (sum from all contributions) 
DG-1186, Page 21

5.9.1.5 Location (on-site or offsite and distance/direction) applicable to the highest 

exposed member of the public 

5.10 Dose assessments to show compliance with 10 CFR 20.1301(e) should be reported according to 

the generally applicable environmental radiation standards promulgated by EPA at 

40 CFR Part 190, with consideration of regulatory position 5.7, and in a format similar to that 

listed in Table A-6. 

5.10.1 Report the following: 

5.10.1.1 Whole body dose 

5.10.1.2 Thyroid dose 

5.10.1.3 Dose to any other organ 

5.10.1.4 Sector and distance 

5.10.1.5 Percent of the applicable limit 

5.10.2 One means of demonstrating compliance with 40 CFR 190 is listed in Federal Register 42 

FR 2859 that states: 

“In the case of light water reactors, ... demonstrating 

conformance with Appendix I of 10 CFR 50 are generally 

adequate for demonstrating compliance with [EPA 40 CFR 190].” 

As a result, a licensee who (1) can demonstrate external sources of direct radiation are 

indistinguishable from background and who (2) demonstrates compliance with the numerical 

dose guidance of 10 CFR 50 Appendix I may cite the above reference as the basis for 

demonstrating compliance with 40 CFR 190. 

However, for licensees who (1) have external sources of direct radiation that are above 

background, and who (2) demonstrate compliance with the numerical dose guidance of 10 

CFR 50 Appendix I may not cite the above reference as the basis for demonstrating 

compliance with 40 CFR 190. 

5.10.3 The dose contributions from direct radiation may be estimated based on either (1) 

thermoluminescent dosimetry (TLD) measurements, (2) calculations, or (3) a combination 

of TLD measurements and calculations. When direct radiation dose is determined by 

TLD, background levels of radiation from selected control locations should be subtracted 

from applicable indicator locations. The doses measured from control and indicator 

locations should be taken from the same time period. When choosing the appropriate 

control location(s) consideration should be given to the variability in doses measured at 

the control and indicator locations. 

5.11 An Example Calculation demonstrates how to apply the information contained in regulatory 

positions 5.8 through 5.10 to calculate the three separate types of doses (i.e., Technical 

Specifications required compliance with the numerical guidance of 10 CFR 50 Appendix I, 10 

CFR 20.1301, and 40 CFR 190). For the calculations, assume the dose components below have 

been determined. 
DG-1186, Page 22

5.11.1 Annual Whole Body Doses, nearest residence 

5.11.1.1 Liquid Radwaste 0.001 mrem 

5.11.1.2 Gas Radwaste 0.012 mrem 

5.11.2 Annual Organ Doses, nearest residence 

5.11.2.1 Liquid Radwaste 0.002 mrem 

5.11.2.2 Gas Radwaste 0.015 mrem 

5.11.3 Direct Radiation (from ISFSI, shine, etc), nearest residence 

5.11.3.1 Net annual dose at nearest residence = 48.00-47.92 = 0.08 mrem 

5.11.3.2 The propagation of uncertainty in measurements = 2 2 (4.00) + (3.06) = ± 

5.03 mrem. This calculation may or may not be possible at all sites depending 

on how the Tads are configured in the radiological environmental monitoring 

program. The calculation may be of interest since it may help demonstrate the 

relative significance (or insignificance) of the net annual dose at the nearest 

residence calculated in this example. 

5.11.3.3 In this example, the net dose to the nearest residence was based on one of two 

TLDs at a control location. Either TLD could be used provided it was 

representative of the background. Alternatively, both control TLDs could be 

used to generate an average background (provided both were representative of 

the background). 

5.11.4 Accumulated radioactivity (H-3 in on-site pond from previous years of operation) 

5.11.4.1 Whole body 0.016 mrem, from on-site pond H-3 evaporation to nearest 

residence (inhalation and submersion) 

5.11.4.2 Each organ 0.016 mrem, from on-site pond H-3 evaporation to nearest 

residence (inhalation and submersion) 

5.11.4.3 Whole body and all organs 0.0017 mrem, from on-site pond H-3 evaporation 

to nearest residence (ingestion, leafy vegetables) 

5.11.4.4 Each organ 0.0017 mrem, from on-site pond H-3 evaporation to nearest 

residence (ingestion, leafy vegetables) 

5.11.5 Based on the above information, the dose calculations are performed as shown below. 

5.11.5.1 For 10 CFR 50 Appendix I 

Direct Radiation from TLDs and 

Calculations 

True 

gross 

mrem/yr

Uncertainty 

+/- mrem/yr 

TLD at First Control Location 47.92 3.06 

TLD at Second Control Location 49.95 4.02 

TLD, Indicator, 50 ft away, on-site 152.95 26.32 

TLD, Indicator, 0.75 mi away, on-site 50.68 5.72 

Nearest Residence, 1.2 mi away, 

(calculated based on extrapolation of 

Indicator TLDs, Gross) 

48.00 4.00 
DG-1186, Page 23

5.11.5.1.1 Whole Body Dose = 0.001+0.012+0.016+0.017 = 0.046 mrem 

5.11.5.1.2 Organ Dose = 0.002+0.015+0.016+0.017 = 0.050 mrem 

5.11.5.2 For 10 CFR 20.1301(a) thru (c) (TEDE) 

5.11.5.2.1 TEDE = 0.001+0.012+0.08+0.016+0.017= 0.126 mrem 

5.11.5.3 For 40 CFR 190 

5.11.5.3.1 Whole Body Dose = 0.001+0.012+0.08+0.016+0.017 = 0.126 

mrem 

5.11.5.3.2 Organ Dose = 0.002+0.015+0.08+0.016+0.017 = 0.130 mrem 

6. Solid Radioactive Waste Shipped for Processing or Disposal 

The total curie quantity and radionuclide composition of the solid waste shipped off site should be 

determined and reported in a format similar to Table A-3 in Appendix A to this guide. The data should be 

reported separately for shipments made to offsite waste processors and those made directly to low-level 

waste (LLW) disposal facilities. Shipments for direct disposal using authorizations obtained under 

10 CFR 20.2002, “Method for Obtaining Approval of Proposed Disposal Procedures,” or equivalent State 

regulations should also be reported separately. Shipments of processed waste and used (irradiated) fuel for 

storage should be reported. 

7. Reporting Errata in Effluent Release Reports 

Errors in radioactive effluent release reports should be classified and reported as follows: 

7.1 Small errors may be any of the following: 

7.1.1 Inaccurate reporting of dose that equates to ≤ 10% of the public dose, after correction 

7.1.2 Inaccurate reporting of curies (or release rates, volumes, etc) that equate to ≤ 10% of the 

affected curie total (or release rate, volume, etc), after correction 

7.1.3 Omissions that do not impede the NRCs ability to adequately assess the information 

supplied by the licensee 

7.1.4 Typographical errors or other errors that do not alter the intent of the report 

7.2 Small errors should be corrected, and the affected pages should be submitted as a supplemental 

report (referencing the affected calendar year). The supplemental report (or addendum) to the 

radiological effluent release report may be submitted coincidental with the next (normally 

scheduled) submittal of the ARERR. 

7.3 Large errors may be any of the following 

7.3.1 Inaccurate reporting of dose that equates to > 10% of public dose, after correction 

7.3.2 Inaccurate reporting of curies (or release rates, volumes, etc) that equate to > 10% of the 

affected curie total (or release rate, volume, etc), after correction 

7.3.3 Omissions that may impede the NRCs ability to adequately assess the information 

supplied by the licensee 

7.3.4 Typographical errors or other errors that do significantly alter the intent of the report 

7.4 Large errors should be corrected, and the affected pages should be submitted as a supplemental 

report (referencing the affected calendar year). The supplemental report (or addendum) to the 
DG-1186, Page 24

radiological effluent release report should be submitted to the NRC within 90 days of discovery of 

the error. 

8. Data Trending 

The annual effluent release data should be trended over a 10-year period and submitted with the ARERR 

as follows: 

8.1 Liquid effluent trends: 

8.1.1 source term for curies of total mixed fission and activation products 

8.1.2 source term for curies of tritium 

8.1.3 Appendix I dose (percent of total body dose numerical guidance) 

8.1.4 Appendix I dose (percent of maximum organ dose numerical guidance) 

8.2 Gaseous effluent trends: 

8.2.1 source term for curies of total mixed fission and activation products 

8.2.2 source term for curies of particulates 

8.2.3 source term for curies of iodine 

8.2.4 source term for curies of tritium 

8.2.5 Appendix I dose (percent of gamma air dose numerical guidance) 

8.2.6 Appendix I dose (percent of beta air dose numerical guidance) 

8.2.7 Appendix I dose (percent of maximum organ dose numerical guidance) 

8.3 Direct radiation trends 

8.3.1 Dose to the highest exposed real individual 

9. Format and Content of the Annual Radioactive Effluent Release Report (ARERR) 

In accordance with 10 CFR 50.4, “Written Communications,” the annual report should be submitted 

electronically or in a written communication in a format similar to that given in Tables A-1 through A-6 in 

Appendix A to this guide. Effluent data should be summarized on a quarterly and annual basis. For 

purposes of compliance with 10 CFR 50.36a, the ARERR must be submitted by May 1 (unless a licensing 

basis exists for a different submittal date) for effluents and solid waste from the previous calendar year. 

Measurements should be reported for positive values; if results are determined to be below detectable 

levels, the table entry should be left blank. 

The format specified in revision 2 of this regulatory guide differs slightly from that specified in revision 1. 

 The format and content as specified in revision 2 (and listed in the paragraphs below) are one acceptable 

method of reporting the data. Other formats may be used (e.g., some tables may be combined) as long as 

the specified content is satisfied (e.g., quarterly totals and annual totals by each release category). All 

plants are encouraged to use the format listed below to maximize consistency in data reporting. The aspect 

of consistency aids review by members of the public and allows easier industry-wide comparisons of the 

data. 

9.1 Gaseous Effluents

9.1.1 Table A-1—Gaseous Effluents – Summation of All Releases contains a summation of all 

gaseous effluent releases from all release points and all modes of release. The data is 

subdivided by quarter and year for each radionuclide category: (a) fission and activation 

gases, (b) iodines/halogens, (c) particulates, (d) tritium, and (e) gross alpha. 

9.1.2 Table A-1A, Gaseous Effluents—Ground-Level Release—Batch Mode contains a 

summation of gaseous effluent releases from ground-level release points in the batch mode 
DG-1186, Page 25

of release for the five radionuclide categories of fission gases, iodines/halogens, 

particulates, tritium, and gross alpha. Report the following: 

9.1.2.1 curies of each radionuclide released by quarter and year, and 

9.1.2.2 total curies released in each radionuclide category (fission and activation 

gases, iodines/halogens, particulates, tritium, and gross alpha) by quarter and 

year. 

9.1.3 Table A-1B, Gaseous Effluents—Ground-Level Release—Continuous Mode contains a 

summation of gaseous effluent releases from ground-level release points in the continuous 

mode of release for the five radionuclide categories of fission gases, iodines/halogens, 

particulates, tritium, and gross alpha. Report the following: 

9.1.3.1 curies of each radionuclide released by quarter and year, and 

9.1.3.2 total curies released in each radionuclide category by quarter and year. 

9.1.4 Table A-1C, Gaseous Effluents—Elevated Release—Batch Mode contains a summation 

of gaseous effluent releases from elevated release points in the batch mode of release for 

the five radionuclide categories of fission gases, iodines/halogens, particulates, tritium, 

and gross alpha. Report the following: 

9.1.4.1 curies of each radionuclide released by quarter and year, and 

9.1.4.2 total curies released in each radionuclide category by quarter and year. 

9.1.5 Table A-1D, Gaseous Effluents—Elevated Release—Continuous Mode contains a 

summation of gaseous effluent releases from elevated release points in the continuous 

mode of release for the five radionuclide categories of fission gases, iodines/halogens, 

particulates, tritium, and gross alpha. Report the following: 

9.1.5.1 curies of each radionuclide released by quarter and year, and 

9.1.5.2 total curies released in each radionuclide category by quarter and year. 

9.1.6 Table A-1E, Gaseous Effluents—Mixed Mode Release—Batch Mode contains a 

summation of gaseous effluent releases from mixed-mode release points in the continuous 

mode of release for the five radionuclide categories of fission gases, iodines/halogens, 

particulates, tritium, and gross alpha. Report the following: 

9.1.6.1 curies of each radionuclide released by quarter and year, and 

9.1.6.2 total curies released in each radionuclide category by quarter and year. 

9.1.7 Table A-1F, Gaseous Effluents—Mixed Mode Release—Continuous Mode contains a 

summation of gaseous effluent releases from mixed-modes release points in the 

continuous mode of release for the five radionuclide categories of fission gases, 

iodines/halogens, particulates, tritium, and gross alpha. Report the following: 

9.1.7.1 curies of each radionuclide released by quarter and year, and 

9.1.7.2 total curies released in each radionuclide category by quarter and year. 

9.2 Liquid Effluents

A summary of liquid effluent release information should be reported in the format of Appendix A, Table 

A-2. The quarterly and annual sums of all releases of radioactive materials in liquid effluents (i.e., routine 

and abnormal occurrences, continuous, and batch) should be reported in a format similar to that of the 

Tables A-2A and A-2B. 

9.2.1 Table A-2, Liquid Effluents - Summation of All Releases contains a summation of all 

liquid effluent releases from all release points and all modes of release. The data is 

subdivided by quarter and year for each radionuclide category: (a) fission and activation 

products, (b) tritium, (c) dissolved and entrained noble gases, and (d) gross alpha. Also 
DG-1186, Page 26

include total volume of primary coolant waste (typically batch mode releases), before 

dilution. This is generally the higher activity waste that is processed through the liquid 

radioactive waste system. Report the total measured volume or average flow rate of waste 

from secondary or balance of plant systems (e.g., steam generator blow-down). This is 

typically very low activity waste that is generally not processed with the liquid radioactive 

waste treatment system. Report internal plant dilution flow rate during periods of release

(in-plant dilution flow, before effluent discharge to receiving water body), and external 

dilution flow rate, average (river or stream flow rates). 

9.2.2 Table A-2A, Liquid Effluents - Batch Mode contains a summation of liquid effluent 

releases in the batch mode of release. The table is divided into the four categories of 

fission and activation products, tritium, dissolved and entrained gases, and gross alpha. 

Report the following: 

9.2.2.1 curies of each radionuclide and gross alpha released by quarter and year, 

9.2.2.2 total curies in each radionuclide category by quarter and year, 

9.2.3 Table A-2B, Liquid Effluents—Continuous Mode contains a summation of liquid 

effluent releases in the continuous mode of release. The table is divided into the four 

radionuclide categories of particulates and iodines, dissolved and entrained gases, tritium, 

and gross alpha. Report the following: 

9.2.3.1 curies of each radionuclide and gross alpha released by quarter and year, 

9.2.3.2 total curies in each radionuclide category by quarter and year, 

9.3 Solid Waste Shipments

Appendix A, Table A-3 summarizes the solid radioactive waste (LLW) and used (irradiated) fuel 

shipments during the reporting period. Quantities should be reported for the following wastes: 

9.3.1 shipped for direct disposal to licensed waste disposal facilities, 

9.3.2 shipped for processing before disposal or before returned to site, 

9.3.3 returned to site for long-term storage, 

9.3.4 shipped for 10 CFR 20.2002 disposal, and 

9.3.5 used (irradiated) fuel shipments. 

The total volume of waste, activity, and principal isotopes should be specified for Waste Classifications A 

and B. Other waste that does not fit into these categories should also be reported. For the used fuel 

shipments, the report should summarize the number of shipments, destination, and mode of transport (e.g., 

rail or truck). 

9.4 Dose Assessments

The annual evaluations of public dose should be calculated using the regulatory guidance of section 5 and 

should be reported in the format of Table A-4, Table A-5, and Table A-6. Dose assessments should be 

performed to demonstrate compliance with the following: 

9.4.1 10 CFR Part 50, Appendix I (Table A-4): 

9.4.1.1 Report the calculated dose from liquid effluents on a quarterly and annual 

basis to the total body and maximum organ and the percentage of the 

Appendix I design objectives. This is the dose from the annual discharge of 

liquid effluents to a real individual in the unrestricted area. If a particular 

pathway is not applicable (i.e., it does not exist at a site) no dose should be 

calculated for that pathway. 
DG-1186, Page 27

9.4.1.2 Report the highest air dose from gaseous effluents on a quarterly and annual 

basis at any location that could be occupied by individuals in the unrestricted 

area and the percentage of the Appendix I design objectives. 

9.4.1.3 Report the highest organ dose from iodine, tritium, and particulates with a 

half-life greater than 8 days to any individual in an unrestricted area from all 

pathways of exposure (e.g., submersion and ingestion). 

9.4.2 10 CFR 20.1301(a)-(c) (Table A-5): 

9.4.2.1 Report the TEDE dose from both licensed and unlicensed sources (including 

accumulated radioactivity from prior-year releases) to a real individual likely 

to receive the highest dose, whether in the controlled area or unrestricted area. 

 The TEDE can be calculated directly as TEDE dose or converted from whole 

body and organ dose. Identify whether the real individual likely to receive the 

highest dose is located within the controlled area or the unrestricted area. 

9.4.3 10 CFR 20.1301(e) and 40 CFR Part 190 (Table A-6): 

9.4.3.1 Report the whole body, thyroid, and highest dose to any other organ from 

licensed and unlicensed sources in the uranium fuel cycle, excluding 

background, to the real individual likely to receive the highest dose. Identify 

the location (sector and distance) of the highest exposed real individual. 

9.5 Supplemental Information

Table A-7 in the appendix can be used to provide supplemental information in a descriptive format in 

a narrative form. Relevant information and a description of circumstances should be provided as 

appropriate for each the following categories, adding categories as appropriate. Use the annotation 

N/A if not applicable. 

9.5.1 Abnormal Releases

9.5.1.1 Specific information should be reported concerning abnormal (airborne and/or 

liquid) releases to the site environs. The report should describe the event in a 

way that would enable the NRC to adequately understand how the material 

was released and if there is any offsite impact. In the case of a liquid release, 

the report should describe the potential impact on ground water or drinking 

water pathways, as applicable. 

9.5.1.2 Low-level radioactive system leakage from minor equipment failures and 

component aging (wear and tear) may be expected to occur as an anticipated 

part of the plant operation. If such leakage is directed to a system designed to 

accept and handle radioactive material, it is not considered an abnormal 

release. For example, gaseous effluent leakage into a plant ventilation system 

that was designed to handle leakage from various plant components is not an 

abnormal release. 

9.5.1.3 The following are the thresholds for reporting abnormal releases in the 

supplemental information section. 

9.5.1.3.1 abnormal releases to on-site areas voluntarily reported to local 

authorities under the industry Ground Water Protection Initiative 

(Ref. 31); 

9.5.1.3.2 abnormal releases to on-site areas exceeding 100 gallons of 

radioactive liquid (or where the volume is unknown), where the 

presence of licensed radioactive material is positively identified 

(in either the site environs or in the source of the leak/spill) as 
DG-1186, Page 28

greater than the minimum detectable activity (MDA) for the 

laboratory instrumentation; 

9.5.1.3.3 abnormal releases to on-site areas that result in detectable residual 

radioactivity after remediation 

9.5.1.3.4 abnormal releases that result in a high effluent radiation alarm 

without an anticipated system trip occurring. 

9.5.1.3.5 abnormal releases to an unrestricted area 

9.5.1.4 Information on abnormal releases should include the following 

9.5.1.4.1 date of the release and duration, 

9.5.1.4.2 location of the release, 

9.5.1.4.3 volume of the release, 

9.5.1.4.4 estimated activity of each radionuclide, 

9.5.1.4.5 effluent monitoring results (if any), 

9.5.1.4.6 environmental monitoring results (if any), 

9.5.1.4.7 classification(s) of subsurface aquifer(s) (Class I drinking water, 

Class II impaired aquifer, or Class III unfit for drinking water), 

9.5.1.4.8 size and extent of any ground water plume, 

9.5.1.4.9 expected movement/mobility of any ground water plume, 

9.5.1.4.10 land use characteristics, 

9.5.1.4.11 remedial actions considered or taken and results obtained, 

9.5.1.4.12 calculated member of the public dose attributable to the release, 

9.5.1.4.13 actions taken to prevent recurrence, as applicable, 

9.5.1.4.14 whether the NRC and/or local officials were notified and the 

date(s) and contact person(s), 

9.5.1.4.15 residual radioactivity levels and whether any levels exceed the 

DCGLs provided in Table H.2 in Appendix H to NUREG-1757, 

Volume 2, 

9.5.2 Nonroutine Planned Discharges

9.5.2.1 Report pumping of leaks and spills for remediation or results of ground water 

monitoring to quantify effluent releases to the offsite environment 

9.5.3 Radioactive waste treatment system changes

9.5.4 Annual Land Use Census Changes

9.5.5 Effluent Monitoring System Inoperability

9.5.5.1 Report all effluent radiation monitoring systems that are out of service 30 

consecutive days or more in accordance with NUREG-1301 and NUREG1302, Section 3.3.3.10.b. 

9.5.5.2 Note that per NUREG-1301 and NUREG-1302, Section 3.3.3.10.b, a report is 

only required when the minimum channels operability requirement is not 

achieved. 

9.5.6 Offsite Dose Calculation Manual Changes

9.5.7 Process Control Program Changes

9.5.8 Supplemental Reports for previous Annual Radioactive Effluent Release Reports

9.5.8.1 Omissions from previously submitted Arrears 
DG-1186, Page 29

9.5.8.2 Errata to correct errors identified in previous Arrears 

9.5.9 Other (narrative descriptions of other information related to radioactive effluents)
DG-1186, Page 30

D. IMPLEMENTATION 

The purpose of this section is to provide information to applicants and licensees regarding the NRC’s plans 

for using this draft regulatory guide. The NRC does not intend or approve any imposition or backfit in 

connection with its issuance. 

The NRC has issued this draft guide to encourage public participation in its development. The NRC will 

consider all public comments received in development of the final guidance document. In some cases, 

applicants or licensees may propose an alternative or use a previously established acceptable alternative 

method for complying with specified portions of the NRC’s regulations. Otherwise, the methods described 

in this guide will be used in evaluating compliance with the applicable regulations for license applications, 

license amendment applications, and amendment requests.

REGULATORY ANALYSIS

1. Statement of the Problem 

Regulatory Guide 1.21, Revision 1, was issued in June 1974. The regulatory guide described programs 

acceptable to the regulatory staff for measuring, evaluating, and reporting releases of radioactive materials 

in liquid and gaseous effluents and guidelines for classifying and reporting the categories and curie content 

of solid wastes. 

The methods of measuring, reporting, and evaluating releases have evolved and improved over the past 30 

years. Revision 1 of Regulatory Guide 1.21 does not fully reflect current staff positions that have changed 

based on the lessons learned and operating experience gained over the past 30-plus years. 10 CFR 

50.75(g) provides that: 

“Each licensee shall keep records of ... spills or other unusual 

occurrences ... when there is reasonable likelihood that contaminants 

may have spread to inaccessible areas ... [and] ... these records must 

include any known information on identification of involved 

radionuclides, quantities, forms, and concentrations,” 

This revision of Regulatory Guide 1.21 informs licensees of staff approved methods of improved sampling 

and analysis techniques, including detection and measurements of radioactivity in spills and leaks. 

On March 10, 2006, the NRC Executive Director for Operations established the Liquid Radioactive 

Release Lessons Learned Task Force in response to incidents at some nuclear power plants related to 

unplanned, unmonitored releases of radioactive liquids into the environment. The task force issued the 

“Liquid Radioactive Release Lessons Learned Task Force Final Report,” dated September 1, 2006 (Ref. 

26), which recommended the revision of effluent and environmental monitoring program requirements and 

guidance and the provision of additional guidance on detecting, evaluating, and monitoring unplanned and 

unmonitored releases of radioactive liquids into the environment. 

2. Objective 

The objective of this regulatory action is to update the regulatory guide to describe the improved methods 

of measuring, evaluating, and reporting radioactivity in solid waste and radioactivity in liquid and gaseous 

effluents. In addition, this regulatory action would provide other editorial corrections and revisions to 

enhance clarity. 
DG-1186, Page 31

3. Alternative Approaches 

The NRC staff considered the following alternative approaches: 

• Do not revise Regulatory Guide 1.21. 

• Update Regulatory Guide 1.21. 

3.1 Alternative 1: Do Not Revise Regulatory Guide 1.21

Under this alternative, the NRC would not revise the guidance and the current guidance would be retained. 

 If NRC does not take action, there would not be any changes in costs or benefit to the public, licensees or 

NRC. However, this “no-action” alternative would not address identified concerns with the current version 

of the regulatory guide. This alternative provides a baseline condition from which any other alternatives 

will be assessed. 

3.2 Alternative 2: Update Regulatory Guide 1.21

Under this alternative, the NRC would update Regulatory Guide 1.21 to provide current staff guidance. 

The impact to the NRC would be the costs associated with preparing and issuing the regulatory guide 

revision. There will be minimal impact to licensees since the information to be supplied in accordance 

with this draft guide (with the exception of the groundwater data) was required by the previous version 

(Revision 1) of this Regulatory Guide. Since licensees have been reporting ground water information in 

Arrears since 2006 (in accordance with industry guidance, Ref. 31) there will be minimal impact on 

licensees. The impact to the public would be the voluntary costs associated with reviewing and providing 

comments to NRC during the public comment period. The value to NRC staff and users of the regulatory 

guide would be the benefits associated with enhanced efficiency and effectiveness gained by using a 

common guidance document as the technical bases for license applications and other interactions between 

the NRC and its regulated entities. There is additional benefit in updating this Regulatory Guide in that the 

revised guidance is current with new industry practices and new (or revised) analytical techniques 

developed since publication of the last revision of this Regulatory Guide. 

4. Conclusion 

Based on this regulatory analysis, the staff recommends revision of Regulatory Guide 1.21. The staff 

concludes that the proposed action will enhance compliance with NRC regulations associated with 

measuring, evaluating and reporting radioactive releases, and guidance on detecting, evaluating and 

monitoring unplanned and unmonitored releases of radioactive liquids into the environment. 
DG-1186, Page 32

GLOSSARY

A priori - Not based on prior study or examination; before the fact. In this regulatory guide, “a priori” 

describes the concept that the sensitivity (lower limit of detection) of an analytical technique 

(isotopic radiological measurement) should be determined (1) before an analysis is run and (2) 

before interference from other radionuclides occurs during actual measurements. 

Abnormal Release - An unplanned or uncontrolled release of licensed radioactive material, including 

leaks and spills, to the site environs. The site environs in this context encompasses locations 

outside of nuclear power plant systems, structures, and components (SSCs), where the SSCs are as 

described in the final safety analysis report or ODCM. Abnormal releases can occur in restricted 

areas, controlled areas, or unrestricted areas. System leakage captured by effluent ventilation 

control systems or sumps is not an abnormal release. See also the definitions of “unplanned 

release” and “uncontrolled release” below. In certain circumstances, there may be some 

subjectivity (or measures of degree) associated with the definitions of “unplanned release” and 

“uncontrolled release.” In these situations, additional circumstances should be considered before a 

final determination of abnormal release is made. Generally, releases that may reasonably be 

categorized as both unplanned and uncontrolled should be considered abnormal releases. A 

volume of gas from the containment atmosphere that escapes the equipment hatch during a RFO 

(especially during the time interval when the containment purge exhaust fans are off) would 

generally not be considered an abnormal release if (1) the duration was short, (2) the containment 

activity (gas, particulate, and iodine) was known and very low, and (3) an evaluation was 

completed to identify a limiting (or “trigger”) level of activity that would initiate 

remedial/mitigating action (e.g., close the equipment hatch to control gases escaping containment). 

Accumulated Radioactivity – Radioactivity from prior-year effluent releases that may still be present in 

the media of concern. 

Batch Release – The discharge of a discrete volume of radioactive gases or liquids occurring over a finite 

period of time (usually hours or days). 

Channel Calibration - The adjustment, as necessary, of the channel output such that it responds within 

the necessary range and accuracy to known values of the parameter that the channel monitors. See 

also the definition in NUREG-1302/1302. 

Channel Check - The qualitative assessment, by observation, of channel behavior during operation. This 

determination shall include, where possible, comparison of the channel indication and/or status to 

other indications and/or status derived from independent instrument channels measuring the same 

parameter. 

Channel Operational Test (COT) - The injection of a simulated or actual signal into the channel as close 

to the sensor as practicable to verify operability of all devices in the channel required for channel 

operability. The COT shall include adjustments, as necessary, of the required alarm, interlock, and 

trip setpoints required for channel operability such that the setpoints are within the necessary range 

and accuracy. The COT may be performed by means of any series of sequential, overlapping, or 

total channel steps. 

Continuous Release - An essentially uninterrupted release of gaseous or liquid effluent for extended 

periods during normal operation of the facility. A continuous release is the discharge of a 

nondiscrete volume of radwaste (i.e., from a volume or system that has input flow during the 

release). 
DG-1186, Page 33

Controlled Area - The licensee-defined area, outside of a restricted area but inside the site boundary, to 

which the licensee can limit access for any reason. 

Conversion Factor - A factor (e.g., microcuries per cubic centimeter per counts per minute (μCi/cc/cpm)) 

used to estimate a radioactivity concentration in an effluent based on a gross radioactivity 

measurement (e.g., cpm). 

D/Q – A dispersion parameter for estimating the dose to an individual at a specified (e.g., controlling) 

location. D/Q may be described as the downwind surface or ground concentration (D) (e.g., in 

units of microcuries per square meter (μCi/m2

)) of radioactive material at a location, divided by the 

release rate (Q) (e.g., in units of microcuries, μCi). D/Q is thus a normalized downwind surface 

concentration per unit release and can be used to determine the surface or ground radioactivity 

concentration during a measured effluent release. The units of D/Q are reciprocal square meters. 

Determination - A quantitative evaluation of the release or presence of radioactive material under a 

specific set of conditions. A determination may be made by direct or indirect measurements (e.g., 

with the use of scaling factors). 

Effluent Discharge (Radioactive) - A discharge of licensed material through a liquid or gaseous pathway 

from a facility into the site environs. 

Effluent Monitor Inoperability - A monitor is classified as inoperable if the instrument is unavailable for 

a period of time greater than 30 continuous days (in accordance with NUREG-1301 and NUREG1302, Section 3.3.3.10.b). 

Elevated Release - A gaseous effluent release made from a height that is more than twice the height of 

adjacent solid structures, or releases made from heights sufficiently above adjacent solid structures 

that building wake effects are minimal or absent. 

Ground-level Release - A gaseous effluent release made from a height that is at – or less than – the height 

of adjacent solid structures, or where the degree of plume rise is unknown or is otherwise 

insufficient to avoid building wake effects. 

Ground Water - All subsurface water, or simply water in the ground, regardless of its quality, including 

saline, brackish, or fresh water. Ground water can be moisture in the ground that is above the 

regional water table in the unsaturated or vadose zone, or ground water can be at and below the 

water table in the saturated zone. 

Impacted Areas – An area with some reasonable potential for residual radioactivity in excess of natural 

background or fallout levels (see 10 CFR 50.2, “Definitions,” and NUREG-1757, “Consolidated 

Decommissioning Guidance,” issued September 2006). Impacted areas include locations where 

radiological leaks or spills have occurred within the onsite environs (i.e., outside of the facility’s 

systems, structures, and components). (See also the definition of “significant contamination.”) 

Leach ate - Water containing contaminants that is percolating downward from a pond or lake into the 

subsurface. 

Licensed Material - Source material, special nuclear material, or byproduct material received, possessed, 

used, transferred, or disposed of under a general or specific license issued by the Commission. 
DG-1186, Page 34

Lower Limit of Detection (LLD) - The a priori detection capability for the smallest concentration of 

radioactive material in a sample that will yield a net count, above system background, that will be 

detected with 95-percent probability with only 5-percent probability of falsely concluding that a 

blank observation represents a real signal (NUREG-1301, NUREG-1302, and NUREG/CR-4007, 

“Lower Limit of Detection Definition and Elaboration of a Proposed Position for Radiological 

Effluent and Environmental Measurements,” issued September 1984 (Ref. 27). 

Member of the Public - Any individual except an individual receiving an occupational dose. This 

includes onsite personnel who are not receiving an occupational dose. 

Minimum Detectable Concentration - The smallest activity concentration measurement that is practically 

achievable with a given instrument and type of measurement procedure. It depends on factors 

involved in the survey measurement process (surface type, geometry, backscatter, and selfabsorption). (See NUREG-1507, “Minimum Detectable Concentrations with Typical Radiation 

Survey Instruments for Various Contaminants and Field Conditions,” issued June 1998) (Ref. 28). 

Minor Release Point – A specific component, system, or building which is a source of (liquid or gaseous) 

radwaste and for which an evaluation indicates (1) the dose is less than 1% of the total from all 

release points, and (2) the activity (e.g., curies) is less than 1% of the total effluent activity from all 

release points (where liquid and gas releases are considered separately, and nuclides are totaled 

separately by the categories listed in regulatory position 9.1.1 and 9.2.1, and the summing intervals 

for doses and activity are either quarterly, annually, each fuel cycle, or longer period as specified 

by the licensee). The evaluation should be documented and may be based on samples, bounding 

assessments, or a combination of both. 

Mixed Mode Release – A gaseous effluent release made from a height higher than a ground-level release, 

but less than an elevated release, where, due to lack of plume rise (e.g., buoyancy, momentum, and 

wind speed), a proper estimate of radionuclide transport and dispersion require mathematically 

splitting the plume into (1) and elevated component, and (2) a ground level component to properly 

account for building wake effects. See Regulatory Guide 1.111 (Ref. 17) for further guidance. 

Monitoring – The process of measuring and analyzing the release characteristics of radioactive material. 

Non-routine, Planned Discharge – An effluent release from a release point that is not defined in the 

ODCM but that has been planned, monitored, and discharged in accordance with 10 CFR 20.2001 

(e.g., the discharge of water recovered during a spill or leak from a temporary storage tank). 

Nuclear Fuel Cycle – The operations defined to be associated with the production of electrical power for 

public use by any fuel cycle through utilization of nuclear energy. (40 CFR 190.02) 

Operability (Operable) – The ability of a system, subsystem, train, component, or device to perform its 

specified safety function(s) and the ability of all necessary attendant instrumentation, controls, 

normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary 

equipment (required for the system, subsystem, train, component, or device to perform its specified 

safety function(s)) to perform their related support function(s). 

Principal Exposure Pathways – The primary exposure pathways to mankind (e.g., direct radiation, 

airborne exposure, waterborne exposure, and ingestion exposure pathways). 
DG-1186, Page 35

Principal Radionuclide – A radionuclide that contributes in excess of 1 percent of the dose to a member 

of the public or in excess of 1 percent of the total effluent activity in a particular reporting interval 

(e.g., a calendar year), and for the particular radionuclide categories listed in Table A-1 (for gas 

releases) and Table A-2 (for liquid releases). 

Release Point – A location from which radioactive materials are released to the environment from a 

system, structure, or component (including evaporative releases and leaching from ponds in the 

controlled or restricted area before discharge under 10 CFR 20.2001) (see also the definition for 

“significant release point”). 

Restricted Area – An area where the licensee limits access for the purpose of protecting individuals 

against undue risks from exposure to radiation and radioactive materials. 

Scaling Factor – A factor used to estimate the activity of an unknown radionuclide based on its ratio to the 

activity of a readily measured radionuclide or other parameter (e.g., carbon-14 scaled to power 

generation). 

Significant Contamination (as used for 10 CFR 50.75(g) recordkeeping) – A quantity and/or 

concentration of residual radioactivity that would require remediation during decommissioning in 

order to terminate the license by meeting the unrestricted use criteria stated in 10 CFR 20.1402, 

“Radiological Criteria for Unrestricted Use” (see NUREG-1757). 

Significant Residual Radioactivity – Synonymous with significant contamination. 

Significant Exposure Pathway – An exposure pathway that contributes in excess of 10% of the public’s 

effluent dose from all exposure pathways (see also the definition for “principal exposure 

pathway”). 

Significant Release Point – A specific component, system, or building which is a source of (liquid or 

gaseous) radwaste and for which both (1) the dose is greater than or equal to 1% of the total from 

gaseous release points or liquid release points (considered separately), and (2) the activity (e.g., 

curies) is greater than or equal to 1% of the total effluent activity from gaseous release points or 

liquid release points (considered separately). Additionally, nuclides are totaled separately by the 

categories as listed in regulatory position 9.1.1 (e.g., fission and activation gases, iodines/halogens, 

particulates, tritium, and gross alpha) and 9.2.1 (fission and activation products, tritium, dissolved 

and entrained noble gases, and gross alpha). Lastly, the summing intervals for dose and activity 

calculations are either quarterly, annually, each fuel cycle, or longer period as specified by the 

licensee. 

Site Boundary – That line beyond which the licensee does not own, lease, or otherwise control the land or 

property. 

Site Environs – Locations outside of the nuclear power plants systems, structures, or components as 

described in the final safety analysis report or the ODCM. 

Source Check – A qualitative assessment of the channel (instrument) response when the channel sensor is 

exposed to a radioactive source. 

Unlicensed Material – Byproduct material that is no longer under a general or specific license issued by 

the Commission. This includes radioactive material in the site environs that has been disposed of 
DG-1186, Page 36

in accordance with 10 CFR 20.2001, unless concentrations exceed the limits in 10 CFR 30.14, 

“Exempt Concentration” (Ref. 28). In addition, exempt radioactive sources under 10 CFR 30.15, 

“Certain Items Containing Byproduct Material,” or 10 CFR 30.18, “Exempt Quantities,” are 

unlicensed material. 

Uncontrolled Release – The release of licensed radioactive material in which (1) the release path is not 

monitored (e.g., with a continuously indicating/recording radiation monitor in the discharge 

pathway, or with a flow indicator/recorder in the discharge pathway, or without benefit of grab 

samples), AND (2) there is reasonable potential that the combination of flow rate and release 

activity could result in significant amounts of radioactive material discharged, AND (3) there is 

not a preplanned method of terminating the release to the environment. 

Unplanned Release – The unintended discharge of a volume of liquid or airborne radioactivity to the 

environment. Examples of unplanned releases include 

(1) the unintentional discharge of a wrong waste gas decay tank (or bulk liquid radwaste tank), 

(2) the failure of a radiation monitor to divert liquid to the radwaste system in the case where 

radioactivity is present and the automatic alarm/trip function fails to divert material to liquid 

radwaste and that material (or a portion of that material) is instead discharged to the environment, 

or 

(3) the release of a sufficiently large amount of radioactive material (via a leak or a plant 

operational occurrence) such that a 10 CFR 50.72 or a 10 CFR 50.73 report is required. 

If a licensee has prepared a preliminary release permit in advance of a discharge, it is a planned 

release. If a preliminary release permit was not prepared (or considered/evaluated) prior to the 

discharge, additional information (as described below) should be considered before categorizing 

the release as an "unplanned" release. The magnitude of the release should be considered before 

categorizing a discharge as “unplanned.” For example, for routine small liquid leaks from pipes 

and valves where the liquid leakage is collected in sumps or drains (as designed), and where the 

discharge is monitored in accordance with the licensee’s radioactive effluent control program (as 

defined in the ODCM), the release is not unplanned. Similarly, for minor leaks (e.g., valves, 

pipes, or routine plant evolutions) that result in some radioactive gases escaping to the plant 

ventilation system, where the gases do not cause the installed plant vent stack radiation monitoring 

system to deviate significantly from the routine variance in the baseline, and where accountability 

is controlled and reported in accordance with the licensee’s radioactive effluent control program 

(as defined in the ODCM), such routine, typical releases are not considered unplanned releases. 

(Ref. 30). 

Unrestricted Area – An area for which the licensee neither controls nor limits access. 

Uranium Fuel Cycle – The operations of milling of uranium ore, chemical conversion of uranium, 

isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a lightwater-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to 

the extent that these directly support the production of electrical power for public use utilizing 

nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation 

of any radioactive material in support of these operations, and the reuse of recovered non-uranium 

special nuclear and by-product materials from the cycle. 

χ/Q - A mathematical model depicting the atmospheric transport and diffusion of radioactive material in 

gaseous effluents at some distance from the source (origin). χ/Q is the average atmospheric 

effluent concentration, χ, normalized by source strength, Q, at a distance (or location) in a given 

downwind direction. Expressed in another way, χ/Q is the concentration (χ) of airborne 
DG-1186, Page 37

radioactive material (e.g., in units of μCi/m3

) divided by the release rate (Q) (e.g., in units of μCi/s) 

at a specified distance and direction downwind of the release point. χ/Q is thus a normalized 

airborne radioactivity concentration per unit release rate and as such is readily suitable for use in 

determining the ground level airborne radioactivity concentration, at some distance from the origin 

of the release, simply by multiplying by a known source release rate Q. 
DG-1186, Page 38

REFERENCES 

1. 10 CFR Part 20, “Standards for Protection Against Radiation,” U.S. Nuclear Regulatory 

Commission, Washington, DC. 

2. 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” U.S. Nuclear 

Regulatory Commission, Washington, DC. 

3. Regulatory Guide 1.21, “Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and 

Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled 

Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, Washington, DC. 

4. Regulatory Guide 4.1, “Programs for Monitoring Radioactivity in the Environs of Nuclear Power 

Plants,” U.S. Nuclear Regulatory Commission, Washington, DC. 

5. Regulatory Guide 4.15, “Quality Assurance for Radiological Monitoring Programs (Inception 

through Normal Operations to License Termination)—Effluent Streams and the Environment,” 

U.S. Nuclear Regulatory Commission, Washington, DC. 

6. NUREG-1301, “Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 

Controls for Pressurized Water Reactors,” U.S. Nuclear Regulatory Commission, Washington, 

DC, April 1991. 

7. NUREG-1302, “Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 

Controls for Boiling Water Reactors,” U.S. Nuclear Regulatory Commission, Washington, DC, 

April 1991. 

8. EPRI Report No. 101173, “Ground Water Monitoring Guidance for Nuclear Power Plants,” 

Electric Power Research Institute, Palo Alto, CA, September 2005. 

9. NCRP Report No. 81,”Carbon-14 in the Environment,” National Council on Radiation Protection 

and Measurements, Bethesda, MD, January 1985. 

10. NUREG-1757, “Consolidated Decommissioning Guidance,” U.S. Nuclear Regulatory 

Commission, Washington, DC, September 2006. 

11. SECY-03-0069, “Results of the License Termination Rule Analysis,” U.S. Nuclear Regulatory 

Commission, Washington, DC, May 2, 2003. 

12. NUREG/BR-0308, “Effective Risk Communication,” U.S. Nuclear Regulatory Commission, 

Washington, DC, January 2004. 

13. ASTM D 3370-07, “Standard Practices for Sampling Water from Closed Conduits,” American 

Society for Testing and Materials, West Conshohocken, PA, 2007.

14. ANSI N42.18-2004, Specification and Performance of On-Site Instrumentation for Continuously 

Monitoring Radioactivity in Effluents,” American National Standards Institute, New York, NY, 

January 2004. 
DG-1186, Page 39

15. ANSI/HPS N13.1-1999, “Sampling and Monitoring Releases of Airborne Radioactive Substances 

from the Stacks and Ducts of Nuclear Facilities,” American National Standards Institute, New 

York, NY, January 1999. 

16. Regulatory Guide 1.23, “Meteorological Monitoring Programs for Nuclear Power Plants,” U.S. 

Nuclear Regulatory Commission, Washington, DC. 

17. Regulatory Guide 1.111, “Methods for Estimating Atmospheric Transport and Dispersion of 

Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,” U.S. Nuclear 

Regulatory Commission, Washington, DC. 

18. Regulatory Guide 1.113, “Estimating Aquatic Dispersion of Effluents from Accidental and 

Routine Reactor Releases for the Purpose of Implementing Appendix I,” U.S. Nuclear Regulatory 

Commission, Washington, DC. 

19. ANSI/ANS 2.17-2008, “Evaluation of Subsurface Radionuclide Transport at Commercial Nuclear 

Power Production Facilities,” American National Standards Institute, New York, NY (draft 2008). 

20. NUREG/CR-6948, “Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities 

and Sites: Logic, Strategic Approach and Discussion,” U.S. Nuclear Regulatory Commission, 

Washington, DC, November 2007. 

21. NUREG-1576, “Multi-Agency Radiological Laboratory Analytical Protocols Manual,” U.S. 

Nuclear Regulatory Commission, Washington, DC, July 2004. 

22. 40 CFR Part 190, “Environmental Radiation Protection Standards for Nuclear Power Operations,” 

U.S. Environmental Protection Agency, Washington, DC. 

23. Regulatory Guide 1.109, “Calculation of Annual Doses to Man from Routine Releases of Reactor 

Effluents for the Purpose of Demonstrating Compliance with 10 CFR Part 50, Appendix I,” U.S. 

Nuclear Regulatory Commission, Washington, DC. 

24. 10 CFR Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 

High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste,” U.S. Nuclear 

Regulatory Commission, Washington, DC. 

25. ICRP Publication 71, “Age-Dependent Doses to Members of the Public from Intake of 

Radionuclides: Part 4, Inhalation Dose Coefficients,” International Commission on Radiological 

Protection, Stockholm, Sweden. 1995. 

26. Liquid Radioactive Release Lessons Learned Task Force Final Report, U.S. Nuclear Regulatory 

Commission, Washington, DC, September 1, 2006 (ADAMS Accession No. ML062650312). 

27. NUREG/CR-4007, “Lower Limit of Detection: Definition and Elaboration of a Proposed Position 

for Radiological Effluent and Environmental Measurements,” U.S. Nuclear Regulatory 

Commission, Washington, DC, September 1984. 

28. NUREG-1507, “Minimum Detectable Concentrations with Typical Radiation Survey Instruments 

for Various Contaminants and Field Conditions,” U.S. Nuclear Regulatory Commission, 

Washington, DC, June 1998. 
DG-1186, Page 40

29. 10 CFR Part 30, “Rules of General Applicability to Domestic Licensing of Byproduct Material,” 

U.S. Nuclear Regulatory Commission, Washington, DC. 

30. NUREG/CR-5569, “Health Physics Positions Data Base,” HPPOS-254, “Definition of Unplanned 

Release,” U.S. Nuclear Regulatory Commission, Washington, DC. 

31. NEI 07-07, “Industry Ground Water Protection Initiative – Final Guidance Document,” Nuclear 

Energy Institute, August 2007 

32. EPRI Report No. 1015118, “Ground Water Protection Guidelines for Nuclear Power Plants,” 

Electric Power Research Institute, Palo Alto, CA, November 2007. 

33. Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental 

Surveillance, January 1991, U.S. Department of Energy, DOEBB-0173T, DE 91 013607 

34. NUREG/CR-6204, “Questions and Answers Based on Revised 10 CFR Part 20, U.S. Nuclear 

Regulatory Commission, Washington, DC, May 1994. 

35. NUREG-1022, “Event Reporting Guidelines 10 CFR 50.72 and 50.73.” U.S. Nuclear Regulatory 

Commission, Washington, DC, October 2000. 

36. NRC IE Bulletin No. 80-10, “Contamination of Nonradioactive System and Resulting Potential for 

Unmonitored, Uncontrolled Release of Radioactivity to Environment," U.S. Nuclear Regulatory 

Commission, Washington, DC, May 6, 1980. 
DG-1186, Page 41

BIBLIOGRAPHY

40 CFR Part 141, “National Primary Drinking Water Regulations,” U.S. Environmental Protection 

Agency, Washington, DC. 

ANSI N13.30-1996, “Performance Criteria for Radiobioassay,” American National Standards Institute, 

New York, NY, May, 1996. 

 

ANSI/ANS 3.11-2005, “Determining Meteorological Information at Nuclear Facilities,” American 

National Standards Institute, New York, NY, January 2005. 

ANSI N42.14-1999, “Calibration and Use of Germanium Spectrometers for the Measurement of GammaRay Emission Rates of Radionuclides,” American National Standards Institute, New York, NY, May 1999. 

ANSI/NCSL Z540-2-1997 (reapproved 2002), “American National Standard for Expressing Uncertainty--

U.S. Guide to the Expression of Uncertainty in Measurement,” American National Standards Institute, 

New York, NY, January 1997 

NIST Technical Note 1297, “Guidelines for Evaluating and Expressing the Uncertainty of NIST 

Measurement Results,” National Institute of Standards and Technology, Gaithersburg, MD, September 

1994.

NUREG-800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power 

Plants,” March 2007 (Section 2.3.5), U.S. Nuclear Regulatory Commission, Washington, D.C., May 

1980. 

NUREG-0324, “XOQDOQ, Program for the Meteorological Evaluation of Routine Effluent Releases at 

Nuclear Power Stations,” U.S. Nuclear Regulatory Commission, Washington, DC, September 1977. 

NUREG-0543, “Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle 

Standard (40 CFR Part 190),” February 1980. 

NUREG/CR-2919, “XOQDOQ Computer Program for the Meteorological Evaluation of Routine Effluent 

Releases at Nuclear Power Stations,” U.S. Nuclear Regulatory Commission, Washington, DC, September, 

1982 

Regulatory Guide 1.143, Revision 2, “Design Guidance for Radioactive Waste Management Systems, 

Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants,” U.S. Nuclear 

Regulatory Commission, Washington, DC, November 2001. 
Appendix A to DG-1186, Page A-1

APPENDIX A 

Table A-1, Gaseous Effluents - Summation of All Releases 

Total Release Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Fission and 

Activation Gases 

Ci 

Iodines Ci 

Particulates Ci 

Tritium Ci 

Alpha Ci 
Appendix A to DG-1186, Page A-2

Table A-1A, Gaseous Effluents - Ground-Level Release - Batch Mode 

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 Ci 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Other Ci 

 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-3

Table A-1B, Gaseous Effluents - Ground-Level Release - Continuous Mode

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 Ci 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-4

Table A-1C, Gaseous Effluents - Elevated Release - Batch Mode 

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 Ci 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Other Ci 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-5

Table A-1D, Gaseous Effluents - Elevated Release - Continuous Mode 

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Other Ci 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-6

Table A-1E, Gaseous Effluents - Mixed Mode Release - Batch Mode 

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Other Ci 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-7

Table A-1F, Gaseous Effluents - Mixed Mode Release - Continuous Mode 

Fission Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

C-14 Ci 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131M Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Xe-138 Ci 

Other Ci 

Total Ci 

Iodines/halogens Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

I-131 Ci 

I-132 Ci 

I-133 Ci 

I-134 Ci 

I-135 Ci 

Total Ci 

Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Zn-72 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Nb-95m Ci 

Tc-99m Ci 

Te-132 Ci 

Cs-134 Ci 

Cs-138 Ci 

Ba-139 Ci 

La-142 Ci 

Ce-143 Ci 

Ce-144 Ci 

Rh-188 Ci 

Other Ci 

Total Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-8

Table A-2, Liquid Effluents 

Summation of All Releases 

Total Release Units Quarter 

1 

Quarter 

2 

Quarter 

3 

Quarter 

4 

Total Uncertainty

(%) 

Fission and 

Activation 

Products 

(excluding 

tritium, gases, 

and alpha) 

Ci 

Tritium Ci 

Dissolved and 

Entrained 

Gases 

Ci 

Alpha Ci 

Volume of 

Primary System 

Liquid Effluent 

(Before Dilution)

Liters 

Volume of 

Secondary or 

Balance of Plant 

Liquid Effluent 

(e.g., LowActivity or 

unprocessed) 

(Before Dilution) 

Liters 

Dilution Water 

Used 

Liters 

Average Stream 

Flow 

m3

/sec 
Appendix A to DG-1186, Page A-9

Table A-2A, Liquid Effluents - Batch Mode 

Fission and 

Activation 

Products 

Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Cr-51 Ci 

Mn-54 Ci 

Fe-55 Ci 

Fe-59 Ci 

Co-57 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Y-91m Ci 

Y-92 Ci 

Nb-95 Ci 

Nb-95m Ci 

Mo-99 Ci 

Tc-99m Ci 

Ru-103 Ci 

Ru-106 Ci 

Ag-110m Ci 

Sn-113 Ci 

Sb-122 Ci 

Sb-124 Ci 

Sb-125 Ci 

Te-129 Ci 

Te-132 Ci 

I-131 Ci 

I-133 Ci 

I-135 Ci 

Cs-134 Ci 

Cs-137 Ci 

Ba-139 Ci 

Ba-140 Ci 

La-140 Ci 

Ce-143 Ci 

Cs-144 Ci 

W-187 Ci 

Np-239 Ci 

Other Ci 

Totals Ci 
Appendix A to DG-1186, Page A-10

Table A-2A (continued), Liquid Effluents - Batch Mode 

Dissolved and 

Entrained Gases 

Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131m Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Other Ci 

Totals Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-11

Table A-2B, Liquid Effluents - Continuous Mode 

Particulates and 

Iodines 

Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Cr-51 Ci 

Mn-54 Ci 

Fe-55 Ci 

Fe-59 Ci 

Co-57 Ci 

Co-58 Ci 

Co-60 Ci 

Sr-89 Ci 

Sr-90 Ci 

Y-91m Ci 

Y-92 Ci 

Nb-95 Ci 

Nb-95m Ci 

Mo-99 Ci 

Tc-99m Ci 

Ru-103 Ci 

Ru-106 Ci 

Ag-110m Ci 

Sn-113 Ci 

Sb-122 Ci 

Sb-124 Ci 

Sb-125 Ci 

Te-129 Ci 

Te-132 Ci 

I-131 Ci 

I-133 Ci 

I-135 Ci 

Cs-134 Ci 

Cs-137 Ci 

Ba-139 Ci 

Ba-140 Ci 

La-140 Ci 

Ce-143 Ci 

Cs-144 Ci 

W-187 Ci 

Np-239 Ci 

Other Ci 

Totals Ci 
Appendix A to DG-1186, Page A-12

Table A-2B (continued), Liquid Effluents - Continuous Mode 

Dissolved and 

Entrained Gases 

Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total 

Ar-41 Ci 

Kr-85 Ci 

Kr-85m Ci 

Kr-87 Ci 

Kr-88 Ci 

Xe-131m Ci 

Xe-133 Ci 

Xe-133m Ci 

Xe-135 Ci 

Xe-135m Ci 

Other Ci 

Totals Ci 

Tritium Ci 

Gross Alpha Ci 
Appendix A to DG-1186, Page A-13

Table A-3, Low-Level Waste and Used Fuel 

1 LLW Shipped for Direct Disposal (to 

licensed waste disposal facilities) 

Number of Shipments Volume Activity Principal 

Isotopes 

Class A Resins, Filters, Evap bottoms ft3 Ci 

Class B Resins, Filters, Evap bottoms ft3 Ci 

Class A Dry Active Waste ft3 Ci 

Class A Irradiated Components ft3 Ci 

Class A Other Waste (e.g., sewage sludge, 

ventilation charcoal) ft3 Ci 

Other Waste (not fitting in above 

categories) 

ft3 Ci 

2 LLW Shipped for Processing (before 

disposal or return to site) 

Number of Shipments Volume Activity Principal 

Isotopes 

Class A Resins, Filters, Evap bottoms ft3 Ci 

Class B Resins, Filters, Evap bottoms ft3 Ci 

Class A Dry Active Waste ft3 Ci 

Class A Irradiated Components ft3 Ci 

Class A Other Waste (e.g., sewage sludge, 

ventilation charcoal) ft3 Ci 

Other Waste (not fitting in above 

categories) 

ft3 Ci 

3 LLW Returned to Site 

(from radioactive waste processors for 

long-term storage) 

Number of Shipments Volume Activity Principal 

Isotopes 

Class A Resins, Filters, Evap bottoms ft3 Ci 

Class B Resins, Filters, Evap bottoms ft3 Ci 

Other Waste 

ft3 Ci 

4 LLW Shipped for 10 CFR 20.2002 

Disposal 

Number of Shipments Volume Activity Principal 

Isotopes

Class A Resins, Filters, Evap bottoms ft3 Ci 

Class A Dry Active Waste ft3 Ci 

Class A Irradiated Components ft3 Ci 

Class A Other Waste (e.g., sewage sludge, 

ventilation charcoal) ft3 Ci 

Other Waste (not fitting above categories) 

ft3 Ci 

5 Used (Irradiated) Fuel Shipments 

(Offsite Disposal)

Number of Shipments Destination Mode of 

Transport 

Principal 

Isotopes

 
Appendix A to DG-1186, Page A-14

Table A-4, Dose Assessments, 10 CFR Part 50, Appendix I 

Quarter 1 Quarter 2 Quarter 

3 

Quarter 

4 

Yearly Sector/ 

Distance 

Liquid Effluent 

Dose Limit 

Total Body 1.5 mrem 1.5 mrem 1.5 mrem 1.5 mrem 3 mrem 

Total Body Dose 

% of Limit 

Liquid Effluent 

Dose Limit 

Any Organ 5 mrem 5 mrem 5 mrem 5 mrem 10 mrem 

Organ Dose 

% of Limit 

Gaseous 

Effluent 

Dose Limit 

Gamma Air 

Dose 

5 mrad 5 mrad 5 mrad 5 mrad 10 mrad 

Gamma 

Air Dose 

 

% of Limit 

Gaseous 

Effluent 

Dose Limit 

Beta Air Dose 

10 mrad 10 mrad 10 mrad 10 mrad 20 mrad 

Beta Air Dose 

% of Limit 

Gaseous 

Effluent 

Dose Limit 

Any Organ 

(Iodine, Tritium, 

Particulates with >8-

day half-life) 

7.5 mrem 7.5 mrem 7.5 mrem 7.5 mrem 15 mrem 

Gaseous 

Effluent 

Organ Dose 

(Iodine, Tritium, 

Particulates with > 8-

Day half-life) 

 

% of Limit 
Appendix A to DG-1186, Page A-15

Table A-5, Dose to a Member of the Public Per 10 CFR 20.1301 (a) thru (c)

Liquid 

Effluents 

(mrem) 

Gaseous 

Effluents 

(mrem) 

Direct 

Radiation 

(mrem) 

Total 

(mrem) 

Location/ 

Area 

(i.e., controlled area 

or unrestricted area) 

Sector/ 

Distance 

TEDE 

Limit 

100 mrem controlled area or 

unrestricted area 

TEDE 

Dose 

 

% of 

Limit 

Table A-6, EPA 40 CFR Part 190 Individual in the Unrestricted Area 

Whole 

Body 

Thyroid Any other organ Sector/Distance 

Dose Limit 25 mrem 75 mrem 25 mrem 

Dose 

% of Limit 
Appendix A to DG-1186, Page A-16

Table A-7, Supplemental Information 

1. Abnormal releases (e.g., leaks and spills)

2. Nonroutine, Planned Discharges (e.g., pumping of leaks and spills for remediation, results of 

ground water monitoring to quantify effluent releases to the offsite environment)

3. Radioactive Waste Treatment System Changes

4. Annual Land-Use Census Changes

5. Effluent Monitor Instrument Inoperability

6. Offsite Dose Calculation Manual Changes 

7. Process Control Program Changes

8. Errata/Supplemental Report(s) (e.g., errors or omissions in previous reports)

9. Other (narrative description of other information that is provided to the U.S. Nuclear Regulatory 

Commission)
Appendix A to DG-1186, Page A-17

Table A-8, Residual Radioactivity Screening Values for Surface Soil 

(Partial List of Common Nuclear Power Plant Radionuclides) 

(from Appendix H to NUREG-1757, Volume 2, Revision 1) 

Radionuclide Symbol pCi/g pCi/kg 

Hydrogen-3 3H 110 110,000 

Carbon-14 14C 12 12,000 

Manganese-54 54Mn 15 15,000 

Iron-55 55Fe 10,000 10,000,000 

Cobalt-60 60Co 3.8 3,800 

Nickel-59 59Ni 5,500 5,500,000 

Nickel-63 63Ni 2,100 2,100,000 

Strontium-90 90Sr 1.7 1,700 

Niobium-94 94Nb 5.8 5,800 

Technetium-99 99Tc 19 19,000 

Cesium-134 134Cs 5.7 5,700 

Cesium-137 137Cs 11 11,000 

Notes: 

These values represent surficial surface soil concentrations of individual radionuclides that would be deemed in 

compliance with the 25 millirem/year (0.25 millisievert/year) unrestricted release dose limit in 10 CFR 20.1402, 

“Radiological Criteria for Unrestricted Use.” For radionuclides in a mixture, the “sum of fractions” rule applies (see 

Section 2.7 of NUREG-1757). 

Screening values are in units of picocuries per gram equivalent to 25 millirem/year. These values were derived using 

Decontamination and Decommissioning (D&D) screening methodology (NUREG/CR-5512, Volume 3, “Residual 

Radiation Contamination from Decommissioning: Parameter Analysis,” issued October 1999). They were derived 

based on selection of the 90th percentile of the output dose distribution for each specific radionuclide (or 

radionuclide with the specific decay chain). Behavioral parameters were set at the mean of the distribution of the 

assumed critical group. The metabolic parameters were set at “Reference Man” or at the mean of the distribution for 

an average human.