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{{Adams
{{Adams
| number = ML091170109
| number = ML003739960
| issue date = 06/30/2009
| issue date = 06/30/1974
| title = Measuring, Evaluating, and Reporting Radioactive Materials in Liquid and Gaseous Effluents and Solid Waste
| title = Measuring,Evaluating & Reporting Radioactivity in Solid Wastes & Releases of Radioactive Materials in Liquid & Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants
| author name =  
| author name =  
| author affiliation = NRC/RES
| author affiliation = NRC/RES
Line 9: Line 9:
| docket =  
| docket =  
| license number =  
| license number =  
| contact person = O'Donnell, Edward, RES/RGB
| contact person =  
| case reference number = DG-1186
| document report number = RG-1.21, Rev 1
| document report number = RG-1.021, Rev. 2
| package number = ML091170100
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 72
| page count = 22
}}
}}
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION
{{#Wiki_filter:U.S. ATOMIC ENERGY COMMISSION  
June 2009
REGULATORY  
 
DIRECTORATE OF REGULATORY STANDARDS
Revision 2
Revision 1 June 1974 GUIDE
 
REGULATORY GUIDE 1.21 MEASURING, EVALUATING, AND REPORTING RADIOACTIVITY IN
REGULATORY GUIDE
SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID  
OFFICE OF NUCLEAR REGULATORY RESEARCH
AND GASEOUS EFFLUENTS FROM LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
 
The NRC issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff needs in reviewing applications for permits and licenses.  Regulatory guides are not substitutes for regulations, and compliance with them is not required.  Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.
 
This guide was issued after consideration of comments received from the public.
 
Regulatory guides are issued in 10 broad divisions:  1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health;
9, Antitrust and Financial Review; and 10, General.
 
Electronic copies of this guide and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc- collections/reg-guides/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML091170109
 
REGULATORY GUIDE 1.21 (Draft was issued as DG-1186, dated October 2008)
 
MEASURING, EVALUATING, AND REPORTING RADIOACTIVE  
MATERIAL IN LIQUID AND GASEOUS EFFLUENTS AND
SOLID WASTE


==A. INTRODUCTION==
==A. INTRODUCTION==
This guide describes methods the staff of the U.S. Nuclear Regulatory Commission (NRC)
General Design Criterion 60, "Control of releases of radioactive materials to the environment,"
considers acceptable for use:  (1) in measuring, evaluating, and reporting plant-related radioactivity (excluding background radiation) in effluents and solid radioactive waste shipments from NRC licensed facilities, (2) in assessing and reporting the public dose from facility operations, and (3) on complying with
of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10CFR Part 50, "Licensing of Production and Utilization Facilities," requires that the nuclear power plant design include means to control the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactcr operation, including anticipated operational occurrences.
40 CFR 190 in accordance with the requirements of 10 CFR 20.1301(e). 
 
This guide incorporates the risk-informed principles of the Reactor Oversight Process.  A risk- informed, performance-based approach to regulatory decision-making combines the risk-informed and performance-based elements discussed in the staff requirements memorandum on SECY-98-144, White Paper on Risk-Informed and Performance-Based Regulation, dated March 1, 1999 (Ref. 1). 
 
The following regulations and design criteria establish the regulatory basis for the radiological effluent control program: 
 
1.
 
Title 10 of the Code of Federal Regulations (10 CFR) Section 20.1501, Surveys (Ref. 2), 
 
Rev. 2 of RG 1.21, Page 2
2.
 
10 CFR 50.36a, Technical Specifications on Effluents from Nuclear Power Reactors (Ref. 3),
 
3.
 
10 CFR 20.1302, Compliance with Dose Limits for Individual Members of the Public, 
 
4.
 
10 CFR 72.44(d), License Conditions (Ref. 4), 


5.
General Design Criterion
64,
"Monitoring radioactivity releases,"
requires that nuclear powver plant designs provide means for monitoring effluent discharge paths for radioactivity that may be released
_-
from normal operations, including anticipated operational occurrences, and from postulated accidents.


Section IV.B of 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, to
Section 20.106, "Concentrations in effluents to unrestricted areas," of 10 CFR Part 20, "Standards for Protection Against Radiation," provides that a licensee shall not release to an unrestricted area, radioactive materials in concentrations which exceed limits specified in 10 CFR Part 20 or as otherwise authorized in a license issued by the Commission. Section 20.201,  
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.
"Surveys," of 10 CFR Part 20 further requires that a licensee conduct surveys of concentrations of radioactive materials as necessary to demonstrate compliance with AEC regulations.


6.
Paragraph (a)(2)
of
§50.36a,
"Technical specifications on effluents from nuclear power reactors,"
of 10 CFR Part 50 provides that technical specifications for each license will include a requirement that the licensee submit a report to the Commission within 60
days after January 1 and July 1 of each year which specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents during the previous 6 months of operation, and such other information as may be required by the Commission to estimate maximum potential annual radiation doses to the public resulting from effluent releases.


General Design Criterion 60, Control of releases of radioactive materials to the environment, of Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), Domestic Licensing of Production and Utilization Facilities.
Paragraph (c) of §20.1, "Purpose," of 10 CFR Part
20 states that every reasonable effort should be made by AEC licensees to maintain radiation exposure, and releases of radioactive materials in effluents to unrestricted areas, as far below the limits specified in Part 20 as practicable, i.e., as low as is practicably achievable, taking into account the state of technology, and the economics of improvements in relation to benefits to the public health and safety and in relation to the utilization of atomic energy in the public interest.


7.
This guide describes programs acceptable to the Regulatory staff for measuring, reporting, and evaluating releases of radioactive materials in liquid and gaseous effluents and guidelines for classifying and reporting the categories and curie content of solid wastes. Other programs for the reporting of operating information, including abnormal occurrences, are presented in Regulatory Guide
1.16,
"Reporting of Operating Information." In some cases, specific programs should be supplemented because of individual plant design features or other factors. The need for supplemental or modified programs will be determined on a case-by-case basis.


General Design Criterion 64, Monitoring radioactivity releases, of Appendix A,
The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.
General Design Criteria for Nuclear Power Plants, to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), Domestic Licensing of Production and Utilization Facilities.


10 CFR 20.1501 requires surveys that may be necessary and are reasonable to evaluate the magnitude and extent of potential radiological hazards.  In 10 CFR Part 20, Standards for Protection against Radiation, survey is defined as an evaluation of the radiological conditions and potential hazards related to radioactive material or other sources of radiation, including (1) a physical survey of the location of radioactive material and (2) measurements or calculations of levels of radiation or concentrations or quantities of radioactive material present.  The design objectives set out in
USAEC REGULATORY GUIDES
10 CFR Part 50, Appendix I, provide numerical guidance on limiting conditions for operation for light- water cooled nuclear power reactors to meet the requirement that radioactive materials in effluents discharged to unrestricted areas be kept as low as is reasonably achievable (ALARA).  
Copies of published guld.


10 CFR 50.36a requires establishing technical specifications with procedures and controls over effluents, including reporting (1) the quantity of each of the principal radionuclides discharged 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.
may be obtained by request indicating the divslions desired to the U.S. Atomic Energy Commission, Washington, D.C. 20545, Regulatory Guides we issued to describe and make available to the public Attention: Director of Regulatory Standarde. Comments end euggestions for methods acceptable to the AEC Regulatory staff of implementing specific parts of Improvements In these guides are encouraged and should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commission, U.S. Atomic Energy Commission, Washington, D.C. 20545, weluating specfic problems or postulated accidents, or to provide guidance to Attention: Chief, Public Proceedings Staff.


In 10 CFR 20.1302, the NRC establishes requirements for surveys in the unrestricted and controlled areas and for radioactive materials in effluents discharged to unrestricted and controlled areas.
applicants. Regulatory Guides are not substitutes for regulations and compliance with them is not required. Methods and solutions different from those set out in The guides are issued in the following ten broad divisions:
the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission.


The purpose of these surveys is to demonstrate compliance with the dose limits of 10 CFR 20.1301, Dose Limits for Individual Members of 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 10 CFR 20.1302(b)(1) to determine the total effective dose equivalent to the individual likely to receive the highest dose.  This requirement is based on actual, realistic exposure pathways to a real individual.  (See also 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.  5) and Attachment 6 to SECY-03-0069, Results of the License Termination Rule Analysis, dated May 2, 2003 (Ref.  6)). 
===1. Power Reactors ===


Rev. 2 of RG 1.21, Page 3 In 10 CFR 72.44(d), the NRC establishes environmental monitoring requirements for each facility holding a specific license under Part 72 authorizing receipt, handling, and storage of spent fuel, high-level radioactive waste, and/or reactor-related greater than class C waste.  This regulatory guide describes a method for reporting these results.
===6. Products ===
2. Research and Test Reactors


The general design criteria, Criterion 60, specifies nuclear power units shall control liquid and gaseous effluents and handle solid waste for both normal and anticipated operational occurrences.
===7. Transportation ===
3. Fuels and Materials Facilities
8. Occupational Health Published guides will be revised periodically, as appropriate, to accommodate
4. Environmental and Siting
9. Antitrust Review comments ard to reflect new information or experience.


The general design criteria, Criterion 64, specifies that a means shall be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released during both normal and anticipated operational occurrences.
5. Materials and Plant Protection
1


The reports required under (1) Subpart M, Reports, of 10 CFR Part 20 (related to reports of exposures, radiation levels, and concentrations or radioactive material), (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 these applicable regulations.  In addition, effluent discharges and radioactive material losses reported under those regulatory provisions should also be reported in the Annual Radioactive Effluent Release Report (ARERR) described in this regulatory guide.
===0. General===
 
This regulatory guide contains information collection requirements 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.
 
Rev. 2 of RG 1.21, Page 4 TABLE OF CONTENTS
 
==A. INTRODUCTION==
..................................................................................................................... 1
 
==B. DISCUSSION==
............................................................................................................................ 6
1.  Regulatory Guidance .............................................................................................................. 6
2.  Objectives of the Radiological Effluent Control Program ..................................................... 7
 
==C. REGULATORY POSITION==
..................................................................................................... 9
1.  Effluent Monitoring ................................................................................................................ 9
1.1  Guidance for Effluent Monitoring .................................................................................... 9
1.2  Release Points for Effluent Monitoring ............................................................................ 9
1.3  Monitoring a Significant Release Point .......................................................................... 10
1.4  Monitoring a Less-Significant Release Point ................................................................. 10
1.5  Monitoring Leaks and Spills ........................................................................................... 11
1.6  Monitoring Continuous Releases .................................................................................... 13
1.7  Monitoring Batch Releases ............................................................................................. 14
1.8  Principal Radionuclides for Effluent Monitoring ........................................................... 14
1.9  Carbon-14 ....................................................................................................................... 15
1.10  Abnormal Releases ....................................................................................................... 16
2.  Effluent Sampling ................................................................................................................. 17
2.1  Representative Sampling ................................................................................................ 17
2.2  Sampling Liquid Radioactive Waste .............................................................................. 18
2.3  Sampling Gaseous Radioactive Waste ........................................................................... 18
2.4  Sampling Bias ................................................................................................................. 18
2.5  Composite Sampling....................................................................................................... 19
2.6  Sample Preparation and Preservation ............................................................................. 19
2.7  Short-Lived Nuclides and Decay Corrections ................................................................ 19
3  Effluent Dispersion (Meteorology and Hydrology) ............................................................... 19
3.1  Meteorological Data ....................................................................................................... 19
3.2  Atmospheric Transport and Diffusion ............................................................................ 20
3.3  Release Height ................................................................................................................ 20
3.4  Aquatic Dispersion (Surface Waters) ............................................................................. 20
3.5  Spills and Leaks to the Ground Surface ......................................................................... 21
3.6  Spills and Leaks to Ground Water .................................................................................. 21
4.  Quality Assurance ................................................................................................................. 23
4.1  Regulatory Guidance ...................................................................................................... 23
4.2  Quality Control Checks .................................................................................................. 24
4.3  Functional Checks .......................................................................................................... 24
4.4  Procedures ...................................................................................................................... 24
4.5  Calibration of Laboratory Equipment and Radiation Monitors ...................................... 24
4.6  Calibration of Measuring and Test Equipment ............................................................... 25
4.7  Calibration Frequency..................................................................................................... 25
4.8  Measurement Uncertainty ............................................................................................... 25
5.  Dose Assessments for Members of the Public ..................................................................... 25
5.1  Bounding Dose Assessments .......................................................................................... 26
 
Rev. 2 of RG 1.21, Page 5
5.2  Members of the Public .................................................................................................... 27
5.3  Occupancy Factors .......................................................................................................... 27
5.4  10 CFR Part 50, Appendix I ........................................................................................... 27
5.5  10 CFR 20.1301(a) through (c) ...................................................................................... 28
5.6  10 CFR 20.1301(e) ......................................................................................................... 28
5.7  Dose Assessments for 10 CFR Part 50, Appendix I ....................................................... 29
5.8  Dose Assessments for 10 CFR 20.1301(e) ..................................................................... 30
5.9  Dose Calculations ........................................................................................................... 31
6.  Solid Radioactive Waste Shipped for Processing or Disposal ............................................. 31
7.  Reporting Errata in Effluent Release Reports ...................................................................... 32
7.1  Examples of Small Errors ............................................................................................... 32
7.2  Reporting Small Errors ................................................................................................... 32
7.3  Examples of Large Errors ............................................................................................... 33
7.4  Reporting Large Errors ................................................................................................... 33
8.  Format and Content of the Annual Radioactive Effluent Release Report ............................ 33
8.1  Gaseous Effluent ............................................................................................................. 34
8.2  Liquid Effluents .............................................................................................................. 36
8.3  Solid Waste Storage and Shipments ............................................................................... 37
8.4  Dose Assessments ........................................................................................................... 37
8.5  Supplemental Information .............................................................................................. 38
 
==D. IMPLEMENTATION==
.............................................................................................................. 40
GLOSSARY .................................................................................................................................. 41 REFERENCES .............................................................................................................................. 50
BIBLIOGRAPHY .......................................................................................................................... 54 APPENDIX A - TABLES .......................................................................................................... A-1
 
Rev. 2 of RG 1.21, Page 6


==B. DISCUSSION==
==B. DISCUSSION==
1.  Regulatory Guidance 
Information on the identity and quantity of radionuclides in liquid and gaseous effluents and solid wastes from light-water-cooled nuclear power plants, together with meteorological data representative of principal release points, are needed:  
 
Six basic documents contain the regulatory guidance for implementing the 10 CFR Part 20 and
10 CFR Part 50 regulatory requirements and plant technical specifications related to monitoring and reporting of radioactive material in effluents and environmental media, solid radioactive waste disposal, and the public dose that results from licensed operation of a nuclear power plant:
 
1.
1.


Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste,
For evaluation by the licensee and the Regulatory staff of the environmental impact of radioactive materials in effluents and solid wastes, including estimates of the potential annual radiation doses to the public;
 
2.
2.


Regulatory Guide 4.1, Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants (Ref. 7), 
To ascertain whether AEC regulatory requirements and limiting conditions of operation have been met and whether concentrations of radioactive materials in liquid and gaseous effluents have been kept as low as practicable;
 
3.
3.


Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring Programs (Inception Through Normal Operations to License Termination)Effluent Streams and the Environment (Ref. 8), 
For evaluation by the licensee and the Regulatory staff of the adequacy and performance of containment, waste treatment methods, and effluent controls.


4.
It is essential to have a degree of uniformity in the methods used for measuring, evaluating, recording, and reporting data on radioactive material in effluents and solid wastes. The methods described in this guide provide a uniform basis for comparison of data from different sources and permit the preparation of consistent summaries of data for use by the Regulatory staff as bases for the assessment of a licensee's effluent controls and the potential environmental impact of radioactive materials in effluents and solid wastes.


NUREG-1301, Offsite Dose Calculation Manual Guidance:  Standard Radiological Effluent Controls for Pressurized Water Reactors (Ref. 9), 
This guide outlines general guidelines for monitoring and reporting programs. Detailed specifications for sampling and analysis of effluents are not included since they need to be tailored to the requirements of each specific plant. Standardized methods for monitoring, sampling, and analysis should be used to the extent practicable. The following is an example of a standard which is appropriate for these purposes.


5.
The American National Standards Institute (ANSI)
has developed a standard' which includes general prin ciples and guidance for sampling airborne radioactive materials.


NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Boiling Water Reactors (Ref. 10), and
To assure uniformity of interpretation, the following definitions of terms used in this guide are provided:  
Abnormal releases-unplanned or uncontrolled release of radioactive material from the site boundary.


6.
SANSI N. 13.1-1969, "Guide to Sampling Airborne Radio active Materials in Nuclear Facilities." Copies may be obtained from the American National Standards Institute, Inc., 1430
 
Broadway, New York, N.Y. 10018.
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.
 
These six documents, when used in an integrated manner, provide the basic guidance and implementation details for developing and maintaining effluent and environmental monitoring programs at nuclear power plants. The four regulatory guides specify the guidance for radiological monitoring and the assessment of dose, and the two NUREGs provide the specific implementation details for effluent and environmental 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.  The guide 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.
Batch releases-discontinuous release of gaseous or liquid effluent which takes place over a finite period of time, usually hours or days.


Regulatory Guide 4.1 addresses the environmental monitoring program.  The guide 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 environmental monitoring (both on-site and offsite), including the exposure pathways. The guide defines the exposure pathways, the program scope of sampling media and sampling frequency, and
Continuous release-release of gaseous or liquid effluent which is essentially uninterrupted for extended periods during normal operation of the facility.


Rev. 2 of RG 1.21, Page 7 the methods of comparing environmental measurements to effluent releases in the Annual Radiological Environmental Operating Report.
Determined (or a determination)-a quantitative evaluation of the release or presence of radioactive material under a specific set of conditions.


Regulatory Guide 4.15 provides the basic principles of QA in all types of radiological monitoring programs for effluent streams and the environment. The guide addresses all types of licenses including nuclear power plants. The guide provides the principles for structuring organizational lines of communication and responsibility, using qualified personnel, implementing standard operating procedures, defining data quality objectives (DQOs), performing quality control (QC) checking for sampling and analysis, auditing the process, and taking corrective actions.
A
determination may be made by direct or indirect measurements. In some cases it may not be practical to make direct measurements of specific radionuclides in effluent or waste; e.g., the concentrations may be too low for measurement in a reasonable or practical volume of sample, certain nuclides may be masked by other radionuclides in the sample, or as in the case of solid or concentrated wastes, it may be difficult to obtain a representative sample. Under these circumstances, it may be more appropriate to calculate releases using previously estaibli-shed ratios with those nuclides which are readily measurable.


NUREG-1301 and NUREG-1302 provide the detailed implementation guidance by describing 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.
Such a procedure would constitute a determination.


Regulatory Guide 1.109 provides the detailed implementation guidance for demonstrating that radioactive effluents conform to the As Low as is Reasonably Achievable (ALARA) design objectives of
Elevated release point-the point of release of gaseous waste for which credit was given as such in the determination of the technical specification limit for that release point.
10 CFR 50, Appendix I.  The regulatory guide describes calculational models and parameters for estimating dose from effluent releases, including the dispersion of the effluent in the atmosphere and different water bodies.


Note:  The dose to occupational workers, including contributions from activities associated with effluent programs (such as low-level waste processing, storage and shipping, as well as dose from handling resins and filters for gaseous and liquid radioactive waste) is occupational dose associated with the licensed operation and is not included in RG 1.21.
Ground-level release point-the point of release for gaseous waste which is treated in the technical specifications as having zero height.


The NRC issues regulatory guides to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants.
This guide, which is a revised and rewritten version of Regulatory Guide 1.21 (issued as Safety Guide 21 December 29, 1971), describes acceptable programs for measuring, evaluating, and reporting release of radioactive material in liquid and gaseous effluents and solid wastes from nuclear power plants. It also provides guidelines for calculating potential annual radiation doses to individuals and populations using appropriate models and parameters and pertinent recorded effluent and meteorological data. Significant changes from the previous version are identified below:
 
1. There has been a major change in the format of this guide. The more detailed recommendations concerning radionuclide measurements are presented in Appendix A
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.  Nuclear power reactor licensees may continue to use Revision 1 of Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Waste and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-water Cooled Nuclear Power Plants, issued June 1974, or may adopt other procedures or practices that provide for the measuring, evaluating, and reporting of radioactive material in liquid and gaseous effluents and solid waste.
and the reporting recommendations are indicated in Appendix B.
 
2. Objectives of the Radiological Effluent Control Program
 
The requirements for the radiological effluent control program appear in 10 CFR Part 20 and the technical specifications which are part of a license, including limitations on dose conforming to
10 CFR Part 50, Appendix I. In addition, a facilitys technical specifications describe specific requirements.  These regulatory requirements, 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 analyze
 
Rev. 2 of RG 1.21, Page 8 effluent discharges, 


2.
2.


ensure that effluent treatment systems are used to reduce effluent discharges to ALARA
In many cases the criteria for sensitivity of effluent measurements have been modified to reflect as low as practicable dose considerations in the offsite environs;
levels,
i.e., the sensitivity of effluent measurements should be sufficient to detect concentrations which, when dispersed in the offsite environs, would result in a dose to individuals of a small fraction of natural background radiation.


3.
3.


establish instantaneous release rate limitations on the concentrations of radioactive material, 
Some changes have been made in the frequency of analysis for certain radionuclides in several categories of effluents.
 
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, and 
 
6.
 
evaluate the dose to members of the public.
 
The Annual Radioactive Effluent Release Report (ARERR), submitted before May 1 (unless a licensing basis exists for a different submittal date), and the Annual Radiological Environmental Operating Report (AREOR) submitted annually by May 15 (unless a licensing basis exists for a different submittal date), are used to demonstrate compliance with the facilitys technical specifications 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 discharged in effluents and resultant radiation dose to individuals, 
1.21-2


4.
4.


compliance with the radiation dose limits to members of the public established by the NRC and the U.S. Environmental Protection Agency (EPA), and 
Provisions for monitoring and reporting of solid wastes and for reporting of meteorological measurements, categories not considered in the earlier guide, have been included.


5.
5.


compliance with the effluent reporting requirements of 10 CFR 50.36a.
Provisions for applying the measured meteorological and effluent data to acceptable dose models 2 in calculating potential doses to individuals and populations, and for reporting of these dose estimates have been included.
 
Licensees may also, if they choose to do so, 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 Annual Radioactive Effluent Release Report (ARERR) since the reporting dates may conflict.  If the dates are coincident, or can be met with a single report, licensees may use the ARERR to fulfill the 10 CFR 72.44(d) reporting requirements provided a copy is submitted as specified in 10 CFR 72.44(d)(3). 
 
Rev. 2 of RG 1.21, Page 9


==C. REGULATORY POSITION==
==C. REGULATORY POSITION==
1. Effluent Monitoring 
1.


1.1  Guidance for Effluent Monitoring 
Meteorology A knowledge of meteorological conditions in the vicinity of the nuclear plant is essential to make valid estimates of maximum potential annual radiation doses resulting from radioactive materials released in gaseous effluents. Meteorological measurements should be made in accordance with the guidance set forth in Regulatory Guide 1.23 (Safety Guide 23), "Onsite Meteorological Programs." A summary report of the meteorological measurements taken during each calendar quarter in the
6-month period should be submitted with the semiannual Effluent and Waste Disposal Report as joint frequency distributions of wind direction and wind speed by atmospheric stability class in the format presented in Table 4A of Appendix B to this guide.


Monitoring programs should be established to identify and quantify principal radionuclides in effluents.  NUREG-1301 (for pressurized-water reactors (PWRs)) and NUREG-1302 (for boiling-water reactors (BWRs)) specify the generic controls and surveillance requirements, including the frequency, duration, and methods of measurement. These NUREGs provide specifications for LLDs, requirements for batch releases and continuous releases, sampling frequencies, analysis frequencies and timelines, and composite sample requirements.  Site-specific radiological effluent control programs may differ from the generic NUREG-1301 and NUREG-1302 guidance provided there is either a documented evaluation or justification for such deviations as part of an offsite dose calculation manual (ODCM) authorized change, or if submitted as part of the original ODCM in accordance with Generic Letter 89-01, Implementation of Programmatic and Procedural Controls for Radiological Effluent Technical Specifications, (Ref. 11)
Hourly meteorological data for batch releases should be recorded for the periods of actual release, and quarterly summaries should be reported separately from the summaries of all observations taken during each quarter. The batch release data and the quarterly summaries of all observations should each be given in the format presented in Table 4A of Appendix B.
dated January 31, 1989, and approved by the NRC.


1.2  Release Points for Effluent Monitoring 
For abnormal releases, hourly meteorological data should be recorded for the periods of actual release and should be included in the quarterly summaries of batch releases.


The ODCM should identify the facilitys significant release points (see glossary) used to quantify liquid and gaseous effluents discharged to the unrestricted area.  For those release points containing contributions from two or more inputs (or systems), it is preferable to monitor each major input (or system)
2.
individually to avoid dilution effects, which may impede or prevent radionuclide identification.  NUREG-
1301 and NUREG-1302 contain detailed guidance for the content and format of a licensees ODCM.  For purposes of effluent and direct radiation monitoring, the ODCM should list and/or describe the following: 
 
1.


Significant release points include stacks, vents, and liquid radioactive waste discharge points, among others.
Location of Monitoring All major and potentially significant paths for release of radioactive material during normal reactor
2 Draft Regulatory Guide 1.AA, "Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Implementing Appendix I," Draft Regulatory Guide 1.DD, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents from Routine Releases," and Draft Regulatory Guide 1.EE, "Analytical Models for Estimating Radioisotope Concentration in Different Water Bodies", in Attachment to Concluding Statement, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactors, Docket RM-50-2, USAEC, February 20, 1974.


2.
3 "Final Environmental Statement-Numerical Guides for Design Objectives and Limiting Conditions for Operation to
-'
Meet the Criterion 'As Low as Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," WASH-1258, Vol. 1, Directorate of Regulatory Standards, USAEC, July 1973.


Other release points should be listed in the ODCM if they are not normally classified as one of the significant release points but could become a significant release point based on expected operational occurrences (e.g., primary to secondary leakage for PWRs or failed fuel). This list does not need to be exhaustive or all-inclusive but instead should demonstrate that the licensee has reasonably anticipated expected operational occurrences and their effects on radioactive discharges. Examples may include main steam line safety valves, steam-driven feedwater pumps, turbine building sumps, containment ice condensers, leachate seepage from unlined ponds, or evaporative releases from ponds in the restricted or controlled areas.
operation, including anticipated operational occurrences, should be monitored. Measurements of effluent volume, rates of release, and specific radionuclides should be made, insofar as practicable, at the point(s) which would provide data that are the most representative of effluent releases to the plant environs. For those effluent discharge points which have input from two or more contributing sources within the plant, monitoring of the major contributing sources should also be considered from the standpoint of more effective process and effluent control. In many cases, monitoring of each of the major contributing sources may be a preferable or more sensitive alternative to monitoring the total effluent release when dilution with other less concentrated effluent streams makes the resultant effluent concentrations too low for accurate measurements.


3.
3.


The site environs map should show the following:
Type of Monitoring The type of monitoring selected, including the frequency, duration, and methods of measurement, depends to a large degree on the objectives of the monitoring program. Effluent monitoring is required to (a) demonstrate compliance with technical specification and/or 10 CFR Part 20 effluent limits, (b) allow evaluation of the performance of containment, waste treatment, and effluent controls, and (c)
 
permit evaluation of environmental impact and estimation of the potential annual radiation doses to the public.
a.


significant release points, b.
Because radiation dose is dependent on the radionuclide(s) to which the individual is exposed, monitoring programs should provide accurate information on the identity and quantity of specific radionuclides in effluents and wastes.
 
boundaries of the restricted area and the controlled area (per 10 CFR Part 20
definitions),
c.
 
boundary of the unrestricted area for liquid effluents (e.g., at the end of the pipe or entrance to a public waterway), and d.
 
boundary of the unrestricted area for gaseous effluents (e.g., the site boundary). 
 
Rev. 2 of RG 1.21, Page 10


4.
4.


Dose calculation methodologies should be described for exposure pathways and routes of exposure that are identified in Regulatory Guide 1.109, if applicable.
Gross Radioactivity Measurements Gross radioactivity measurements alone are generally not acceptable for showing compliance with effluent release limits. However, gross radioactivity measurements are often the only practicable means of continuously monitoring effluents and therefore are acceptable under certain specified conditions. Gross radioactivity measurements are acceptable for the purpose of quantifying radioactivity (a) when gross total radioactivity concentrations are a small fraction of the maximum permissible concentrations (MPCs)
for
"unidentified mixtures" as specified in the notes of Appendix B to 10 CFR Part 20 or (b) when gross radioactivity measurements are shown to be truly indicative of the actual quantity and/or concentration of radionuclides released.


5.
5.


Dose calculation methodologies for direct radiation should be described if necessary (e.g.,
Measurements of Specific Radionucfides Measurements should be made to identify specific radionuclides in batch releases prior to their release to the environment. In those cases where analysis of specific radionuclides such as strontium-89 and strontium-90
when assessing direct radiation from the facility). The methodology should include background subtraction, or if appropriate, extrapolation of radiation measurements to points of interest (e.g., to the individual members of the public likely to receive the highest dose).  
cannot be made prior to release, representative samples should be collected from each
1.21-3


The unrestricted area may be defined separately for each of the following:  (1) liquid effluents,
batch of effluents for the purpose of analysis at some later time. The use of composite samples is acceptable, and analyses of such samples should be performed at scheduled frequencies.
(2) gaseous effluents, and (3) if appropriate, for other radiological controls such as direct radiation.


1.3  Monitoring a Significant Release Point
Measurements should be made to quantify specific radionuclides in continuous releases by analyses of grab samples collected at scheduled frequencies.


A significant release point is any location, from which radioactive material is released, that contributes greater than 1 percent of the activity discharged from all the release points for a particular type of effluent considered.  Regulatory Guide 1.109 lists the three types of effluent as (1) liquid effluents, (2)
The frequency of radionuclide analyses should be based on the degree of variance of the concentrations and mixture compositions from an established norm. Continuous monitoring data as well as grab sample data should be the bases for identifying this variance.
noble gases discharged to the atmosphere, and (3) all other radionuclides discharged to the atmosphere.


The ODCM should list significant release points.  Significant release points should be monitored in accordance with the ODCM.  If a new significant release point is identified and is not listed in the ODCM, licensees should (1) establish an appropriate sampling interval (e.g., in site-specific procedures)
SFrequent comparisons should be made between gross r'adioactivity measurements of continuous monitors and analyses of specific radionuclides. These comparisons should be the bases for calibrating cohitinuous monitors to establish relationships between monitor readings and concentrations or release rates of radionuclides in continuous effluent releases.
and (2) update the ODCM within a reasonable timeframe (e.g., yearly).  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 that the amount of radioactive material released, and the corresponding doses, are not substantially underestimated (see 10 CFR Part 50, Appendix I, Section III, Implementation).  If activity is detected when monitoring a significant release point, the radionuclides detected should be reported in the effluent totals (including those with half-lives less than 8 days) in the ARERR (i.e., in Table A-1 or Table A-2), provided that the amount discharged is significant to the three-digit exponential format required for the ARERR.


1.4  Monitoring a Less-Significant Release Point
6.


NUREG-1301/1302 provides tables designating sampling and analysis frequencies for release points. Historically these tables together with the guidance from Revision 1 of RG 1.21 provided the sampling and analysis frequencies.  Licensees may continue to use this guidance from NUREG-1301 and NUREG-1302 and Revision 1 of RG 1.21.  This method of assigning sample frequencies is simple to implement, but in certain cases, it may entail an inappropriately large number of samples for less- significant release points which have no - or extremely low - impact on the parameters reported in the ARERR.  As a result, for less-significant release points, licensees may evaluate and assign more appropriate sample frequencies.  If a licensee wishes to deviate from the sample frequencies listed in NUREG-1301 and NUREG-1302, the licensees evaluation, showing that the effectiveness of the radioactive effluent control program is not reduced, should be maintained in site documentation.
Representative Samples A sample should be representative of the bulk stream or volume of effluent from which it is taken.


Regardless of the surveillance frequencies, if activity is detected when monitoring a less-significant release point, the licensee must (per 10 CFR Part 50.36a and 10 CFR Part 50, Appendix I, Section III.A.1) report the cumulative activity in the effluent totals (i.e., in Table A-1 or Table A-2) in the ARERR (provided that the amount discharged is significant to the three-digit exponential format required for the ARERR).  
Provisions should be made to assure that representative samples are obtained from well-mixed streams or volumes of effluent by the selection of proper sampling equipment, the proper location of sampling points, and the development and use of vroper sampling procedures.


Rev. 2 of RG 1.21, Page 11
Prior to sampling, large volumes of liquid waste should be mixed in as short a time interval as practicable to assure that any sediments or particulate solids are distributed uniformly in the waste mixture. Sample points should be located where there is a minimum of disturbance of flow due to fittings and other physical characteristics of the equipment and components.


Site documentation should identify less-significant release points, to the extent reasonable, but it is not necessary to list all possible release points in site documentation. Releases from a less-significant release point may be assessed (see section 5.1, Bounding Assessments) to the extent reasonable using assumptions and bounding calculations (in lieu of, or in addition to, sampling and analysis).  When plant conditions change, and such changes may reasonably affect the status of a less-significant release point (e.g., significant change in primary-to-secondary leakage in PWRs or substantial cross-contamination between systems), sampling and analysis of the affected less-significant release points should be conducted. These sample results should be evaluated to (1) confirm the continued validity of the bounding calculations (if used) regarding effluent accountability and (2) determine the impact (if any) on effluent accountability.  The guidance in this regulatory guide regarding monitoring less-significant 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 Inspection and Enforcement (IE) Bulletin 80-10, Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment, dated May 6, 1980 (Ref.  12).  
Sample nozzles should be inserted into the flow or liquid volume to ensure sampling the bulk volume of pipes and tanks. Sample lines should be flushed for a sufficient period of time prior to sample extraction in order to remove sediment deposits and air and gas pockets.


1.5  Monitoring Leaks and Spills
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 assure that individual samples are indeed representative of the effluent mixture.


An area where an unplanned release occurred into the on-site environs (e.g., a leak or spill) should be identified as an impacted area for decommissioning purposes in accordance with NUREG-1757, Consolidated Decommissioning Guidance, issued September 2006 (Ref.  13).  A leak or spill should be assessed to obtain the necessary information for the ARERR as specified in Regulatory Position 8.5.1, Abnormal Releases or Abnormal Discharges (see glossary).  Leaks or spills to the ground will be diluted on contact with soil and water in the environment.  Samples of the undiluted liquid (from the source of the leak or spill) and samples of the affected soil (or surface water or ground water) should be analyzed 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, ground water monitoring may be used as a surrogate for soil sampling.  If sampling is not practical, the 10 CFR 50.75(g)
The general principles for obtaining valid samples of airborne radioactive material, the methods and materials for gas and particle sampling, and the guides for sampling from ducts and stacks contained in ANSI
records should describe why sampling was not conducted (e.g., the area was inaccessible or there were safety considerations).  The location and estimated volume of the leak or spill should be recorded to identify the extent of the impacted area and predicted size or extent of the contaminant plume.  If a spill is promptly remediated (e.g., within 48 hours) and if subsequent surveys of the remediated area indicate no detectable residual radioactivity remaining in the soil or ground water (see paragraph below), then, for purposes of reporting discharges in the ARERR, there was no liquid discharge to the unrestricted area, and the spill need not be reported in the ARERR.  However, the decommissioning file should be updated to include a description of the event as specified by 10 CFR 50.75(g).  Licensees should review the decommissioning files before generating the ARERR to ensure that the ARERR includes the necessary information regarding leaks and spills.
N13.1-1969 1 are generally acceptable and provide ade quate base, for the design and conduct of monitoring programs ior airborne effluents.


When evaluating areas that have been remediated, the licensee should survey for residual radioactivity.  There may be times when the licensee wants to verify that an area contains no residual radioactivity.  There is existing regulatory guidance and information on analytical detection capabilities.
7.


Licensees should ensure that surveys are appropriate and reasonable (as defined in 10 CFR 20.1501). 
Composite Samples To be representative of the average quantities and concentrations of radioactive materials released in liquid and in particulate form in gaseous effluents, samples for compositing 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. Prior to analysis, the composite should be thoroughly mixed S'
Licensees should generally ensure that surveys are conducted using the appropriate sensitivity levels (e.g., refer to the environmental LLDs in NUREG-1301 and NUREG-1302, Table 4.12-1, Detection Capabilities for Environmental Sample Analysis, or LLDs determined by using the methodology outlined in NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols Manual, (MARLAP)
that the sample is representative of the average effluent release.
issued July 2004 (Ref. 14)).  Additionally, licensees should apply plant-process-system knowledge when evaluating leaks and spills. For example, consider a hypothetical case of a leak in a condensate storage


Rev. 2 of RG 1.21, Page 12 tank.  Assume that the tanks contents were analyzed 30 days before the leak and determined to contain
Periods of collection for composites should be as short as practicable to preclude the loss of radioactive material by deposition on walls of the sample container or volatilization of potentially volatile material. Periodic checks should be performed to identify any such changes in composite samples.
1.2x10-6 microcuries per milliliter (uCi/mL) of tritium (1,200 picocuries/liter (pCi/L)).  Additionally, assume that historical records indicate that the tank contained detectable levels of tritium about 50 percent of the time, and that tritium concentrations never exceeded 2,000 pCi/L of tritium.  In this example, the licensee discovers a leak in the tank and is able to fix the leak after 400 gallons (1,500 liters) of water leaked to the ground surface.  The licensee confirms the presence of tritium by sampling the tank contents and/or the wetted soil.  Based on those results, the licensee chooses to remediate the affected soil and excavates the affected soil and places the removed soil into suitable containers.  The licensee then samples undisturbed soil from several locations within the excavated area and analyzes the soil for tritium.  The licensee adjusts the analytical method and the analytical sensitivity to allow detection of (the equivalent of)
1,000 pCi/L of tritium in the water fraction.  The licensee analyzes the soil (for gamma activity) and the water fraction of soil (for tritium activity) from the excavated area and detects no radionuclides.  The licensee also confirms radioactive material did not reach the water table by verifying the excavated area is above the water table. The NRC would find this to be an acceptable method for the licensee to use in concluding that there is no detectable residual radioactivity from the spill listed in this example.


This regulatory guide provides guidance regarding information the licensees should provide in the ARERR.  In that context, when leaks and spills of radioactive material are identified, prompt response and timely actions should be taken to the extent reasonable to (1) evaluate radiological conditions and
8.
(2) ensure proper reporting of materials discharged off site.  To realize these two goals, it may be necessary to isolate the leak or spill at the source, prevent the spread of the leak or spill, and remediate the affected area (if the licensee deems remediation to be reasonable and necessary).  For leaks and spills involving the discharge of radioactive material to the unrestricted area, the dose to members of the public from the leak or spill should be evaluated using realistic or bounding exposure scenarios.  (See Attachment 6 to SECY-
03-0069 for more information on use of realistic scenarios.)  However, for leaks or spills that occur on site, a realistic dose assessment to an offsite member of the public may become complicated especially if (1) no radioactive material has entered the unrestricted area and (2) there are no members of the public on site.


For leaks and spills, licensees should perform surveys that are reasonable to evaluate the potential radiological hazard (as described in 10 CFR 20.1501).  As a result, for leaks and spills, licensees may choose to use bounding assessments to estimate the potential hazard.  For example, if a leak occurs on site and radioactive material is released at or below the ground surface, the licensee may choose to assess the potential hazard by assuming that a conservatively large (e.g., bounding) volume of water is part of an assumed exposure pathway (e.g., drinking water).  Such assumptions would allow the licensee to assess the potential hazard to a hypothetical individual member of the public. A hazard assessment of this sort would be appropriate for inclusion in the supplemental information section of the ARERR.  In such cases where there is no real exposure pathway to a member of the public, the licensee should indicate that the hazard assessment is a bounding estimate of the dose to a hypothetical individual member of the public and no actual exposure was received by a real individual member of the public.
Time between Collection and Analysis Measurements should be made as soon as practicable after collection to minimize loss of short-lived radionuclides by decay. Measurement of longer-lived radionuclides sometimes can be simplified by allowing sufficient time before their analysis for the decay of short-lived radionuclides.


If licensees choose to notify local authorities of spills or leaks (e.g., because of local ordinances or local and State government agreements), the licensee should review the reporting requirements of
Procedures should be instituted for handling, packaging, and storing samples to assure that loss of radioactive matenals or other factors causing sample deterioration do not invalidate the analysis.
10 CFR 50.72(b)(xi) and information in NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and
50.73, (Ref. 15), for applicability.  In such situations, licensees should ensure effective communication using the guidance provided in NUREG/BR-0308, Effective Risk Communication, (Ref. 16), especially with respect to ensuring that the risk is described in the appropriate context.  In general, licensees should notify the NRC when significant public concern is raised, in accordance with 10 CFR 50.72(b)(xi).  


Rev. 2 of RG 1.21, Page 13 Although the licensee may choose to use its problem identification and resolution program (corrective action program) to document the evaluation of the spill or leak, appropriate documentation should be placed in, or cross-referenced to, the decommissioning files as required by 10 CFR 50.75(g). 
9.


Remediation should be evaluated and implemented as appropriate based on licensee evaluations and decision-making.  Evaluation factors should include (1) the location and accessibility, (2) the concentrations of radionuclides and extent of the residual radioactivity, (3) the efficacy of monitored natural attenuation, (4) the volume of the release, (5) the mobility of the radionuclides, (6) the depth of the water table and (7) whether significant residual radioactivity (see glossary) is expected at the time of decommissioning.  Since the contaminants, concentrations, and extent of contamination are expected to vary over time or plant life (either increase based on anticipated future leaks and spills or decrease based on remediation or monitored natural attenuation), no one set of numerical values defines significant residual radioactivity.  However, licensees may make remediation decisions based on their expectations of being able to meet the decommissioning criteria of 10 CFR 20.1402, Radiological Criteria for Unrestricted Use, at the anticipated time of decommissioning.
Corrections for Decay Decay corrections should be made as though the effluent were released uniformly throughout the sampling period unless it is shown that most of the effluent was released during a particularly short interval.


Information that may be useful in this decision-making includes (1) NUREG-1757, Volume 1, Appendix H, Memorandum of Understanding between the Environmental Protection Agency and the Nuclear Regulatory Commission, (2) NUREG-1757, Volume 2, Derived Concentration Guideline Levels in Table H.1, and (3) the derived concentration guideline levels that have been authorized for decommissioned nuclear power plants.  For a more detailed analysis, licensees may use the RESRAD
The exact time or time intervals of sample collection should be recorded. To estimate radioactive decay in composite or pooled samples, weighting should be applied to the delay time of each portion and to the quantity of each portion in relation to the total quantity of the sample.
computer codes available from Argonne National Laboratory (Refs. 17, 18, and 19) or equivalent.


1.6  Monitoring Continuous Releases 
10. Sensitivity The sensitivity limits given for radioactivity analyses in Appendix A of this guide are based on the potential significance in the environment of the quantities of radioactive materials released. For some radionuclides, lower detection limits than those given herein may be readily achievable and when measurements below the stated sensitivity limits are attained, the results should be recorded and reported.


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, are generally not acceptable for demonstrating compliance with effluent discharge limits.
For certain mixtures of gamma-emitting nuclides, it may not be possible to measure certain radionuclides at the stated sensitivity limits when other radionuclides are present in the sample in much greater concentrations.


The use of continuously indicating radiation monitoring system results may be combined with sample analyses to more fully characterize and quantify a discharge. This technique may have particular applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during a release or (2) when there is a desire to verify whether a preliminary grab sample is representative. In these instances the radiation monitor responses (i.e., the radiation monitor efficiencies) for various radionuclides should be well characterized.
Also, it may not be possible to measure certain radionuclides whose gamma ray yields are low (e.g.,
Kr-8 5, Cr-5 1, etc.) at the stated sensitivity limits. Under these circumstances, and in the case of radionuclides
1.21-4


Grab samples should be collected at scheduled frequencies (see NUREG-1301 and NUREG-1302 or as approved in Generic Letter 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 discharge and (2) the relative radionuclide composition from an established norm.  Where the magnitude of the discharge 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 discharge.  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 verify (or modify) the conversion factors that correlate radiation monitor readings and concentrations of radionuclides in effluents.
which have no gamma rays and weak beta radiation (e.g.,  
Fe-55, Ni-63, etc.), it may be more appropriate to calculate releases of such radionuclides using measured ratios of these radionuclides to those radionuclides which are routinely identified and measured.


Rev. 2 of RG 1.21, Page 14
Measurements should be made periodically to establish and assure the continued validity of the ratios used. Any reported data determined by this method should be clearly identified.


1.7  Monitoring Batch Releases 
11. Accuracy of Measurements a.


For batch releases, measurements should be performed to identify principal radionuclides before a release.  In those cases in which an analysis of specific hard-to-detect radionuclides (such as strontium-
Errors in Measurements An estimate should be made of the error associated with measurement of radioactive materials in effluents and solid wastes. Counting statistics can provide an estimate of the minimum error involved in radioactivity analyses.
89/90 and iron-55 in liquid releases) 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 NUREG-1301 and NUREG-1302 or, for less-significant release points, at the frequencies specified by the licensee.  (See Regulatory Position 1.4.) 


The use of continuously indicating radiation monitoring system results may be combined with sample analyses to more fully characterize and quantify a discharge.  This technique may have particular applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during a discharge or (2) when there is a desire to verify whether a preliminary grab sample is representative.  In these instances the radiation monitor responses (i.e., the radiation monitor efficiencies) for various radionuclides should be well characterized.
Counting statistics (e.g.,  
one-sigma counting error) should be included in the records of measurements, since they provide a readily calculable estimate of the statistical uncertainty due to counting.


1.8  Principal Radionuclides for Effluent Monitoring
The total or maximum error associated with the effluent measurement will include the cumulative errors resulting from the total operation of sampling and measurement. Because it may be very difficult to assign error terms for each parameter affecting the final measurement, detailed statistical evaluations of error are not suggested. The objective should be to obtain an overall estimate of the error associated with measurements of radioactive materials released in liquid and gaseous effluents and solid waste.


During analysis of samples, licensees should apply the appropriate analytical sensitivities to ensure adequate surveys are conducted.  NUREG-1301/1302 provides a list of principal gamma emitters for which an LLD control applies.  Historically, this list together with the guidance from Revision 1 of RG
b.
1.21 provided the appropriate sensitivity levels for an analysis.  Licensees may continue to use this guidance, which essentially classifies all radionuclides as principal radionuclides, and apply the analytical sensitivity levels (e.g., LLDs) directly from NUREG-1301 and NUREG-1302 and Revision 1 of RG 1.21.


This method is simple to implement, but in certain cases, it may entail inappropriately long count times or it may involve alternate (or unnecessary) methods of analysis for low-activity radionuclides with no - or extremely low - dose significance.
Quality Controls Control checks and tests should be applied to the analytical process by the use of blind duplicate analyses of selected effluent samples and by cross-check analysis of selected samples with an independent laboratory. Quality controls should also be applied to the entire sample-collection procedure to assure that representative samples are obtained and that samples are not changed or affected prior to their analysis because of handling or because of their storage environment.


Although the LLD list from NUREG-1301 and NUREG-1302 may be used for determination of principal radionuclides, in reality, the principal radionuclides at a site will be dependent on site-specific factors such as (1) the amount of failed fuel, (2) the extent of system leakage, (3) the sophistication of radioactive waste processing equipment, and (4) the level of expertise in operating radioactive waste processing system.
c.


Since the principal radionuclides will vary from site to site, licensees who wish to deviate from the historical method of determining principal radionuclides (as described above) may adopt a risk-informed approach to identify principal radionuclides (and the associated sensitivity levels) at a site.
Calibrations Individual written procedures should be prepared and utilized for specific methods of calibrating radiological monitoring systems and measuring equipment. Calibration practices for ancillary equipment and systems are described in Regulatory Guide 1.23,
"Onsite Meteorological Programs," and elsewhere, I and where appropriate, they should be utilized and included as a part of the 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.


This regulatory guide introduces the term principal radionuclide in a risk-informed context.  A
Calibration procedures should identify the specific equipment or group of instruments to which the procedures apply.
licensee may evaluate the list of principal radionuclides for use at a particular site.  The principal radionuclides may be determined based on their relative contribution to (1) the public dose compared to the 10 CFR 50 Appendix design objectives or (2) the amount of activity discharged compared to other site radionuclides.  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) outlined in a licensees ODCM.  This implementation of primary radionuclides ensures both
(1) radionuclides that are present in relatively large amounts but that contribute very little to dose, and (2)
radionuclides that are present in very small amounts but that have a relatively high contribution to dose are appropriately included in the ARERR.


Rev. 2 of RG 1.21, Page 15 NOTE:  With respect to principal radionuclides, dose is the measure of risk whereas activity is not. For example, a relatively large amount of tritium released into a large body of water has little dose significance.
Calibrations of measuring equipment should be performed using reference standards certified by the National Bureau of Standards or standards that have been calibrated against standards certified by the National Bureau of Standards. Calibration standards should have the necessiry accuracy, stability, and range required for their intended use.


If adopting a risk-informed perspective, a radionuclide is considered a principal radionuclide if it contributes either (1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective dose for all radionuclides in the type of effluent being considered, or (2) greater than 1 percent of the activity of all radionuclides in the type of effluent being considered. Regulatory Guide 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble gases released to the atmosphere, and (3) all other radionuclides released to the atmosphere. In this context, the term principal radionuclide has special significance with respect to the required sensitivity levels (e.g., LLDs) for an analysis.  The LLDs specified in NUREG-
Calibrations should generally be performed at regular intervals. Frequency of calibration should be based on the reproducibility and time stability of the system. 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 which gives measurements within a more narrow range. In many cases, it would be more appropriate to calibrate measuring equipment before and after use in addition to or instead of calibration at arbitrarily scheduled intervals. Calibration of measuring equipment before and after use permits detection of any erroneous readings or malfunctions that may have occurred during use.
1301/1302 may be used, or LLDs may be determined based on the other methodologies (e.g., as outlined in MARLAP). Once principal radionuclides are identified, they should be monitored in accordance with the sensitivity levels (e.g., LLDs) listed in the ODCM.


For radionuclides that are not identified as principal radionuclides, licensee discretion may be applied to the sensitivity of analysis provided that there is no reduction in the effectiveness of the radioactive effluent control program. If analytical sensitivities are chosen that are different from those in NUREG-1301 and NUREG-1302, the basis for the deviations should be documented. For example, data quality objectives (DQOs) and other concepts from Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring Programs (Inception through Normal Operations to License Terminations)
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. Functional checks, i.e., routine checks performed to demonstrate that a given instrument is in working condition and functioning properly, may be performed using radioactive sources that are not standards.
Effluent Streams and the Environment, Revision 2, issued July 2007 (Ref. 20), may be useful for determining risk-informed sensitivity levels for an analytical method.


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 becomes the controlling document and specifies the list of principal radionuclides.  If adopting this method, the ODCM should be updated with the list of principal radionuclides within 1 year of their identification.  Licensees are allowed to revise the ODCM in accordance with the ODCM change process as described in the plants technical specifications  (which includes documented evaluations of such changes).  
Continuous radioactivity monitoring systems should be calibrated against appropriate standards and the relationship established between concentration and monitor readings over the full range of the readout device. Adequacy of the system should be judged on the basis of reproducibility, time stability, and sensitivity.


The concept of principal radionuclides does not reduce the requirement for reporting radionuclides detected in effluents. In addition to principal radionuclides, other radionuclides detected during routine monitoring of release points should be reported in the radioactive effluent release report and included in dose assessments to members of the public.
Periodic inservice calibrations should also be performed to relate monitor "readings"
to the concentrations and/or release rates of radioactive material in the monitored release path. These calibrations should be based on the results of analyses for specific radionuclides in grab samples from the release path.


1.9  Carbon-14
12. Expression of Results of Measurements a.


Carbon-14 (C-14) is a naturally occurring isotope of carbon.  Nuclear weapons testing in the 1950s and 1960s significantly increased the amount of C-14 in the atmosphere.  C-14 is also produced in commercial nuclear reactors, but the amounts produced are much less than those produced naturally or from weapons testing.  Since the NRC published Regulatory Guide 1.21, Revision 1, in 1974, the analytical methods for determining C-14 have improved.  Coincidentally the radioactive effluents from commercial nuclear power plants over the same period have decreased to the point that C-14 is likely to be a principal radionuclide (as defined in this document) in gaseous effluents.
Units The information and data on effluent releases included in reports to the Commission should be expressed in the units given in Appendix B of this guide and reported in the form given in paragraphs b and c below.


Rev. 2 of RG 1.21, Page 16 C-14 releases in PWRs occur primarily as a mix of organic carbon and carbon dioxide released from the waste gas system.  In BWRs, C-14 releases occur mainly as carbon dioxide in gaseous waste (Ref.  21).  Because the dose contribution of C-14 from liquid radioactive waste is much less than that contributed by gaseous radioactive waste, evaluation of C-14 in liquid radioactive waste is not required.
b.


Many documents provide information about the magnitude of C-14 in typical effluents from commercial nuclear power plants (e.g., Refs. 21, 22).  Those documents suggest nominal annual releases of C-14 in gaseous effluents are approximately 5 to 7.3 curies from PWRs and between 8 to 9.5 curies from BWRs.
Significant Figures To avoid ambiguity, significant figures should be used in recording the results of effluent
1.21-5


Licensees should evaluate whether C-14 is a principal radionuclide for gaseous releases from their facility.
measurements. When several numbers are multiplied or divided together, the result should be rounded off to as few significant figures as are present in the factor with the fewest significant figures. When numbers are added or subtracted, the number with the fewest decimal places, not necessarily the fewest significant figures, puts the limit on the number of places that may justifiably be carried in the sum or difference.


10 CFR 50.36a requires that operating procedures be developed for the control of effluents and that quantities of principal radionuclides be reported.  The quantity of C-14 discharged can be estimated by sample measurements 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, issued January 1985 (Ref. 23)) or estimated by use of the GALE code from NUREG-
For the purpose of reporting in the format of Appendix B of this guide, numerical values should be rounded off to three figures.
0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors PWR-GALE Code, April 1985 (Ref. 22).  Because the production of C-14 is expected to be relatively constant at a particular site, if sampling is performed for C-14 (instead of estimating C-14 discharges based on calculations from a normalized source term), the sampling frequency may be adjusted to that interval that allows adequate measurement and reporting of effluents.  If estimating C-14 based on scaling factors and fission rates, a precise and detailed evaluation of C-14 is not necessary.  It is not necessary to calculate uncertainties for C-14 or to include C-14 uncertainty in any subsequent calculation of overall uncertainty.


1.10  Abnormal Releases and Abnormal Discharges
c.


In the previous revision of the Regulatory Guide 1.21, the terms release and discharge were synonymous.  This regulatory guide uses the term release to describe an effluent from the plant (regardless of where the effluent is deposited), whereas the term discharge is used only to describe an effluent that enters the unrestricted area. Although the term release includes effluents to either (1) the on-site environs or (2) the unrestricted area, for purposes of this regulatory guide, the use of the term release will generally be reserved for those instances when an effluent is released from the power plant into the on-site environs. The on-site environs in this context encompass locations outside of nuclear power plant systems, structures, and components as described in the final safety analysis report or ODCM.
Numerical Values Results of measurements, including percentages, should be reported in external floating point form, using the letter "E" to denote the exponent to the base 10. For example: 2% should appear as
2.00E+00; 0.00032 should appear as 3.20E-04; 157.6 should appear as 1.58E+02; 2.67 should appear as  
2.67E+00.


This is a change in terminology with respect to the definition of abnormal release in Regulatory Guide 1.21, Revision 1, which defined abnormal releases to be from the site boundary.
The term "not detected" should not be used. If radioactivity in the sample(s) is less than the maximum sensitivity of measurement, the value should be reported as less than the maximum sensitivity. For example, if the maximum sensitivity is 3 x 10-9 uCi/ml, the values should be reported as <3.OOE-09.


An abnormal release (see glossary) is an unplanned or uncontrolled release of licensed radioactive material from the plant. Abnormal releases may be categorized as either batch or continuous depending on the circumstances.  By contrast, an abnormal discharge (see glossary) is an unplanned or uncontrolled release of licensed radioactive material to the unrestricted area.  Abnormal discharges may also be categorized as either batch or continuous depending on the circumstances. The distinction between the terms abnormal release and abnormal discharge is important for describing the staff position for measuring, evaluating, and reporting releases and discharges, especially where leaks and spills are involved.
13. Radiological Impact on Man Estimations of doses to individuals and populations are necessary for the assessment of the radiological impact on man from the operation of nuclear power plants. Dose calculations should be made using the measured effluent and meteorological data and acceptable dose models such as those provided in draft regulatory guides for implementation of numerical guides. 2 To the extent that they are not inconsistent with the models provided in these draft guides, other dose models such as those given in WASH-1258 3 or those used for calculating the estimated dose values given in the licensee's Environmental Report are also acceptable as bases for making dose calculations.


That portion of an abnormal release that is discharged to the unrestricted area is reported as a abnormal discharge in the year in which the discharge occurred.  The portion of an abnormal release that remains on site is considered residual radioactivity (see 10 CFR 20) and is documented in accordance with
14. Other Provisions The provisions and principles presented in Appendices A and B of this guide are acceptable to the Regulatory staff as bases for measuring and reporting of radioactive materials in liquid and gaseous effluents and solid wastes from nuclear power plants, as well as for estimating doses to individuals and populations in the offsite environs.
10 CFR 50.75(g).  


Rev. 2 of RG 1.21, Page 17
1.21-6


Low-level radioactive system leakage resulting 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 captured by, or directed to, a system designed to accept and handle radioactive material including the subsequent planned and controlled discharge of the radioactive material (e.g., as described in the FSAR or ODCM), that evolution is not considered an abnormal release.  Normal system leakage captured by effluent ventilation control systems or sumps is not an abnormal release (provided that, before discharge of the radioactive material, the discharge is planned and controlled).  (See also the definitions of unplanned release and uncontrolled release in the glossary.) 
APPENDIX A
MEASURING RADIOACTIVE MATERIALS IN LIQUID
AND GASEOUS EFFLUENTS AND SOLID WASTE
This appendix describes a monitoring program that is acceptable to the Regulatory staff. The frequencies of sampling and analysis and the types of measurements described are considered to be the minimum acceptable.


In certain circumstances, some subjectivity may be associated with the definitions of unplanned release and uncontrolled release.  In these situations, additional circumstances should be considered to determine if an abnormal release occurred.  A well-designed and documented evaluation of a release point can include an evaluation of the potential for an unplanned or uncontrolled release. The evaluation can establish bounding criteria that establish a threshold for an abnormal release based on planning and control.  Generally, releases that may reasonably be categorized as both unplanned and uncontrolled should be considered abnormal releases.
In some cases, this program should be supplemented with additional measurements because of individual plant design features or other factors. The need for supplemental or modified programs is determined on a case-by-case basis.


For example, consider an underground pipe that carries radioactive liquid to an outside storage tank.  If this pipe develops a leak, and licensed radioactive material escapes into the surrounding soil, it is considered an abnormal release if some portion or all of the radioactive material remains on site.  This type of leak should be reported as an abnormal release in the next ARERR.  If the licensee predicts (e.g., based on site conceptual model and subsequent ground water monitoring results) that the radioactive material will enter the unrestricted area in 2 years, the resulting radioactive discharge (that would occur 2 years hence) will be considered an abnormal discharge.  Therefore, the resulting radioactive discharge should be reported along with other data for the affected calendar year in a future ARERR (i.e., in this example,
A.
3 years later).  Both releases and discharges (either routine or abnormal) should be reported on a calendar- year basis for the year in which the release or discharge occurred.


Consider another example involving a volume of radioactive gas from the containment atmosphere that escapes the equipment hatch during a refueling outage (especially during the time interval when the containment purge exhaust fans are off).  This would generally not be considered an abnormal discharge if
GASEOUS EFFLUENTS
(1) the duration was preplanned (e.g., for a short duration such as 12 hours), (2) the containment activity (gas, particulate, tritium, and iodine) was preplanned, known, and very low (e.g., such that a bounding estimate of the radioactive material discharged indicated there would be no measurable impact relative to typical discharges), (3) the containment activity was monitored (e.g., by sampling or radiation monitoring equipment), and (4) an evaluation was completed to identify a preplanned limiting (or trigger) level of activity that would initiate remedial or mitigating action (e.g., close the equipment hatch to control gases escaping containment).  In this example, the actions taken (i.e., preplanning and monitoring) before and during the evolution are sufficient to establish control of this discharge.  As a result, this type of evolution should not be categorized as an abnormal discharge.
Continuous monitoring should be conducted along principal gaseous effluent discharge paths.


===2. Effluent Sampling ===
The radionuclide composition and quantities and concentrations of radioactive material in gaseous effluents should be determined and recorded. For the periods of release, the records should also show, on an hourly basis, the existing meteorological conditions of wind direction, wind speed, and atmospheric stability which are representative of conditions at the principal points of release (see Regulatory Guide 1.23, "Onsite Meteorological Programs"). 
The single Poisson (one sigma) error for discrete measurements should be less than 50 percent for release rates at the design objective level, less than 30 percent at twice the design objective release rate, and less than 20
percent at eight times the design objective release rate.


2.1  Representative Sampling 
1.


A typical schedule for radioactive effluent sample collection and analyses appears in NUREG-
Fission and Activation Gases During the release of gaseous wastes from the primary system waste gas holdup system, the effluent monitor should be operating and set to alarm and to initiate the automatic closure of the waste gas discharge valve before the limits specified in the technical specifications are exceeded.
1301 and NUREG-1302.  Some licensees may have modified these sampling schedules (typically contained in the ODCM) as part of implementing Generic Letter 89-01 as approved by the NRC.


Rev. 2 of RG 1.21, Page 18 Additional samples should be obtained as needed to characterize abnormal releases, abnormal discharges, or other 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.
a.


2.2  Sampling Liquid Radioactive Waste
Continuous Releases For reactors which release gases continuously, a sample of the gaseous effluent should be analyzed within one month after the date of initial criticality of the reactor and at least weekly thereafter to determine the identity and quantity of the principal radionuclides being released. A similar analysis of samples should be performed following each refueling, process change, or other occurrence that could alter the mixture of radionuclides. For those processes or other conditions that change significantly (e.g., when the average daily gross radioactivity release rate equals or exceeds that given in the technical specifications or when the steady-state gross radioactivity release rate increases by
50% over the previous steady-state release rate at the same power level), an analysis should'be done following each change until it is shown that a pattern exists that can be used to predict the isotopic composition of the effluent. In addition, radionuclide analyses should be performed when continuous monitoring shows an unexplained variance from an established norm which may be indicative of a change in the concentration and composition. The norm should be established as a range of readings that may be expected due to normal operating conditions including anticipated operational occurrences.


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 recirculated three or more 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. 24).  If tank-mixing practices deviate from industry standards (i.e., those for recirculation or other), 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)
The calibration of continuous gross radioactivity monitoring systems should be performed by normalizing against the results of specific radionuclide analyses using established ratios of the respective radionuclides to total activity.
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 that 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 before 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.


2.3  Sampling Gaseous Radioactive Waste 
When calibrated in this fashion, the gross radioactivity measurements obtained from continuous monitors may be used to determine the total quantity of radioactivity released.


Although all licensees may not be committed to 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. 25), 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. 26), the documents contain the general principles for designing and conducting monitoring programs for airborne effluents.  The cited references also contain recommendations for obtaining valid samples of airborne radioactive material in effluents and the guidelines for sampling from ducts and stacks.  Licensees should use the appropriate licensing documents to evaluate the validity of representative samples (e.g., evaluate the potential for inaccurate sampling of gaseous effluents that may bypass a particulate filter and collect on an iodine collection cartridge) and to identify any inaccurate sample analyses configurations or counting geometries.
b.


2.4  Sampling Bias
Batch Releases For reactors which release gases intermittently, an analysis should be made of a representative sample of each planned release prior to discharge to determine the identity and quantity of the principal radionuclides released.


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. 26) or equivalent.
Continuous monitoring should also be conducted at appropriate points to obtain information on the quantity and pattern of abnormal releases.


Rev. 2 of RG 1.21, Page 19
c.
2.5  Composite Sampling


Composite samples should be representative of the average quantities and concentrations of radioactive materials discharged in liquid and gaseous effluents.  Composite samples should be collected in proportion to the effluent flow rate or in proportion to the volume of each batch of effluent discharges.
Sensitivity For those discharge points which have input from two or more contributing sources within the plant, separate monitoring of the major sources should be performed as a more sensitive alternative to monitoring the composite effluent stream when bulk dilution results in concentrations too low for accurate measurements.


2.6  Sample Preparation and Preservation
The sensitivity of gross radioactivity measurements of fission and activation gases, as a minimum, should be sufficient to permit measurement of a small fraction of the activity which would result in
(1) an annual air dose of 10 millirads due to gamma radiation at any location near ground level at or beyond the site boundary and (2) an annual air dose of 20
millirads due to beta radiation at any location near ground level at or beyond the site boundary.


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 material discharged.
The sensitivity of analysis for each of the principal radioactive gases in representative samples of gaseous effluents should be such that concentrations of  
1 e ACi/cc are measurable.


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.
1.21-7


2.7  Short-Lived Radionuclides and Decay Corrections 
2.


In the analysis of short-lived radionuclides (e.g., short-lived noble gases), measurements should generally be made as soon as practical after collection to minimize loss by radioactive decay.  In other cases, when needed to improve the detection of the longer-lived radionuclides, time should be allowed for the decay of short-lived, interfering radionuclides.
lodines a.


Some special considerations may be applicable in those instances where short-lived radionuclides are being measured. In general, sample collection (or analysis frequencies) should take into account the half-lives of the radionuclides being measured.  This may have special applicability for continuous samples or composite samples.  It is generally best to select a compositing interval (and analysis frequency)
Monitoring A representative sample from the principal discharge paths should be drawn continuously through an iodine sampling device. The sample collected in the device should be analyzed at least weekly for iodine- 131.
appropriate for the effluent (radionuclide) being analyzed.  In cases where the compositing interval is selected appropriately, analytical bias is minimized.  One way to avoid analytical bias is to decrease the composite sampling interval (and analysis frequency).  


To minimize bias in measurements, it may be necessary to decay correct analysis results for short-lived radionuclides.  Licensees should be cognizant of those situations in which analytical bias may be introduced when analyzing short-lived radionuclides and should select appropriate methods to minimize such bias.
An analysis should also be made monthly or more often for iodine-133 and iodine-135.


3  Effluent Dispersion (Meteorology and Hydrology)
The results of these analyses should be used as the basis for recording, evaluating, and reporting the quantities of radioiodines released during the sampling period. In estimating releases for periods when analyses were not performed, the average of the two adjacent data points spanning this period should be used. These estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.


3.1  Meteorological Data 
b.


Gaseous effluents discharged into the atmosphere are transported and diluted as a function of  
Sensitivity The sensitivity of the analysis of radioiodines should be sufficient to permit measurement of a small fraction of the activity which would result in annual exposures of 15 millirems to the thyroid of individuals in unrestricted areas.
(1) the atmospheric conditions in the local environment, (2) the topography of the region, and (3) the characteristics of the effluents.  Licensees should consider the guidance in Regulatory Guide 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (Ref. 27), in the development and implementation of site programs designed to collect site-specific meteorological data.  The meteorological data do not need to be reported in the ARERR, but the data should be summarized and maintained as


Rev. 2 of RG 1.21, Page 20
3.
documentation (records).  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.


3.2  Atmospheric Transport and Diffusion 
Particulates a.


Site-specific meteorological data collected should be analyzed and used to generate gaseous effluent dispersion factors (/Q) and deposition factors (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. 28).  The use of annual average meteorological conditions to determine /Q and D/Q is appropriate for continuous releases and for establishing instantaneous release set points (see NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, issued October 1978 (Ref. 29)).  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, 5 (or more) years of data should be used.  If long-term, annual average /Q and D/Q values are used in determining dose to individual members of the public, the values should be revalidated or updated periodically (e.g., every 3 to 5 years).  If the evaluation indicates the long- term, annual average /Q and D/Q are nonconservative by 10 percent or more, either revise the affected values or document the reason why such changes are not deemed necessary.
Monitoring A representative sample from the discharge paths should be drawn continuously through a
particulate filter. Measurements should be made on these filters to determine the quantities of radionuclides with half-lives greater than 8 days that are released in particulate form to the environment.


3.3  Release Height
(1) The particulate filters should be changed and analyzed at least weekly for the principal gamma-emitting nuclides (at least for the radionuclides barium-lanthanum-140 and iodine-13 1). When quantities of released radioactive materials are at low levels, precluding accurate measurement of principal radionuclides, gross beta radioactivity measurements should be made as a basis for estimating the quantity Of radioactive material released in the week.


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 maximum exposed individual member of the public.  From a dispersion perspective, when determining the maximum 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 /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 () at a more distant location, possibly in a different sector.  See Regulatory Guide 1.111 for a more complete discussion of release height.
(2) A quarterly analysis for strontium-89 and strontium-90 should be made on a composite of all filters from each sampling location collected during the quarter.


3.4  Aquatic Dispersion (Surface Waters)
(3)
A
monthly analysis for gross alpha radioactivity should be made on a composite of all filters collected during the month from each sampling location.


Liquid radioactive effluents may be disposed in accordance with 10 CFR 20.2001, General Requirements, into a variety of receiving surface water bodies, including non-tidal rivers, lakes, reservoirs, settling ponds, 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 flow-through 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 plant- induced, 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. 30), describes calculational models for estimating aquatic dispersion to surface water
The results of these analyses should be used as the basis for recording and reporting the quantities of radioactive material in particulate form released during
,he sampling period. In estimating releases for periods when analyses were not performed, the average of the two adjacent data points spanning this period should be used. These estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.


Rev. 2 of RG 1.21, Page 21 bodies.  However, the dispersion characteristics may be highly site dependent and local characteristics should be considered when performing dispersion modeling and dose assessments.
b.


3.5  Spills and Leaks to the Ground Surface
Sensitivity The sensitivity of analysis for radioactive material in particulate form should be sufficient to permit measurement of a small fraction of the activity which would result in annual exposures of 15 millirems to any organ of an individual in an unrestricted area.


Liquid releases onto the land surface are transported and diluted as a function of site-specific hydrologic features, events, and processes and properties of the effluent.  The releases may temporarily accumulate, pool, or runoff 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 through surface water bodies and considers water usage effects.
4.


Spills or leaks of radioactive material to the ground surface should initiate characterization of the runoff.
Tritium a.


The 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 and NUREG-1302 typically associated with planned effluents (e.g., sampling before discharge to unrestricted areas).  Refer to Regulatory Positions 8.5.1, 8.5.2, and 8.5.9 in this guide for recommendations on the general format for reporting abnormal releases to on-site areas and abnormal discharges to unrestricted areas.
Monitoring The release of tritium to the atmosphere should be determined for each batch released on an intermittent basis, and at least monthly for continuous releases.


3.6  Spills and Leaks to Ground Water
b.


Liquid radioactive leaks and spills are sometimes released to on-site ground water or discharged to offsite ground water.  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 local water table.  The dispersion of this material depends on the local subsurface geology and hydrogeologic 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) and will eventually be released to the unrestricted area.
Sensitivity The sensitivity of analysis of tritium released to the atmosphere should be such that a concentration of  
10-6 pCi/cc (of air) is measurable.


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 abnormal releases of liquid effluents into ground water (see Regulatory Guide 4.1).  References that can be used in developing an adequate ground water site conceptual model include the following: 
B.


1.
LIQUID EFFLUENTS
During the release of radioactive wastes, the effluent control monitor should be set to alarm and to initiate automatic closure of the waste discharge valve prior to exceeding the limits specified in the technical specifications.


ANSI/American Nuclear Society (ANS) 2.17, Evaluation of Subsurface Radionuclide Transport at Commercial Nuclear Power Production Facilities (Ref. 31); 
Continuous monitoring should be provided' for liquid effluent releases. The radionuclide mixture of liquid effluents should be determined and recorded. For the period(s) of release, the records should also show the volume of water used to dilute the liquid effluent and the resultant concentrations at the point(s) of release to unrestricted areas. If the effluent passes into a flowing stream, data on the average flow of the stream during periods of effluent release should be collected and reported in the Supplemental Information section of the report. (See Effluent and Waste Disposal Semiannual Report, Appendix B.)  
The single Poisson (one sigma) error for discrete measurements should be less than 50 percent for release rates at the design objective level, less than 30 percent at twice the design objective release rate, and less than 20
percent at eight times the design objective release rate.


2.
1.


NUREG/CR-6948, Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities and Sites, issued November 2007 (Ref. 32); and 
Batch Releases a.


3.
A representative sample of each batch of liquid effluent released should be analyzed for the principal gamma-emitting radionuclides.


Electric Power Research Institute (EPRI) Report No. 1011730, Ground Water Monitoring Guidance for Nuclear Power Plants, issued September 2005 (Ref.  33).  
1.21-8


4.
When operational or other limitations preclude specific gamma radionuclide analysis of each batch, gross radioactivity measurements should be made to estimate the quantity and concentrations of radioactive material released in the batch, and a weekly sample composited from proportional aliquots from each batch released during the week should be analyzed for the principal gamma-emitting radionuclides.


NUREG/CR-6805, A Comprehensive Strategy of Hydrogeology Modeling and Uncertainty Analysis for Nuclear Facilities and Sites, July, 2003 (Ref 54).  
b.


5.
A
monthly sample composited from proportional aliquots from each batch released during the month should be analyzed for tritium and gross alpha radioactivity.


EPRI Report No. 1015118, Ground Water Protection Guidelines for Nuclear Power Plants, Electric Power Research Institute, Palo Alto, CA, November 2007 (Ref 34).  
c.


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,
A representative sample from at least one representative batch per month should be analyzed for dissolved and entrained fission and activation gases.


Rev. 2 of RG 1.21, Page 22 depth of the release, depth to the local water table, 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 ground water depression(s) that can alter the natural ground water flow.
d.
 
Sites should perform a basic site hydrogeological characterization, in advance of leaks or spills, to be prepared to evaluate potential leaks and spills.  Sites with significant residual radioactivity that are likely to exceed the radiological criteria for unrestricted use at the time of decommissioning (e.g., as described in
10 CFR 20.1402) should perform more extensive evaluation.  Initial assessments should be conducted with relatively simple site conceptual models using scoping surveys and/or bounding assumptions.  The complexity of the models should increase as (1) more knowledge is obtained about the system under evaluation (e.g., source of leak, plume size, concentrations, radionuclides, site characteristics, presence of preferential flow pathways, etc) and as (2) the dose estimates rise above significant residual radioactivity levels (see definition in the glossary).  Industry documents (Refs. 31, 33, and 34) that contain details of various industry practices can be used as part of a ground-water monitoring program.  Sites with low-level spills or leaks generally do not require extensive site characterization and monitoring.
 
Some basic steps in monitoring ground water contamination are summarized below:
 
1.  Use the site conceptual model (as necessary) to assist in monitoring, evaluating, and reporting radioactive releases and radioactive discharges.
 
2.  Collect empirical data by one or more of the following (as necessary):
a.  sample and analyze ground water from existing monitoring wells, and b.  conduct additional hydrogeologic testing using existing wells (or new wells) if required.
 
3.  Test the site conceptual model and radionuclide transport predictions using groundwater sample results and data collected during hydrogeologic testing.
 
4.  Modify site conceptual model and radionuclide transport parameters as necessary to predict discharges and assess doses to members of the public.
 
5.  Return to step 1.
 
The ground water monitoring results should be used in the development and testing of a site conceptual model to predict radionuclide transport in ground water.  A more thorough discussion is contained in the references listed in section C.3.6.  The site conceptual model is generally considered adequate when it predicts the results of monitoring (sometimes called a calibrated model). Ground water monitoring results are used to evaluate the validity of the site conceptual model.  Following a leak or spill of contaminated material, the site conceptual model may be used in conjunction with radionuclide transport modeling and ground water monitoring to comprise a basis for predicting future effluents from the site.  Account should be taken of dispersion and dilution that occurs over time and in three dimensions.


The site conceptual model together with a strategic and carefully planned monitoring program can ensure that necessary and reasonable surveys are performed (i.e., limited scoping surveys or more extensive surveys).  Limited scoping surveys should be performed to determine if significant residual radioactivity exists and to determine if there is adequate protection of public health and safety.  If the limited scoping surveys identify significant residual radioactivity, then the extent of the contamination
A
 
quarterly sample composited from proportional aliquots from each batch released during the three-month period should be analyzed for strontium-89 and strontium-90.
Rev. 2 of RG 1.21, Page 23 should be further evaluated by more extensive surveys (e.g., monitoring wells or other evaluations as appropriate). These survey 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 and the mobility of the radionuclides.  For spills or leaks occurring below the soil surface in inaccessible locations, direct scoping and characterization may not be feasible.  In these cases, indirect monitoring techniques (e.g., ground water monitoring wells in a down gradient direction)
should be used to satisfy existing regulatory requirements.  These survey activities should, at a minimum, satisfy (1) the requirements of 10 CFR 50.75(g) and (2) the effluent reporting requirements of
10 CFR 50.36a for ground water discharges to the unrestricted area.  In general, leaks and spills of radioactive material should be described (reported) in the ARERR for the calendar year the spill or leak occurred.  Additionally, ground water monitoring data should be reported in the ARERR for the calendar year in which the data were collected.  Refer to Regulatory Positions 8.5.1, 8.5.2, and 8.5.9 of this document for guidance on the general format for reporting abnormal releases to on-site areas and abnormal discharges to unrestricted areas.
 
Although licensees may conduct a ground water monitoring effort for different reasons, for purposes of this regulatory guide, the surveys, characterization activities, site conceptual models, and other components of any ground water monitoring effort should be sufficient to do the following: 
 
1.


appropriately report, for purposes of accountability, effluents discharged to unrestricted areas, 
The results of these analyses should be used as the basis for recording and reporting the quantities of radioactive material released in liquid effluents during the sampling period. In estimating releases for a period when analyses were not performed, the average of the two adjacent data points spanning this period should be used. Such estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.


2.
2.


document information in a format consistent with Table A-6 and Regulatory Position 8.5, 
Continuous Releases For continuous releases (e.g., secondary plant leakage),
in addition to continuous monitoring, a representative sample of the liquid effluent should be analyzed at least weekly to determine the identity and quantity of the principal gamma-emitting radionuclides being released. Analysis for other specific radionuclides should be conducted in accordance with 1 above.


3.
3.


provide advance indication of potential future discharges to unrestricted areas (to ensure releases are planned and monitored before discharge), 
Sensitivity The sensitivities of analyses of radioactive materials in liquid effluents should be sufficient to permit the measurement of concentrations of IffipCi/mri by gross radioactivity measurements, 5 x 10-7 yCi/ml of 1icTh gamma-emitting radionuclide, 10' uCi/nml of each of the dissolved and entrained gaseous radionuclides,  
 
10-'
4.
puCi/ml of gross alpha radioactivity, Ilff
 
/Ci/mli of tritium, and 5 x 10-8 pCi/ml of strontium-89 and strontium-90.
demonstrate that significant residual radioactivity has not migrated off site to an unrestricted area in the annual reporting interval, and 
 
5.
 
communicate pertinent information to the NRC.
 
===4. Quality Assurance ===
 
4.1  Regulatory Guidance 
 
A range of QC checks and tests should be applied to the analytical process.  Regulatory Guide 4.15, Revisions 1 and 2, describe the QA program activities for ensuring that radioactive effluent monitoring systems and operational programs meet their intended purpose.  Each licensees licensing basis determines the applicability of Revision 1 or Revision 2.  Licensees with programs in operation before the issuance of Regulatory Guide 4.15, Revision 2, may rely exclusively on Revision
 
===1. Regulatory Guide ===
4.15, Revision 2, contains guidance on determining appropriate sensitivity levels for analytical instrumentation based on data quality objectives (DQOs).  The use of DQOs may provide a better technical basis for determining sensitivity levels (LLDs) than the use of the default values supplied in NUREG-1301 and NUREG-1302.  A combination approach (using both Revision 1 and Revision 2 of Regulatory Guide 4.15) can be used to determine appropriate sensitivity levels (LLDs) different (i.e., higher or numerically larger) than those listed in NUREG-1301 and NUREG-1302.


Rev. 2 of RG 1.21, Page 24
C. SOLID WASTE
4.2  Quality Control Checks 
The total curie quantity and radionuclide composition of the solid waste shipped offsite should be determined. Provisions should be made to monitor and to limit the curie quantity of material and the maximum radiation level of each package of solid waste in order to reduce radiation exposure to personnel and to meet the regulatory requirements of 10 CFR Part 71, "Packaging of Radioactive Material for Transport and Transportation of Radioactive Material under Certain Conditions," and of the Department of Transportation.


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 working condition and functioning properly.  QC records should include results of routine tests and checks, background data, calibrations, and all routine maintenance and service.
Monitoring of solid wastes in storage and preparatory to shipment should be performed to provide assurance that the radiation levels from waste in storage and in transport do not exceed regulatory limits.


4.3  Functional Checks 
1.21-9


Routine qualitative tests and 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 traceable by the National Institute of Standards and Technology (NIST). The schedule for source checks, channel checks, channel calibrations, and channel operational tests should be in accordance with NUREG-1301 and NUREG-1302.
APPENDIX B
EFFLUENT AND WASTE DISPOSAL REPORT
This appendix describes the data and information that should be included in effluent and waste disposal reports. The data and information should be reported in a format similar to that given in Tables 1 through 4 and the Supplemental Information sheet. Except as noted, effluent and solid waste data should be summarized on a quarterly basis, although in some cases more detailed data may be needed. The need fcr reporting of additional data to the Commission will be determined on a case-by-case basis.


4.4  Procedures 
The reporting method includes the use of uniform notation for numerical values and generally defined guidance for reporting certain supplemental information.


Individual written procedures should be used to establish specific methods of calibrating installed radiological monitoring systems and grab sampling equipment.  Written procedures should document calibration practices used for ancillary equipment and systems (e.g., meteorological equipment, airflow measuring equipment, in-stack monitoring pitot tubes).  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.
Data from licensee's effluent and waste disposal reports are compiled, and summary reports of nuclear power plant effluents are prepared by the Commission. The supplemental information reduces errors in processing and compiling of report data.


Calibration procedures should identify the specific equipment or group of instruments to which the procedures apply.
In the report, a separate section should contain a discussion of the radiological impact of facility operation on man. Calculations and estimates of potential doses to individuals and population doses should be summarized for the report (6-month) period, although in some cases more detailed data may be needed. The need for these additional data to be reported to the Commission is determined on a case-by-case basis.


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.
Meteorological data during continuous releases should be submitted in the format presented in Table
4A. (Also see Regulatory Guide 1.23.) Data on meteoro logical conditions during batch releases should be reported separately in the same format. For the purpose of this guide, abnormal releases should be treated as batch releases, and the meteorological data obtained during abnormal releases should be included in the batch release report.


4.5  Calibration of Laboratory Equipment and Radiation Monitors 
A.
 
Calibrations (e.g., of laboratory equipment and continuous radiation monitoring systems used to quantify radioactive effluents) 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 inservice 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 employing a method provided by the instrument manufacturer.
 
Rev. 2 of RG 1.21, Page 25
 
4.6  Calibration of Measuring and Test Equipment 
 
Measuring and test equipment 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.
 
4.7  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 (an increase or decrease) 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.
 
4.8  Measurement Uncertainty
 
The measurement uncertainty (formerly called measurement error) associated with the measurement of radioactive materials in effluents should be estimated.  Counting statistics can provide an estimate of the statistical counting uncertainty involved in radioactivity analyses.  Because it may be difficult to assign error terms for each parameter affecting the final measurement, detailed statistical evaluations of error are not required.  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 measurement uncertainty associated with the effluent measurement should ideally include the cumulative uncertainties resulting from the total operation of sampling and measurement.  Expanded uncertainty should be reported with measurement results.  The objective should be to evaluate only the important contributors and obtain a reasonable measure of the uncertainty associated with reported results. Detailed statistical and experimental evaluations are not required.  The overall objective should be to obtain an overall estimate of measurement uncertainty.  The formula for calculating the total or expanded uncertainty classically includes the square root of the sum of squares of each important contributor to the measurement uncertainty.  Licensees may obtain additional information from NUREG-1576 and ANSI/HPS N13.1-1999 if there is a need to improve the estimate of uncertainty.
 
5.  Dose Assessments for Individual Members of the Public
 
The regulation in 10 CFR 20.1301 establishes dose limits for individual members of the public.
 
The regulations referenced in Regulatory Positions 5.4 through 5.6 contain both dose limits and design objectives that the licensee demonstrates compliance with through calculations.  Table 1 summarizes the fundamental parameters associated with the dose calculations.  Regulatory Positions 5.7 and 5.8 present important concepts for these calculations.  Because of differences between NRC and EPA regulations, only demonstrating compliance with radiological effluent technical specifications (based on Appendix I to
 
Rev. 2 of RG 1.21, Page 26
10 CFR Part 50) does not necessarily ensure compliance with EPAs 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations (Ref. 35), particularly if there is a direct radiation component (e.g., from BWR shine, ISFSI, or radioactive materials storage). 
 
Table 1.  Parameters Associated with Dose Calculations
 
10 CFR Part 50, Appendix I
10 CFR 20.1301(e)
(EPA 40 CFR Part 190)
Dose Whole Body, Max of Any Organ, Gamma Air, and Beta Air Whole Body, Thyroid, and Max of Any Organ Basis ICRP-2 EPA 40 CFR Part 190
Where Unrestricted Area Unrestricted Area Individual Receptor Real Person/Exposure Pathway (nearest real residence, real garden, real dairy/meat animal)
Real Person/Exposure Pathway (nearest real residence, real garden, real dairy/meat animal)
Origin Liquid and Gas Radioactive Waste Liquid and Gas Radioactive Waste Direct Radiation (e.g., shine, nitrogen-16, ISFSI, radioactive materials storage, outside tanks)
Accumulated Radioactive Material (e.g.,
tritium in lake water) Not Already Included in Dose Estimates Radioactive Material Licensed Only Licensed and Unlicensed When Current year Current and Prior Years Operation
 
5.1  Bounding Assessments 
 
Bounding assessments may be useful in those circumstances where compliance can be readily demonstrated using conservative assumptions.  For purposes of this document, the term bounding assessment means that the reported value is unlikely to be substantially underestimated (see 10 CFR 50
Appendix I, Section III). Bounding assessments for the current year do not imply the absolute bounds for future conditions.
 
For example, licensees may use conservative bounding dose assessments in lieu of site-specific dose assessments of the maximum dose to individual members of the public.  Instead of assessing dose from ground level effluent releases to a real individual member of the public located 2 miles from the site boundary, a conservative bounding dose assessment can be performed for a hypothetical individual located at the site boundary.
 
If bounding assumptions are made, the radioactive effluent release report should state such and should annotate the assumptions.  Hypothetical exposure pathways and locations are sometimes used for bounding dose assessments (or hazard evaluations done in accordance with 10 CFR 20.1501).  See the definition of hypothetical exposure pathway in the glossary.
 
Rev. 2 of RG 1.21, Page 27
5.2  Individual Members of the Public 
 
Individual 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 location) as either members of the public or as occupational workers.  (See definition of members of the public in 10 CFR Part 20.)  The annual dose limits for members of the public in the unrestricted area are 25 millirem (mrem) whole body and 75 mrem to the thyroid and 25 mrem to any other organ in accordance with the EPA regulations in 40 CFR Part 190; the limits are 100 mrem in accordance with 10 CFR 20.1301.  In effect, annual dose limits to members of the public while in the unrestricted area are the EPA limits of 25 mrem whole body and 75 mrem to the thyroid and 25 mrem to any other organ;
whereas the annual dose limit for a member of the public in the licensees controlled area is the NRCs total effective dose equivalent limit of 100 mrem.
 
If bounding assessments are not used, licensees should perform evaluations to determine the dose to a real, maximum exposed 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 consider only members of the public in the unrestricted area.  A member of the public is typically 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) and is typically not a fictitious fencepost resident or an exposure pathway that includes a virtual goat or cow.  Licensees are encouraged (but not required) to use real individual members of the public when performing dose assessments for radioactive discharges.  Table 1 in Regulatory Guide 1.109 allows a dose evaluation to be performed at a location where an exposure pathway and dose receptor actually existed at the time of licensing.
 
5.3  Occupancy Factors 
 
For members of the public in the unrestricted area, occupancy factors should be assumed to be
100 percent at locations identified in the land use census, unless site-specific information indicates otherwise.  Occupancy factors may be applied inside the controlled area based on estimated hours spent in the controlled area.
 
5.4  10 CFR Part 50, Appendix I 
 
Appendix I to 10 CFR Part 50 contains numerical guidance for design objectives and limiting conditions of operation for radioactive waste systems to ensure discharges of 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 (see below).  License technical specifications require that exposure to liquid and gaseous effluents conform to the numerical guidance in 10 CFR Part 50, Appendix I.  Per 10 CFR 50.34a, Design Objectives for Equipment to Control Releases of Radioactive Material in EffluentsNuclear Power Reactors, 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 following terms are generally used: 


SUPPLEMENTAL INFORMATION
1.
1.


air doses (gamma and beta), total body doses, and organ doses (based on International Commission on Radiation Protection (ICRP)-2, Report of Committee II on Permissible Dose for Internal Radiation, issued 1959 (Ref. 36)); 
Regulatory Limits The technical specification limits for radioactive materials released in liquid and gaseous effluents should be included in each report. If changes are made in limiting conditions of operation during the report period, the appropriate limits and dates should be included.


2.
2.


effluent discharges only (excludes direct radiation from the facility and ISFSIs); 
Maximum Permissible Concentrations The maximum permissible concentrations (MPC)
used to calculate permissible release rates and concentrations for air and water should be included in each report (if appropriate), i.e., the MPC used in accordance with technical specifications and/or derived from the use of Notes to Appendix B, 10 CFR Part 20.


Rev. 2 of RG 1.21, Page 28
3.
3.


current annual period (excludes accumulated radioactivity from prior-year effluents); and
Average Energy The release rate limits for fission and activation gases in gaseous effluents are usually based on the average energy (FE) of the radionuclide mixture in the effluent. The E value for the gamma and beta energies per disintegration that is used should be included in the report.


4.
4.


unrestricted area (excludes individuals in the restricted areas and controlled areas).  
Measurements and Approximations of Total Radioactivity A summary description should be provided of the method(s)
used to determine or measure total radioactivity in effluent releases (total here means the overall gross curie quantity). For example, gross radioactivity measurements (gross beta and/or gross gamma) may be used to approximate total radioactivity
'n effluents, and/or analyses of specific radionuclides in selected or composited samples may be used to determine the radionuclide composition of the effluent.


When calculating air doses licensees should assure that for any location outside the site boundary doses do not exceed the 10 CFR 50 Appendix I design objectives.  Calculation of air dose at the site boundary would assure the most conservative calculation of air doses for ground-level releases.  This may not be true for elevated releases.  Licensees should select a location that assures the most conservative calculation of air dose.
A summary description of the methods used for estimating overall errors associated with radioactivity measurements should also be provided.
 
5.5  10 CFR 20.1301(a) through (c) 
 
This regulation specifies dose limits for members of the public from licensed operation of the facility.  These limits apply to doses resulting from licensed and unlicensed radioactive material and from radiation sources other than background radiation (see 10 CFR 20.1001, Purpose).  Demonstration of compliance with the limits of 40 CFR Part 190 will be considered to also demonstrate compliance with the
0.1 rem total effective dose equivalent limit of 10 CFR 20.1301(a) (Ref. 37). 
 
5.6  10 CFR 20.1301(e) 
 
For those facilities subject to EPAs generally applicable environmental radiation standards promulgated in 40 CFR Part 190, 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 that are not already included in items 1 and 2.
 
These requirements include the following considerations: 
 
1.
 
Whole body and organ doses (ICRP-2 concepts).
 
2.
 
Any member of the public means any individual except when that individual is receiving an occupational dose.
 
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 is the area outside the site boundary, excluding the controlled area and the restricted area.  (See the definition of generally applicable environmental radiation standards in
10 CFR 20.1003, Definitions.)
 
4.
 
Current-year effluents includes both normal and abnormal discharges to the unrestricted area.


5.
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.
Batch Releases The report should provide information relating to batch releases of liquid and gaseous effluents which are discharged to the environment. This information should include the number of releases, total time period for batch releases, and the maximum, mean, and minimum time period of release.


6.
6.


Cumulative dose means the sum of (1) current-year effluent dose, (2) current-year direct radiation dose, and (3) dose from accumulated radioactivity if not already included in the first two categories.
Abnormal Releases The number of abnormal releases of radioactive material to the environment should be reported. The total curies of radioactive materials released as a result of abnormal releases should be included.


Rev. 2 of RG 1.21, Page 29
This information should be reported separately for liquid and gaseous releases. The activity values should also be included, as appropriate, in Tables 1 and 2.
7.


Accumulated radioactivity includes radioactive material in the unrestricted area from prior-year discharges that remains in the environment (e.g., tritium in lake water or radionuclides).  
Hourly meteorological data should be recorded for the periods of actual release and included in the quarterly summaries for batch releases in the format given in Table
4A.


8.
B. GASEOUS EFFLUENTS
Summary information should be reported in the formats of Tables IA through IC. Table IA values
1.21-10


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 (see definition of uranium fuel cycle in 40 CFR Part 190, also in Glossary of this document).  
should include the sums of all sources of release, i.e.,  
routine and abnormal releases, continuous and batch, elevated and ground level. The reported percent of technical specification limits should be based on the combined releases from multiple sources as given in the technical specifications. This also applies to the releases from multireactor sites.


5.7  Dose Assessments for 10 CFR Part 50, Appendix I
For reactors that have technical specification limits for more than one principal point of release, separate radionuclide data should be reported for each of these release points. Data should be separated by release height, i.e., elevated or ground level, and these data should be further subdivided by release mode, i.e.,  
continuous or batch mode. (See Tables lB and IC.)
Estimates of the total error associated with certain total values should be provided in each report. (See Table 1A.) These error values should be the best effort at an overall estimate of the errors associated with the totals in the report.


Dose assessments to show compliance with technical specification requirements for meeting the numerical values of 10 CFR Part 50, Appendix I, design objectives should include quarterly and annual doses using the considerations of Regulatory Position 5.4. They should be reported in a format similar to that shown in Table A-4 in the appendix to this regulatory guide and include the items listed below:
Report the following information as indicated by Tables IA through IC.


1.
1. Gases a.


doses from liquid effluents a.
Quarterly sums of total curies of fission and activation gases released.


total body dose, quarterly and annual, b.
b.


organ dose, quarterly and annual (maximum, any organ), and c.
Average release rates (guCi/sec) of fission and activation gases for the quarterly periods covered by the report.


percent of limits for each of the above.
c.


2.
Percent of technical specification limit for releases of fission and activation gases. This should be calculated in accordance with technical specification limits.


doses from gaseous effluents a.
d.


beta and gamma air doses, quarterly and annual, b.
Quarterly sums of total curies for each of the radionuclides determined to be released, based on analyses of fission and activation gases. The data should be categorized by (1) elevated releases, batch and continuous modes, and (2) ground-level releases, batch and continuous modes. (See Tables lB and IC.)
2.


organ dose commitment from iodine, tritium, and particulate releases with half-lives greater than 8 days, quarterly and annual, and c.
lodines a.


percent of limit for each of the above.
Quarterly sums of total curies of iodine-131 released.


An evaluation of the local exposure pathways to determine the maximum exposed member of the public should be performed.  However, maximum doses from various exposure pathways are not additive from different locations.  For example, dose from a downstream drinking water exposure pathway should not be added to the dose to an upstream resident whose exposure is from gaseous effluents and direct radiation unless that individuals drinking water is obtained from the down stream location.
b.


Maximum doses to real individuals are assessed as described in Regulatory Guide 1.109.  The locations and exposure pathways are those where real 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 assumptions for food consumption and occupancy factors at actual locations when assessing dose to the maximum exposed individual, unless they have determined and applied site-specific (actual) data.  In lieu of assessing dose to real individuals, bounding dose assessments may also be used for compliance with
Average release rate (pCi/sec) of iodine-131.
10 CFR Part 50, Appendix I (see the section titled Bounding Assessments).  


The objective of Appendix I is to provide numerical guides for design objectives and limiting conditions for operation to ensure that radioactive 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 discharges.  The Appendix I related calculations do not include dose from radioactivity in prior-year, accumulated, effluent discharges (e.g., last years radioactivity remaining in lake water is excluded).  Note:  However, the dose calculations for
c.


Rev. 2 of RG 1.21, Page 30
Percent of technical specification limit for iodine-131.
demonstrating compliance with the EPA limits do include accumulated radioactivity.  (See Section 5.8 below.)


The exposure pathways and routes of exposure identified in Regulatory Guide 1.109 and other exposure pathways and routes of exposure that may arise because of unique conditions at a specific site should be considered if they are likely to contribute significantly to the total dose.  Other exposure pathways are considered significant if a conservative evaluation yields an additional dose increment equal to or more than 10 percent of the total from all exposure pathways considered in RG 1.109 (see the regulatory position C in Regulatory Guide 1.109).  An evaluation of other exposure pathways (not included in dose assessments) should be performed and maintained for purposes of demonstrating compliance with staff position C in Regulatory Guide 1.109.  A thoroughly designed and documented evaluation of a less significant release point could also assist in the evaluation and characterization of abnormal releases and abnormal discharges.
d.
 
Real exposure pathways are identified for routine discharges and direct radiation based on the results of the land use census.  Dose calculations should typically be performed based on real exposure pathways.  Conversely, dose assessments (i.e., surveillances and dose calculations) are not needed for exposure pathways that do not exist at a site.  For example, if the land use census does not identify the existence of an ingestion exposure pathway involving a milk animal, the licensee is not required to assess that route of exposure for the ingestion exposure pathway.  Similarly, if a licensee discharges liquid radioactive waste to a body of water (either surface water or ground water) and that body of water is not used as a source of drinking water (either private or public), a drinking water assessment is not required.
 
For purposes of reporting information in the ARERR, there is a distinction between dose assessments for Appendix I to 10 CFR Part 50 and hazard assessments that may be conducted for on-site spills and leaks as outlined in 10 CFR 20.1501 (where bounding estimates may be necessary).  (See bounding dose estimates in Section 5.1.)
 
5.8  Dose Assessments for 10 CFR 20.1301(e) 
 
To show compliance with 10 CFR 20.1301(e), dose assessments should be reported according to the generally applicable environmental radiation standards promulgated by EPA at 40 CFR Part 190, with consideration of Regulatory Position 5.6 and in a format similar to that shown in Table A-5 of the appendix to this guide.
 
5.8.1 The following should be reported:
1.
 
whole body dose to the maximum individual member of the public
2.


thyroid dose to the maximum individual member of the public
Quarterly sums of total curies of each of the isotopes, iodine-131, iodine-133, and iodine-135 determined to be released. (See B.1.d above and Tables lB and 1C.)
3.
3.


dose to any other organ to the maximum individual member of the public
Particulates a.
4.


percent of the applicable limit
Quarterly sums of total curies of radioactive material in particulate form with half-lives greater than 8 days determined to be released.


5.8.2 One means of demonstrating compliance with 40 CFR Part 190 is listed in the Federal Register
b.
(42 FR 2859), (Ref. 38), which states the following:


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 Part 190].  
Average release rate (pCi/sec) of radioactive material in particulate form with half-lives greater than 8 days.


As a result, a licensee who (1) can demonstrate that external sources of direct radiation are indistinguishable from background and who (2) demonstrates compliance with the numerical dose
c.


Rev. 2 of RG 1.21, Page 31 guidance of 10 CFR Part 50, Appendix I, may cite the above reference as the basis for demonstrating compliance with 40 CFR Part 190.
Percent of technical specification limit for radioactive material in particulate form with half-lives greater than 8 days.


However, licensees who (1) have external sources of direct radiation that are above background and (2) demonstrate compliance with the numerical dose guidance of 10 CFR Part 50, Appendix I, must also include sources of direct radiation from uranium fuel cycle operations (e.g., including direct radiation from the licensed facility as well as co-located or nearby nuclear power facilities if appropriate).  
d.


5.8.3 The dose contributions from direct radiation may be estimated based on either (1) direct radiation measurements (e.g., thermoluminescent dosimeters, optically stimulated devices, or integrating portable ion chambers), (2) calculations, or (3) a combination of measurements and calculations. When direct radiation dose is determined by measurement, estimates of background levels of radiation may be subtracted based on selected control locations. The doses measured from control and indicator locations should be taken from the same time period.  When choosing the appropriate control location(s), licensees should consider the historical variability in doses measured at the control and indicator locations. Several sources contain additional information regarding background subtraction for thermoluminescent dosimeters (Refs. 39, 40, 41, and 42).  Methods of determining dose from direct radiation to the maximum exposed individual member of the public may also include extrapolation methods.
Quarterly sums of total curies for each of the radionuclides in particulate form determined to be released based on analyses performed. (See B.l.d above and Tables IB and IC.)  
e.


Licensees must demonstrate compliance with 10 CFR 20.1301(e) for the generally applicable environmental radiation standards promulgated in 40 CFR Part 190.  These include the concept of a total dose (to the whole body and to any organ) from all sources related to the uranium fuel cycle.
Quarterly sums of total curies of gross alpha radioactivity determined to be released.


Contributions to the total dose from radioactive effluents (liquid and gaseous) and direct radiation should be included, if applicable.  Other sources (e.g., accumulated radioactive materials in offsite ponds or lakes from previous years discharges) should also be included, if applicable, when estimating the total dose.  However, if the contributions from direct radiation or accumulated radioactivity are generally minor (as evaluated and documented in a licensee technical evaluation as not contributing to the total dose), these contributions need not be included in the total dose evaluation, but the basis for exclusion should be documented.
4.


5.9  Dose Calculations 
Tritium a.


Acceptable dose assessment models, such as those provided in Regulatory Guides 1.109, 1.111,
Quarterly sums of total curies of tritium determined to be released in gaseous effluents.
1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light- Water-Cooled Power Reactors, (Ref. 43) and 1.113, should be used to make dose calculations.  When calculating organ doses from airborne effluents, contributions from I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days should be included in the assessment.


6. Solid Radioactive Waste Shipped for Processing or Disposal 
b.


Solid radioactive waste shipments should be reported in a format similar to that of Table A-3 in Appendix A to this guide.  The data should be divided by waste classification and by the waste stream categories listed in Table A-3.  The waste streams are (1) resins, filters, and evaporator bottoms, (2) dry active waste, (3) irradiated components, and (4) other waste.  The data reported should be for the low-level waste (LLW) volumes shipped from a plant site for waste processing or disposal (not the radioactive waste volumes that are ultimately buried).  
Average release rate (pCi/sec) of tritium.


Rev. 2 of RG 1.21, Page 32 Note:  Data on LLW disposed in licensed LLW disposal facilities is available using the Manifest Information Management System (MIMS) operated by the Department of Energy.  There are no requirements for reporting storage of LLW at nuclear power plants.  However, LLW storage records are maintained at nuclear plants and are available for NRC inspection during routine effluent inspections.
c.
 
Shipments that do not need to be reported include shipments of metal melt, contaminated equipment for transfer between licensees or equipment for refurbishment, contaminated laundry (either launderable or dissolvable), or radioactive samples for analysis.  Potentially contaminated dry active waste sent for resurvey and segregation (sometimes referred to as green is clean) does not need to be reported.
 
Equipment shipped for decontamination and free release does not need to be reported.  However, records of these types of shipments should be maintained on site.


The total curie quantity and major radionuclides in the solid waste shipped off site should be determined and reported in a format similar to that of Table A-3.
Percent of appropriate technical specification or MPC limits for tritium.


7. Reporting Errata in Effluent Release Reports 
C. LIQUID EFFLUENTS
Summary information should be reported in the formats of Tables 2A and 2B. Table 2A values should include the quarterly sums of all releases of radioactive materials in liquid effluents, i.e., routine and abnormal occurrences, continuous and batch. The reported percent of technical specification limits should be based on the combined releases from multiple sources as given in the technical specifications. This also applies to the releases from multireactor sites.


Errors in radioactive effluent release reports should be classified and reported as described below.
Estimates of the total error associated with certain total values should be provided in each report. (See Table 2A.) These error values should be the best effort at an overall estimate of the errors associated with the totals in the report.


7.1  Examples of Small Errors 
Report the following information, as indicated by Tables 2A and 2B.
 
Small errors may be any of the following:


1.
1.


inaccurate reporting of dose that equates to 10 percent of the applicable 10 CFR 50
Mixed Fission and Activation Products a.
Appendix I design objective or 10 percent of the EPA public dose criterion, 


2.
Quarterly sums of total curies of radioactive material determined to be released in liquid effluents (not including tritium, dissolved and entrained gases, and alpha-emitting material). (See Table 2A.)
b.


inaccurate reporting of curies (or release rates, volumes, etc.) that equate to 10 percent of the affected curie total (or release rate, volume, etc.), after correction; 
Average concentrations (juCi/ml) of mixed fission and activation products (C.1 .a above) released to unrestricted areas, averaged over the quarterly periods covered by the report.


3.
c.


omissions that do not impede the NRCs ability to adequately assess the information supplied by the licensee, or 
Percent of applicable limit of average concentrations released to unrestricted areas (C.l.b above). Include the limit used and the bases in the supplemental report information.


4.
d.


typographical errors or other errors that do not alter the intent of the report.
Quarterly sums of total curies for each of the radionuclides determined to be released in liquid effluents, based on analyses performed. Data should be separated by type of release mode, i.e., continuous or batch. (See Table 2B.)
1.21-11


7.2  Reporting Small Errors 
2.


Small errors should be corrected within one year of discovery, and the correction may be submitted with the next (normally scheduled) submittal of the ARERR as follows.  A brief narrative explanation of the errors should be included in Section 8, Errata/Corrections to Previous ARERRs, of Table A-6, Supplemental Information.  The narrative should include a statement that the affected pages, in their entirety, are included as attachments to the ARERR.  Additionally, the affected, corrected pages, in their entirety, should be submitted as an attachment (or addendum) to the ARERR.  The corrected pages should reference the affected calendar year and should contain revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections to multiple ARERRs, make a separate attachment (or addendum) for each of the affected years.  Other methods of correcting previous ARERRs may be used provided the corrections are clearly and completely described.
Tritium a.


Rev. 2 of RG 1.21, Page 33
Quarterly sums of total curies of tritium determined to be released in liquid effluents.
7.3  Examples of Large Errors 


Large errors may be any of the following:
b.


1.
Average concentrations (pCi/ml) of tritium released in liquid effluents to unrestricted areas, averaged over the quarterly periods covered by the report.


inaccurate reporting of dose that equates to >10 percent of the Appendix I or EPA
c.
public dose criterion, after correction; 
 
2.


inaccurate reporting of curies (or release rate, volume, et
Percent of applicable limit of average concentrations released to unrestricted areas (C.2.b above), i.e., percent of 3 x 10-3 pCi/mi. Include the limit and the bases in the supplemental report information.
 
====c. that equate to ====
>10 percent of the affected curie total (or release rate, volume, etc.), after correction; "


3.
3.


omissions that may impede the NRCs ability to adequately assess the information supplied by the licensee; and  
Dissolved and Entrained Gases a.


4.
Quarterly sums of total curies of gaseous radioactive material determined to be released in liquid effluents.


typographical errors or other errors that do significantly alter the intent of the report.
b.


7.4  Reporting Large Errors 
Average concentrations (pCi/ml) of dissolved and entrained gaseous radioactive material released to unrestricted areas, averaged over the quarterly periods covered by the report.


Large errors should be corrected within 90 days of discovery.  The correction may be made by special submittal or may be submitted with the next (normally scheduled) ARERR (if the next ARERR is to be submitted within 90 days of discovery of the error).  If corrections are made by special submittal, include a brief narrative explaining the errors.  The narrative should include a statement that the affected pages, in their entirety, are included as an attachment.  Attach the affected, corrected pages, in their entirety.  The corrected pages should reference the affected calendar year and should contain revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections to multiple ARERRs, make a separate attachment (or addendum) for each of the affected years.  If corrections are made coincident with the next (normally scheduled) submittal of the ARERR, use the correction process as specified in section 7.2 (for small errors) above.  Other methods of correcting previous ARERRs may be used provided the corrections are clearly and completely described.
c.


8. Format and Content of the Annual Radioactive Effluent Release Report 
Percent of technical specification limit of average concentrations released to unrestricted areas (C.3.b above). Include the limit used and the bases in the supplemental report information.


In accordance with 10 CFR 50.4, Written Communications, the annual report should be submitted electronically or in a written communication.  The report should consist of a summary of the numerical data in a tabular format similar to Tables A-1 through A-5 in Appendix A to this guide.
d.


Effluent data reported in Tables A-1, A-1A through A-1F, A-2, A-2A, A-2B, and A-4 should be summarized on a quarterly and annual basis.  Tables A-3 and A-5 should be summarized on an annual basis.  In addition to numerical data, additional supplemental information should be included containing all the information in (but not necessarily in the format of) Table A-6.  Additional detail for the information contained in each of these tables is listed below.  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.
Quarterly sums of total curies for each of the radionuclides determined to be released as dissolved and entrained gases in liquid effluents.


Radionuclides that are not detected for the entire reporting period do not need to be listed in the tables (Tables A-1A through A-1F, A-2A, and A-2B). Activity that is detected should be reported in the appropriate tables (i.e., Tables A-1, A-2, A-1A through A-1F, A-2A, and A-2B) in the ARERR, provided that the amount discharged is numerically significant with respect to the three-digit exponential format recommended for the ARERR.  This should not be confused with three significant figure
4.


====s. Licensees may ====
Alpha Radioactivity Quarterly sums of total curies of gross alpha-emitting material determined to be released in liquid effluents.


Rev. 2 of RG 1.21, Page 34 round numbers according to accepted practices (e.g., refer to ASTM E-29, Practice for Using Significant Digits in Test Data to Determine Conformance with Specifications (Ref. 48)); however, after rounding has been completed, values should be reported in the ARERR in a three-digit exponential format.
5.


Measurements should be reported for positive values.  Some radionuclides that are detected in a year may not be detected in all quarters.  If results are determined to be below detectable levels for an entire quarter, the table entry should include a suitable designation (e.g., N/D and an accompanying footnote) to denote that measurements were performed but no activity was detected.
Volumes a.


The format specified in Revision 2 of this regulatory guide differs slightly from that specified in Revision 1 of Regulatory Guide 1.21.  The format and content as specified in Revision 2 are one acceptable method of reporting the data.  Other formats may be used (e.g., some tables may be combined)
Quarterly sums, in liters, of total measured volume, prior to dilution, of liquid effluent released.
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.  This format is designed to be consistent with some commonly used electronic-data-reporting software packages.  Consistency aids review by members of the public and allows easier industry-wide comparisons of the data.


8.1  Gaseous Effluents 
b.


The quarterly and annual sums of all radionuclides discharged in gaseous effluents (i.e., routine and abnormal discharges, continuous, and batch) should be reported in a format similar to that of Tables A-1A through A-1F in Appendix A to this regulatory guide.  The data should then be further summarized and reported in the format of Table A-1.  Additional information on each of these tables is provided below.
Quarterly sums of total determined volume, in liters, of dilution water used during the period of the report.


Table A-1, Gaseous Effluents - Summation of All Discharges, contains a summation of all gaseous effluent discharges from all release points and all modes of release.  The data are 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.
6.


Table A-1A, Gaseous EffluentsGround-Level ReleaseBatch Mode, contains a summation of gaseous effluent releases from ground-level release points in the batch mode of release for five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha. Licensees should report the following:
Stream Flow Where the effluent passes into a flowing stream, data on the average flow of the stream during periods of effluent release should be collected and reported in the Supplemental Information section of the report.


1.
D. SOLID WASTE
The following information should be reported for shipments of solid waste and irradiated fuel transported from the site during the report period:
1. The semiannual total quantity in cubic meters and the semiannual total radioactivity in curies for the categories or types of waste. (See Table 3.)
a.


curies of each radionuclide discharged by quarter and year, and
Spent resins, filter sludges, evaporator bottoms;
2.
b.


total curies discharged in each radionuclide category (fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha) by quarter and year.
Dry compressible waste, contaminated equipment, etc.;
c.


Some licensees may have surveillance requirements allowing the non-noble gas radionuclides (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with continuous release results.  In these instances, the table entries for the affected radionuclides for batch releases should include an appropriate designation (e.g., *) and an accompanying footnote describing this situation.
Irradiated components, control rods, etc.;
 
d.
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 five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha.  Licensees should report the following:
 
Rev. 2 of RG 1.21, Page 35
 
1.


curies of each radionuclide discharged by quarter and year, and  
Other (furnish description).  
2.
2.


total curies discharged in each radionuclide category by quarter and year.
An estimate of the major nuclide composition in the categories of waste in D.1 above.


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 five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha.  Licensees should report the following:
3.


1.
The disposition of solid waste shipments. (Identify the number of shipments, the mode of transport, and the destination.)
4.


curies of each radionuclide released by quarter and year, and 
The disposition of irradiated fuel shipments.
2.


total curies released in each radionuclide category by quarter and year.
(Identify the number of shipments, the mode of transport, and the destination.)
Estimates of the total error associated with certain total values should be provided in each report. (See Table 3.) These error values should be the best effort of an overall estimate of the errors associated with the totals in the report.


Some licensees may have surveillance requirements allowing the non-noble gas radionuclides (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with continuous release results.  In these instances, the table entries for the affected radionuclides for batch releases should include an appropriate designation (e.g., *) and an accompanying footnote describing this situation.
E. RADIOLOGICAL IMPACT ON MAN
Potential doses to individuals and populations should be calculated using measured effluent and meteorological data. A semiannual summary report should be submitted containing the following information:
1. Total body and significant organ doses to individuals in unrestricted areas from receiving water-related exposure pathways.


Table A-1D, Gaseous EffluentsElevated ReleaseContinuous Mode, contains a summation of gaseous effluent releases from elevated release points in the continuous mode of release for five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha.  Licensees should report the following:
1.
curies of each radionuclide released by quarter and year, and 
2.
2.


total curies released in each radionuclide category by quarter and year.
Total body and skin doses to individuals exposed at the point of maximum offsite ground-level concentrations of radioactive materials in gaseous effluents.


Table A-1E, Gaseous EffluentsMixed Mode ReleaseBatch Mode, contains a summation of gaseous effluent releases from mixed-mode release points in the continuous mode of release for five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha. Licensees should report the following:
3.


1.
Organ doses to individuals in unrestricted areas from radioactive iodine and radioactive material in particulate form from all pathways of exposure.


curies of each radionuclide released by quarter and year, and 
4.
2.


total curies released in each radionuclide category by quarter and year.
Total body doses to individuals and populations in unrestricted areas from direct radiation from the facility.


Some licensees may have surveillance requirements allowing the non-noble gas radionuclides (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with continuous release results.  In these instances, the table entries for the affected radionuclides for batch releases should include an appropriate designation (e.g., *) and an accompanying footnote describing this situation.
5.


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 five radionuclide categories:  fission and activation gases, iodines/halogens, particulates, tritium, and gross alpha. Licensees should report the following:
Total body doses to the population and average doses to individuals in the population from all receiving-water-related pathways.


1.
6.
 
curies of each radionuclide released by quarter and year, and 
2.


total curies released in each radionuclide category by quarter and year.
Total body doses to the population and average doses to individuals in the population from gaseous effluents to a distance of 50 miles from the site. If a significantly large population area is located just beyond
50 miles from the site, the dose to this population group should be considered.


Rev. 2 of RG 1.21, Page 36
F. METEOROLOGICAL DATA
8.2  Liquid Effluents
The report should include the cumulative joint frequency distribution of wind speed, wind direction, and atmospheric stability for the quarterly periods.


The quarterly and annual sums of all radionuclides released in liquid effluents (i.e., routine and abnormal discharges, continuous, and batch) should be reported in a format similar to that of the Tables A-
Similar data should be reported separately for the meteorological conditions during batch releases. (See Regulatory Guide 1.23 and Tables 4A and 4B in this appendix.)
2A and A-2B.  The data should then be further summarized and reported in the format of Appendix A,
1.21-12
Table A-2.  The following provides additional information on each of these tables.
 
Table A-2, Liquid Effluents - Summation of All Releases, contains a summation of all liquid radioactive discharges from all release points and all modes of release.  The data are subdivided by quarter and year for each of the radionuclide categories:  (a) fission and activation products, (b) tritium, (c) dissolved and entrained noble gases, and (d) gross alpha.  The total volume of primary coolant waste (typically batch mode releases) before dilution is also included.  In this context, primary coolant waste means the higher activity waste that generally is not discharged directly, but is instead typically processed through the liquid radioactive waste treatment system before discharge. Various methods exist for calculating the dilution water flow rate.  Health Physics Position HPPOS-099, Attention to Liquid Dilution Volumes in Semiannual Radioactive Effluent Release Reports, issued November 1984, indicates that licensees should use the total volume of dilution flow, not just that flow during periods of liquid effluent releases (Ref. 49). Licensees should include information describing how this value is calculated in either the ODCM or the ARERR.  Because the primary coolant waste typically accounts for the vast majority of the radioactive liquid waste discharges, it is recommended the volume and dilution data be summarized separately from the low-activity waste described in the following paragraph.
 
Report the total measured volume or average flow rate of waste from secondary or balance-of-plant systems (e.g., steam generator blowdown, low activity waste sumps, and auxiliary boilers).  In this context, secondary or balance-of-plant waste means the typically very low activity waste that is generally not processed with the liquid radioactive waste treatment system and that collectively represents a very large volume of waste.  Various methods exist for calculating the dilution water flow rate.  Health Physics Position HPPOS-099 indicates that licensees should use the total volume of dilution flow, not just that volume discharged during periods of liquid effluent releases.  Licensees should include information describing how this value is calculated in either the ODCM or the ARERR.  Because of the potentially high volume and extremely low activity of this type of waste, it is recommended the volume and dilution data be summarized separately from the higher activity waste described in the previous paragraph.
 
Licensees should report dilution flow rates during periods of release (before effluent is discharged to the receiving water body) as described above. If calculated differently than described above, the licensee should describe the method of calculation.  Licensees may choose to report near-field dilution if dilution by the receiving water body is taken into account.  Licensees may report the average, minimum, and/or peak river or stream flow rates if applicable.
 
Table A-2A, Liquid EffluentsBatch Mode, contains a summation of liquid effluent discharges in the batch mode of release.  The table is divided into four radionuclide categories:  fission and activation products, tritium, dissolved and entrained gases, and gross alpha.  Licensees should report the following:


EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT
Supplemental Information Facility licensee
1.
1.


curies of each radionuclide and gross alpha discharged by quarter and year, and
Regulatory Limits a.
2.


total curies in each radionuclide category by quarter and year.
Fission and activation gases:
b.


Rev. 2 of RG 1.21, Page 37 Table A-2B, Liquid EffluentsContinuous Mode, contains a summation of liquid effluent discharges in the continuous mode of release.  The table is divided into four radionuclide categories:
lodines:  
fission and activation products, tritium, dissolved and entrained gases, and gross alpha. Licensees should report the following: 
c.


1.
Particulates, half-lives >8 days:
d.


curies of each radionuclide and gross alpha discharged by quarter and year, and
Liquid effluents:
2.
2.


total curies in each radionuclide category by quarter and year.
Maximum Permissible Concentrations Provide the MPCs used in determining allowable release rates or concentrations.
 
8.3  Solid Waste Storage and Shipments
 
Appendix A, Table A-3, summarizes the solid radioactive waste (low-level waste) shipped from the site during the reporting period.  It is the intent that licensees report the volumes shipped and that licensees are not required to report the volumes that are buried.
 
The volume and curies shipped in each Waste Classification A, B, and C should be reported for each of the following waste streams:
 
1.
 
resins, filters, and evaporator bottoms,
2.
 
dry active waste,
3.
 
irradiated components,
4.
 
other waste, and
5.
 
sum of all waste.
 
Excluded from the reporting are those materials that are either being sent for laundry (either for washing or dissolving), metal melt, equipment for decontamination before disposal, and other very low- level waste such as material being surveyed for release in lieu of disposal.  However, records of these types of shipments should be maintained on site.
 
8.4  Dose Assessments
 
The annual evaluations of dose to members of the public should be calculated using the regulatory guidance in Regulatory Position 5 and should be reported in the format of Tables A-4 and A-5. Dose assessments should be performed to demonstrate compliance with the following: 
 
1.
 
Licensees should demonstrate compliance with 10 CFR Part 50, Appendix I (Table A-4),
by doing the following (note that the type of individual or dose receptor should be identified as a real individual or as a hypothetical individual if using bounding dose assessments; the individual/receptor is in the unrestricted area): 


a.
a.


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 for the maximum exposed individual.  If a particular exposure pathway is not applicable (i.e., it does not exist at a site), no dose should be calculated for that exposure pathway.
Fission and activation gases:
 
b.
b.


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.
lodines:
c.


Rev. 2 of RG 1.21, Page 38 c.
Particulates, half-lives >8 days:
d.


Report the organ dose from iodine, tritium, and particulates with a half-life greater than 8 days to the maximum exposed individual in an unrestricted area from all pathways of exposure (e.g., submersion and ingestion).  
Liquid effluents:
3.


2.
Average Energy Provide the average energy (E) of the radionuclide mixture in releases of fission and activation gases, if applicable.


Licensees should demonstrate compliance with 10 CFR 20.1301(e) and 40 CFR Part 190
4.
(Table A-5) by doing the following:
 
a.
 
Report the whole body, thyroid, and highest dose to any other organ from licensed and unlicensed radioactive material in the uranium fuel cycle, excluding background, to the individual member of the public likely to receive the highest dose.
 
8.5  Supplemental Information
 
Table A-6 in the appendix can be used to provide supplemental information in a descriptive, 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.
 
8.5.1  Abnormal Releases or Abnormal Discharges 
 
1.
 
Specific information should be reported concerning abnormal (airborne and/or liquid)
releases on site and abnormal discharges to the unrestricted area.  The report should describe each event in a way that would enable the NRC to adequately understand how the material was released and if there was a discharge to the unrestricted area.  The report should describe the potential impact on the ingestion exposure pathway involving surface water and ground water, as applicable.  The report should also describe the impact (if any)
on other affected exposure pathways (e.g., inhalation).
 
2.


The following are the thresholds for reporting abnormal releases and abnormal discharges in the supplemental information section:
Measurements and Approximations of Total Radioactivity Provide the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.


a.
a.


abnormal releases or abnormal discharges that are voluntarily reported to local authorities under NEI 07-07, Industry Ground Water Protection Initiative Final Guidance Document, (Ref.50); 
Fission and activation gases:
 
b.
b.


abnormal releases or abnormal discharges estimated to exceed 100 gallons (380
lodines:
liters) of radioactive liquid where the presence of licensed radioactive material is positively identified (in either the on-site environs or in the source of the leak or spill) as greater than the minimum detectable activity (the minimum detectable activity is a post-analysis calculation of sensitivity level based on the actual sample measurement) for the laboratory instrumentation; 
 
c.
c.


abnormal releases to on-site areas that result in detectable residual radioactivity after remediation; 
Particulates:
 
d.
d.


abnormal releases that result in a high effluent radiation alarm without an anticipated system trip occurring; and 
Liquid effluents:
 
5.
Rev. 2 of RG 1.21, Page 39 e.
 
abnormal discharges to an unrestricted area.
 
3.


Information on abnormal releases or abnormal discharges should include the following, as applicable: 
Batch Releases Provide the following information relating to batch releases of radioactive materials in liquid and gaseous effluents.


a.
a.


date and duration, b.
Liquid
 
1. Number of batch releases:
location, c.
 
volume, d.
 
estimated activity of each radionuclide, e.
 
effluent monitoring results (if any), 
f.
 
on-site monitoring results (if any), 
g.
 
depth to the local water table, h.
 
classification(s) of subsurface aquifer(s) (e.g., drinking water, unfit for drinking water, not used for drinking water), 
i.
 
size and extent of any ground water plume, j.
 
expected movement/mobility of any ground water plume, k.
 
land use characteristics (e.g., water used for irrigation), 
l.
 
remedial actions considered or taken and results obtained, m.
 
calculated member of the public dose attributable to the release n.
 
calculated member of the public dose attributable to the discharge, o.
 
actions taken to prevent recurrence, as applicable, and p.
 
whether the NRC was notified, the date(s), and the contact organization.
 
8.5.2  Non-routine Planned Discharges
 
Discharges resulting from remediation efforts that are not identified in the ODCM should be reported.  For example, the remediation effort may include pumping of contaminated ground water in response to leaks and spills.
 
8.5.3  Radioactive Waste Treatment System Changes 
 
Report any changes or modifications affecting any portion of the gaseous radioactive waste treatment system, the ventilation exhaust treatment system, or the liquid radioactive waste treatment.
 
8.5.4  Annual Land Use Census Changes 
 
Report any changes or modifications affecting significant aspects of the environmental monitoring program such as receptors, receptor locations, sample media availability, new (or changed) routes of exposure, etc.
 
8.5.5  Effluent Monitoring System Inoperability
 
1.
 
If an effluent radiation monitor is not operable for the consecutive time period listed in the licensees ODCM or technical specifications (typically 30 days), then the ARERR should include the radiation monitors equipment designation, the common name of the effluent radiation monitor, the time period of the inoperability, the reason why this inoperability was not corrected in a timely manner, and any other information required by the licensees ODCM or technical specifications.
 
Rev. 2 of RG 1.21, Page 40
2.
2.


In accordance with NUREG-1301 and NUREG-1302, Sections 3.3.3.10.b and 3.3.3.11.b the information above is required only when the minimum channels operability requirement is not achieved for the consecutive time period listed in the ODCM (typically
Total time period for batch releases:
30 days).  
3. Maximum time period for a batch release:
4.


8.5.6 Offsite Dose Calculation Manual Changes 
Average time period for batch releases:
5. Minimum time period for a batch release:
6.


Report any changes or modifications affecting significant aspects of the ODCM.
Average stream flow during periods of release of effluent into a flowing st ream:
 
b.
8.5.7  Process Control Program Changes
 
Report any changes or modifications affecting significant aspects of the ODCM.
 
8.5.8  Corrections to Previous Reports
 
1.
 
include a brief explanation of the error(s) 


Gaseous
1. Number of batch releases:
2.
2.


include a statement that the affected pages, in their entirety, are included as attachments to this ARERR 
Total time period for batch releases:
 
3.
3.


ensure a copy of the affected page(s), in their entirety, are included as attachments to this ARERR.  The attached pages should reference the affected calendar year and contain revision bars.
Maximum time period for a batch release:  
 
8.5.9  Other (Narrative Descriptions of Other Information Related to Radioactive Effluents) 
 
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regarding the NRCs plans for using this regulatory guide.  The NRC does not intend or approve any imposition or backfit in connection with its issuance.
 
Except in those cases in which an applicant or licensee proposes or has previously established an acceptable alternative method for complying with the specified portions of the NRCs regulations, the NRC staff will use the methods described in this guide in evaluating compliance with the applicable regulations.
 
Rev. 2 of RG 1.21, Page 41 GLOSSARY
 
a priori Before the fact limit representing the capability of a measurement system and not as an after the fact (a posteriori ) limit for a particular measurement.
 
abnormal dischargeThe unplanned or uncontrolled emission of an effluent (i.e., containing plant- related, licensed radioactive material) into the unrestricted area.
 
abnormal releaseThe unplanned or uncontrolled emission of an effluent (i.e., containing plant-related, licensed radioactive material).
 
accumulated radioactivityRadioactivity from prior-year effluent releases that may still be present in the media of concern.
 
ALARAAs Low as Reasonably Achievable 
 
ARERRAnnual Radioactive Effluent Release Report
 
AREORAnnual Radiological Environmental Operating Report
 
background (radiation)Means radiation from cosmic sources; naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material); and global fallout as it exists in the environment from the testing of nuclear explosive devices and from past nuclear accidents such as Chernobyl that contribute to background radiation and are not under the control of the licensee.  Background radiation does not include radiation from source, byproduct, or special nuclear materials regulated by the Commission.
 
batch releaseThe release of liquid (radioactive) wastes of a discrete volume or the release of a tank or purge of radioactive gases into the site environs.
 
channel checkThe qualitative assessment of channel behavior during operation by observation.  This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
 
channel operational testA channel operational test shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify operability of alarm, interlock and/or trip functions.  The channel operational test shall include adjustments, as necessary, of the alarm, interlock, and/or trip setpoints such that the setpoints are within the required range and accuracy.
 
continuous releaseAn essentially uninterrupted release of gaseous or liquid effluent for extended periods during normal operation of the facility where the volume of radioactive waste is non- discrete and there is input flow during the release.
 
controlled area (10 CFR 20)Means an area, outside of a restricted area but inside the site boundary, access to which is limited by the licensee for any reason.
 
Rev. 2 of RG 1.21, Page 42 controlled area (10 CFR 72)Means that area immediately surrounding an Independent Spent Fuel Storage Installation (ISFSI) or a Monitored Retrievable Storage facility (MRS) for which the licensee exercises authority over its use and within which ISFSI or MRS operations are performed.
 
controlled dischargeA radioactive discharge is considered to be controlled if (1) the discharge was conducted in accordance with methods, and without exceeding any of the limits, outlined in the ODCM, or (2) if one or more of the following three items are true:
 
1.  The radioactive discharge had an associated, pre-planned method of radioactivity monitoring that assured the discharge was properly accounted and was within the limits set by 10 CFR 20
and 10 CFR 50.
 
2.  The radioactive discharge had an associated, pre-planned method of termination (and associated termination criteria) that assured the discharge was properly accounted and was within the limits set by 10 CFR 20 and 10 CFR 50.
 
3.  The radioactive discharge had an associated, pre-planned method of adjusting, modulating, or altering the flow rate (or the rate of release of radioactive material) that assured the discharge was properly accounted and was within the limits set by 10 CFR 20 and 10 CFR 50.
 
controlled releaseA radioactive release is considered to be controlled if (1) the release was conducted in accordance with methods, and without exceeding any of the limits, outlined in the ODCM, or
(2) if one or more of the following three items are true:
 
1.  The radioactive release had an associated, pre-planned method of radioactivity monitoring that assured the release was properly accounted and was within the limits set by 10 CFR 20 and 10
CFR 50.
 
2.  The radioactive release has an associated, pre-planned method of termination (and associated termination criteria) that assured the release was properly accounted and was within the limits set by 10 CFR 20 and 10 CFR 50.
 
3.  The radioactive release had an associated, pre-planned method of adjusting, modulating, or altering the flow rate (or the rate of release of radioactive material) that assured the release was properly accounted and was within the limits set by10 CFR 20 and 10 CFR 50.
 
conversion factorA factor (e.g., microcuries per cubic centimeter per counts per minute (Ci/cc/cpm))
used to estimate a radioactivity concentration in an effluent based on a gross radioactivity measurement (e.g., counts per minute). 
 
D/QA 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 (Ci/m2)) of radioactive material at a location, divided by the release activity (Q) (e.g., in units of microcuries, Ci).  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.
 
Rev. 2 of RG 1.21, Page 43 determinationA 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). 
 
dilution water (for liquid radioactive waste)For purposes of this regulatory guide, any water, other than the undiluted radioactive waste, that is mixed with undiluted liquid radioactive waste before its ultimate discharge to the unrestricted area.
 
discharge pointA location at which radioactive material enters the unrestricted area.  This would be the point beyond the vertical plane of the unrestricted area (surface or subsurface). 
 
DQOData Quality Objectives
 
drinking waterWater that does not contain an objectionable pollutant, contamination, minerals, or infective agent and is considered satisfactory for domestic consumption.  This is sometimes called potable water.  Potable water is water that is safe and satisfactory for drinking and cooking.
 
Although EPA regulations only apply to public drinking water sources supplying 25 or more people (refer to EPA for more information), for purposes of the effluent and environmental monitoring programs, the term drinking water includes water from single-use residential drinking water wells.
 
effluentLiquid or gaseous waste containing plant-related, licensed radioactive material, emitted at the boundary of the facility (e.g., buildings, end-of-pipe, stack, or container) as described in the final safety analysis report (FSAR).
 
effluent dischargeThe portion of an effluent release that reaches an unrestricted area.
 
effluent releaseThe emission of an effluent.  (Same as radioactive release.)
 
elevated releaseA 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.
 
exposure pathwayA mechanism by which radioactive material is transferred from the (local)
environment to humans.  There are three commonly recognized exposure pathways; inhalation, ingestion, and direct radiation.  For example, ingestion is an exposure pathway, and it may include dose contributions from one or more routes of exposure.  For example, one route of exposure that may contribute to the ingestion exposure pathway is often referred to as grass-cow-milk-infant- thyroid route of exposure.
 
ground-level releaseA gaseous effluent release made from a height that is ator less thanthe height of adjacent solid structures, or where the degree of plume rise is unknown or is otherwise insufficient to avoid building wake effects.
 
Rev. 2 of RG 1.21, Page 44 ground waterAll water in the surface soil, the subsurface soil, or any other subsurface water.  Ground water is 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.
 
hypothetical exposure pathwayAn exposure pathway in which one or more of the components involved in the transfer of a radionuclide from the environment to the human does not actually exist at the specified location, or if a real human does not consume, inhale, or otherwise become exposed to the radioactive material.  For example, the grass-cow-milk-infant-thyroid route of exposure (associated with the ingestion exposure pathway) would be considered a hypothetical exposure pathway if the grass, the cow, or the milk did not actually exist at a specified location or if an infant did not actually consume the milk.
 
impacted areasMeans the areas with some reasonable potential for residual radioactivity in excess of natural background or fallout levels.  [Note: See 10 CFR 50.2, Definitions, and NUREG-1757 for a discussion of impacted areas.  For example, impacted areas include locations where radiological leaks or spills have occurred within the onsite environs (i.e., outside of the facilitys systems, structures, and components).  (See also the definition of significant contamination.)]
 
ISFSIIndependent Spent Fuel Storage Installation
 
leachateWater containing contaminants that is percolating downward from a pond or lake into the subsurface.
 
less-significant release pointAny location, from which radioactive material is released as a liquid or gaseous effluent, contributing less than or equal to 1 percent of the activity discharged from all the release points for a particular type of effluent considered.  Regulatory Guide 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous radioactive waste, and (3) all other nuclides discharged to the atmosphere in gaseous radioactive waste.
 
Example: If 1000 Ci of tritium are released in all liquid effluents in a given period of time (e.g., a typical calendar year or fuel cycle) and 0.01 Ci of tritium are released in steam generator blow down, then the steam generator blow down would be a less-significant release point.  Similarly, for gaseous releases of radionuclides other than noble gases (i.e., iodine, particulates, and tritium) if the total effluents are 10 Ci (iodine, particulates, and tritium) and the Refueling Water Storage Tank released 0.009 Ci of iodine, particulates, and tritium, then the Refueling Water Storage Tank would be a less-significant release point. In both of these examples the sample frequency can be adjusted to a frequency that is appropriate for that less significant release point.  Samples collected from these systems for other programs (e.g., detection of primary to secondary leakage) must still be collected and analyzed at the frequencies specified by the other programs.
 
licensed materialMeans 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.
 
lower limit of detection (LLD)The a priori smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability
 
Rev. 2 of RG 1.21, Page 45 with only a 5% probability of falsely concluding that a blank observation represents a real signal (see 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.51).
 
maximum individualIndividuals characterized as maximum with regard to food consumption, occupancy, and other usage of the region in the vicinity of the plant site.  As such, they represent individuals with habits that are considered to be maximum reasonable deviations from the average for the population in general.  Additionally, in physiological or metabolic respects, the maximum exposed individuals are assumed to have those characteristics that represent the averages for their corresponding age group in the general population.  (This term typically refers to members of the public).  See Regulatory Guide 1.109 for additional information.)
 
member of the public (10 CFR 20)Means any individual except when that individual is receiving an occupational dose.
 
member of the public (40 CFR 190)Means any individual that can receive a radiation dose in the general environment, whether he may or may not also be exposed to radiation in an occupation associated with a nuclear fuel cycle.  However, an individual is not considered a member of the public during any period in which the individual is engaged in carrying out any operation which is part of a nuclear fuel cycle.
 
minimum detectable concentrationThe 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 self- absorption) and is typically calculated following an actual sample analysis (a posteriori).  (See NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, issued June 1998 (Ref. 52)).
 
mixed mode releaseA gaseous effluent release made from a height higher than a ground-level release but less than an elevated release where, because of a lack of plume rise (e.g., buoyancy, momentum, and wind speed), a proper estimate of radionuclide transport and dispersion requires mathematically splitting the plume into (1) an elevated component and (2) a ground-level component to properly account for building wake effects.  (See Regulatory Guide 1.111 for further guidance.)
 
monitoringRadiation monitoring, radiation protection monitoring means the measurement of radiation levels, concentrations, surface area concentrations or quantities of radioactive material and the use of results of these measurements to evaluate potential exposures and doses.
 
nonroutine, planned dischargeAn 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 cycleThe operations defined to be associated with the production of electrical power for public use by any fuel cycle through the use of nuclear energy (see 40 CFR 190.02).
 
ODCMThe Offsite Dose Calculation Manual.
 
Rev. 2 of RG 1.21, Page 46
 
on-site environsLocation within the site boundary but outside of the systems, structures, or components described in the final safety analysis report or the ODCM.
 
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 radionuclideA principal radionuclide is one of the principal gamma emitters listed in NUREG-1301 and NUREG-1302, Tables 4.11-1 and Table 4.11-2, or alternatively, from a risk- informed perspective, a radionuclide is considered a principal radionuclide if it contributes either
(1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective dose when all radionuclides in the type of effluent are considered, or (2) greater than 1 percent of the activity of all nuclides in the type of effluent being considered.  Regulatory Guide 1.109 lists the three types of effluents as (1) liquid effluents, (2) noble gases discharged to the atmosphere, and (3) all other nuclides discharged to the atmosphere.  In this document, the terms principal radionuclide and principal nuclide are synonymous since this document is only concerned with measuring, evaluating, and reporting radioactive materials in effluents.
 
QAQuality Assurance
 
QCQuality Control
 
radioactive dischargeThe emission of an effluent (i.e., containing plant-related, licensed radioactive material) into the unrestricted area.  (Same as effluent discharge.)
 
radioactive releaseThe emission of an effluent (i.e., containing plant-related, licensed radioactive material).  (Same as effluent release.)
 
real exposure pathwayAn exposure pathway in which plant-related radionuclides in the environment at (or from) a specified location cause exposure to an actual individual.  For example, the grass-cow- milk-infant-thyroid exposure pathway would be considered a real exposure pathway if the grass, the cow, and the milk actually existed at a specified location and an infant actually consumed the milk.  For purposes of compliance with 10CFR50 Appendix I, the individual must be a member of the public.
 
release sourceA system, structure, or component (containing radioactive material under the licensees control) where radioactive materials are contained prior to release.
 
release pointA location from which radioactive materials are released from a system, structure, or component (including evaporative releases and leaching from ponds and lakes in the controlled or restricted area before release under 10 CFR 20.2001).  For release points monitored by plant process radiation monitoring systems, the release point is associated with the piping immediately downstream of the radiation monitor.  (See also the definition for significant release point.) 
Several release sources may contribute to a common release point.
 
Rev. 2 of RG 1.21, Page 47 residual radioactivityResidual radioactivity means radioactivity in structures, materials, soils, ground water, and other media at a site resulting from activities under the licensees control.  This includes radioactivity from all licensed and unlicensed sources used by the licensee, but it excludes background radiation.  It also includes radioactive materials remaining at the site as a result of routine or accidental releases of radioactive material at the site and previous burials at the site, even if those burials were made in accordance with the provisions of 10 CFR Part 20.
 
restricted areaRestricted area means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials.
 
Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area.
 
route of exposureA specific path (or delivery mechanism) by which radioactive material, originally in the environment at a specified location, can eventually cause a radiation dose to an individual.
 
The path typically includes a type of environmental medium (e.g., air, grass, meat, or water) as the starting point and a recipients organ or body as the end point.  For example, the grass-cow-milk- infant-thyroid route of exposure may contribute to the ingestion exposure pathway.  Additionally, several routes of exposure may contribute to a single exposure pathway.
 
scaling factorA factor used to estimate the unknown activity of a radionuclide based on its ratio to the activity of a readily measured radionuclide or other parameter (e.g., C-14 scaled to power generation).
 
significant contaminationAs 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 (see NUREG-1757).
 
significant release pointAny location, from which radioactive material is released, that contributes greater than 1 percent of the activity discharged from all the release points for a particular type of effluent considered.  Regulatory Guide 1.109 lists the three types of effluent as (1) liquid effluents,
(2) noble gases discharged to the atmosphere in gaseous radioactive waste, and (3) all other radionuclides discharged to the atmosphere in gaseous radioactive waste.
 
significant residual radioactivitySynonymous with the term significant contamination.
 
site boundarySite boundary means that line beyond which the land or property is not owned, leased, or otherwise controlled by the licensee.
 
site environsLocations outside of the nuclear power plant systems, structures, or components as described in the final safety analysis report or the ODCM.
 
source checkA source check is a qualitative assessment of the channel response when the channel sensor is exposed to a source of increased radioactivity.
 
surveySurvey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.  When appropriate, such an evaluation includes a physical survey of the location of
 
Rev. 2 of RG 1.21, Page 48 radioactive material and measurements or calculations of levels of radiation, or concentrations or quantities of radioactive material present.
 
TEDETotal Effective Dose Equivalent
 
type of effluentA grouping of radioactive releases into one of the three categories listed in 10 CFR 50
Appendix I, paragraphs A through C.  The three categories are classified in RG 1.109 as (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous radioactive waste, and (3) all other nuclides discharged to the atmosphere in gaseous radioactive waste.
 
unlicensed materialRadioactive material including (1) previously licensed material discharged in effluents, (2) background radioactivity, or (3) global fallout.  Licensed radioactive material becomes unlicensed radioactive material upon discharge in effluents in accordance with 10 CFR
20.2001.
 
uncontrolled dischargeAn effluent discharge that does not meet the definition of a controlled discharge.  See the definition of controlled discharge.
 
uncontrolled releaseAn effluent release that does not meet the definition of a controlled release.  See the definition of controlled release.
 
unplanned dischargeThe unintended or unexpected discharge of liquid or airborne radioactive material to the unrestricted area.  Examples of an unplanned discharge would include: 
 
1.
 
the unintentional discharge of a wrong waste gas decay tank (or bulk liquid radioactive waste tank), or 
 
2.
 
the failure of a radiation monitor to divert liquid to the radioactive waste system in the case where radioactivity is present and the automatic alarm/trip function fails to divert material to liquid radioactive waste and that material (or a portion of that material) is instead discharged to the environment.
 
unplanned releaseThe unintended or unexpected release of liquid or airborne radioactive material to the on-site environment.  An example of an unplanned release would include a plant occurrence that results in a leak or spill of radioactive material to on-site areas requiring a report under 10
CFR 50.72 or 10 CFR 50.73. (See NUREG/CR-5569, Health Physics Positions Data Base, February, 1994, HPPOS-254, Definition of Unplanned Release, (Ref. 53).) 
 
For example, if a licensee has prepared documents describing an intended release (e.g., a preliminary radioactive waste release permit) in advance of the evolution, and the intended release occurs as planned, then the release is a planned release.  If such documents (e.g., a preliminary release permit) are not prepared (or considered/evaluated) before the release, it is potentially an unplanned release (and additional information may be required to determine if it is an unplanned release).
 
unrestricted areaUnrestricted area means an area, access to which is neither limited nor controlled by the licensee.
 
Rev. 2 of RG 1.21, Page 49 uranium fuel cycleThe operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water- 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 byproduct materials from the cycle.
 
/QReferred to as Xi over Q, /Q is the average atmospheric effluent concentration, , normalized by release rate, Q, at a distance (or location) in a given downwind direction.  Expressed in another way, /Q is the concentration () of airborne radioactive material (e.g., in units of Ci/m3) divided by the release rate (Q) (e.g., in units of Ci/s) at a specified distance and direction downwind of the release point.
 
Rev. 2 of RG 1.21, Page 50
 
REFERENCES1
 
1 Publicly available NRC published documents such as Regulations, Regulatory Guides, NUREGs, and Generic Letters listed herein are available electronically through the Electronic Reading room on the NRCs public Web site at: http://www.nrc.gov/reading-rm/doc-collections/.  Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone 301-415-4737 or
(800) 397-4209; fax (301) 415-3548; and e-mail PDR.Resource@nrc.gov.
 
1.
 
Staff Requirements for SECY-98-144, White Paper on Risk Informed and Performance-Based Regulation, U.S. Nuclear Regulatory Commission, Washington, DC, March 1, 1999.  (ADAMS
ML003753593)
2.
 
10 CFR Part 20, Standards for Protection against Radiation, U.S. Nuclear Regulatory Commission, Washington, DC.
 
3.
 
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission, Washington, DC.
 
4.
4.


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.
Average time period for batch releases:
 
5.
5.


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,
Minimum time period for a batch release:
U.S. Nuclear Regulatory Commission, Washington, DC.
 
6.
6.


SECY-03-0069, Results of the License Termination Rule Analysis, U.S. Nuclear Regulatory Commission, Washington, DC, May 2, 2003.
Abnormal Releases a.


7.
Liquid
1. Number of releases:
2.


Regulatory Guide 4.1, Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.
Total activity released:
b.


8.
Gaseous I.


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.
Number of releases:
2.


9.
Total activity released:
1.21-13


NUREG-1301, Offsite Dose Calculation Manual Guidance:  Standard Radiological Effluent Controls for Pressurized Water Reactors, April 1991. (ADAMS Accession No. ML091050061)  
TABLE 1A
10.
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)  
GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES
Unit Quarter Quarter Es tal A. Fission & activation gases


NUREG-1302, Offsite Dose Calculation Manual Guidance:  Standard Radiological Effluent Controls for Boiling Water Reactors, April 1991. (ADAMS Accession No. ML091050059)
===1. Total release Ci E ===
11.
EE
 
E
Generic Letter 89-01, Implementation of Programmatic and Procedural Controls for Radiological Effluent Technical Specifications, U.S. Nuclear Regulatory Commission, Washington, DC,
2. Average release rate for period pCi/sec E
January 31, 1989.
E
 
3. Percent of Technical specification limit
Rev. 2 of RG 1.21, Page 51
%
12.
E
 
E
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.
B. lodines
 
1. Total iodine-131 Ci E
13.
E
 
.
NUREG-1757, Consolidated Decommissioning Guidance, September 2006.
E
 
2. Average release rate for period IpCi/sec E
14.
E
 
3. Percent of technical specification limit
NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols Manual, July 2004.
%
 
E  
15.
.
 
E
NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, October 2000.
C. Particulates
 
1. Particulates with half-lives >8 days Ci E
16.
E
 
E
NUREG/BR-0308, Effective Risk Communication, January 2004.
2. Average release rate for period pACi/sec E
 
E
17.
3. Percent of technical specification limit
 
%
NUREG/CR-6676, Probabilistic Dose Analysis Parameter Distributions Developed for RESRAD
E
and RESRAD-BUILD Codes, U.S. Nuclear Regulatory Commission, Washington, DC, July,
E
2000, (ADAMS Accession No. ML003741920).
4. Gross alpha radioactivity Ci E
18.
E
 
D. Tritium
NUREG/CR-6692, Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes, November, 2000, (ADAMS Accession No. ML003774030).
19.
 
NUREG/CR-6697,  Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0
Computer Codes, December 2000, (ADAMS Accession No. ML010090284).
20.
 
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.
 
21.
 
IAEA Technical Report Series Number 421, Management of Waste Containing Tritium and Carbon-14,  International Atomic Energy Agency, Vienna, 2004.2
22.
 
NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors PWR-GALE Code, Revision 1, April 1985
23.
 
NCRP Report No. 81, Carbon-14 in the Environment, National Council on Radiation Protection and Measurements, Bethesda, MD, January 1985.
 
24.
 
ASTM D 3370-07, Standard Practices for Sampling Water from Closed Conduits, American Society for Testing and Materials, West Conshohocken, PA, 2007.
 
25.
 
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.
 
26.
 
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.
 
2 Copies of the non-NRC documents included in these references may be obtained directly from the publishing organization.
 
Rev. 2 of RG 1.21, Page 52
27.
 
Regulatory Guide 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.
 
28.
 
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.
 
29.
 
NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, October 1978. (ADAMS Accession No. ML091050057)
30.
 
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.
 
31.
 
ANSI/ANS 2.17-2009, Evaluation of Subsurface Radionuclide Transport at Commercial Nuclear Power Production Facilities, American National Standards Institute, New York, NY (draft 2009).
32.
 
NUREG/CR-6948, Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities and Sites: Logic, Strategic Approach and Discussion, November 2007.
 
33.
 
EPRI Report No. 1011730, Ground Water Monitoring Guidance for Nuclear Power Plants, Electric Power Research Institute, Palo Alto, CA, September 2005.
 
34.
 
EPRI Report No. 1015118, Ground Water Protection Guidelines for Nuclear Power Plants, Electric Power Research Institute, Palo Alto, CA, November 2007.
 
35.
 
40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations, U.S. Environmental Protection Agency, Washington, DC.
 
36.
 
ICRP Publication 2, Report of Committee II on Permissible Dose for Internal Radiation, International Commission on Radiation Protection, Pergamon Press, Oxford, 1959
37.
 
Federal Register, 10 CFR 20, Final Rule, Standards for Protection Against Radiation, Volume
56, Number 98, page 23374, U.S. Nuclear Regulatory Commission, Washington, DC, May 21,
1991. (ADAMS Accession No. ML091050050)
38.
 
Federal Register, 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations, Volume 42, Number 9, page 2859, U.S. Nuclear Regulatory Commission, Washington, DC, January 13, 1977.
 
39.
 
NUREG-0543, Methods for Demonstrating Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR Part 190), February, 1980, (ADAMS Accession No. ML081360410)
40.
 
M. Maiello, The Variations in Long Term TLD Measurements of Environmental Background Radiation at Locations in Southeastern New York State and Southern New Jersey, Health Physics, Volume 72, Number 6, June 1997, pp. 915-922.
 
41.
 
ANSI N545-1975, Performance Testing and Procedural Specifications for Thermoluminescence Dosimetry (Environmental Applications), American National Standards Institute, 1975.
 
Rev. 2 of RG 1.21, Page 53
42.
 
ANSI/HPS N13.11-2009, American National Standard for Dosimetry Personnel Dosimetry Performance Criteria for Testing, American National Standard, January 13, 2009.
 
43.
 
Regulatory Guides 1.111, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors, U.S. Nuclear Regulatory Commission, Washington, DC, April, 1976.
 
44.
 
WASH-1258, Final Environmental Statement Concerning Proposed Rule Making Action: 
Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion As Low As Practical for Radioactive Material in Light-Water-Cooled Power Reactor Effluents, July, 1973.
 
45.
 
BNWL-1754, Models and Computer Codes for Evaluating Environmental Radiation Doses, February, 1974.
 
46.
 
ICRP Publication 60, ICRP Publication 60: 1990 Recommendations of the International Commission on Radiological Protection, 60, Annals of the ICRP Volume 21/1-3, International Commission on Radiation Protection, October, 1991.
 
47.
 
Federal Guidance Report Number 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion factors for Inhalation, Submersion, and Ingestion, Oak Ridge National Laboratory and Environmental Protection Agency, 1988.
 
48.
 
ASTM E-29, Practice for Using Significant Digits in Test Data to Determine Conformance with Specifications, American Society for Testing and Materials International, DOI: 10.1520/E0029-
08.
 
49.
 
NUREG/CR-5569, Health Physics Positions Data Base, HPPOS-099, Attention to Liquid Dilution Volumes in Semiannual Radioactive Effluent Release Reports, U.S. Nuclear Regulatory Commission, Washington, DC, November 1984.
 
50.
 
NEI 07-07, Industry Ground Water Protection InitiativeFinal Guidance Document, Nuclear Energy Institute, Washington, DC, August 2007. (ADAMS Accession No. ML072610036)
51.
 
NUREG/CR-4007, Lower Limit of Detection:  Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements, September 1984.
 
52.
 
NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, June 1998.
 
53.
 
NUREG/CR-5569, Health Physics Positions Data Base, HPPOS-254, Definition of Unplanned Release, U.S. Nuclear Regulatory Commission, Washington, DC, February, 1994.
 
54.
 
NUREG/CR-6805, A Comprehensive Strategy of Hydrogeology Modeling and Uncertainty Analysis For Nuclear Facilities and Sites, U.S. Nuclear Regulatory Commission, Washington, DC, July, 2003.
 
Rev. 2 of RG-1.21, Page 54
 
BIBLIOGRAPHY
 
U.S. Nuclear Regulatory Commission Documents 
 
NUREG-Series Reports NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, March 2007 (Section 2.3.5). 
 
NUREG-0324, XOQDOQ, Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, September 1977.
 
NUREG/CR-2919, XOQDOQ Computer Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, September, 1982 
 
Regulatory Guides Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Revision 2, November 2001.
 
U.S. Environmental Protection Agency Documents
40 CFR Part 141, National Primary Drinking Water Regulations, U.S. Environmental Protection Agency, Washington, DC.
 
National Standards 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 Gamma- Ray 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.
 
Appendix A to RG 1.21, Page A-1 APPENDIX A - TABLES


Table A-1. Gaseous EffluentsSummation of All Releases Summation of All Releases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Uncertainty Fission and Activation Gases Ci  
===1. Total release Ci E ===
E.


Average Release Rate Ci/s
EJI
2. Average release rate for period
3. Percent of technical specificati(
pCi/sec E
E
1.21-14 E
i
&deg;.
*


% of Limit
TABLE 1B
%
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)
GASEOUS EFFLUENTS-ELEVATED RELEASE
CONTINUOUS MODE
BATCH MODE
Nuclides Released Unit Quarter Quarter Quarter Quarter
1. Fission gases krypton-85 Ci E
E
E
E
krypton-85m Ci E
E
E
E
krypton-87 Ci E
E
E
E
krypton-88 Ci E
E
E
E
xenon-133 Ci E
E
E
E
xenon-135 Ci E
E
E
E
xenon-135m Ci E
E
E
E
xenon-138 Ci E
E
E
E
Others (specify)
Ci E
E
E
E
Ci E
E
E
E
Ci E
I
E
.
E
E
unidentified Ci E
E
E
E
Total for period Ci I
E
E
E
.
E
2. Iodines iodine-131 Ci E
E
E
E
iodine-133 Ci E
E
E
E
iodine-135 Ci E
E
E
E
Total for period Ci E
E
E
E
3. Particulates strontium-89 Ci E
E
.
E
E
strontium-90
Ci E
E
E
E
cesium-134 Ci E
E
E
E
cesium-137 Ci E
E
E
E
barium-lanthanum-140
Ci E
E
E
E
Others (specify)
Ci E
E
E
E
Ci_.


Iodines (Halogens)
E
Ci  
E
E
E
Ci E
E
E
E
unidentified Ci E
E
E
E
1.21-15


Average Release Rate Ci/s
TABLE 1C
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)
GASEOUS EFFLUENTS-GROUND-LEVEL RELEASES
CONTINUOUS MODE
Nuclides Released Unit Quarter Quarter Quarter Quater
1. Fission gases krypton-85 Ci E
E
E
E
krypton-85m Ci E
E
E
E
krypton-87 Ci E
E
E
E
krypton-88 Ci E
E
E
E
xenon-133 Ci E
E
E
E
xenon-135 Ci E
E
E
E
xenon-135m Ci E
E
E
E
xenon-138 Ci E
E
E
E
Others (specify)
Ci E
E
E
E
Ci E
E
E
E
Ci E
E
E
E
unidentified Ci E
E
EE
.
E
Total for period Ci E
E
E
E
2. Iodines iodine-131 Ci E
E
E
.
E
iodine. 133 Ci E
E
E
E
iodine-135 Ci E
E
E
E
Total for period Ci E
E
E
E
3. Particulates strontium-89 Ci E
E
E
E
strontium-90
Ci E
E
E
E
cesium-134 Ci E
E
E
E
cesium-137 Ci E
E
E
E
barium-lanthanum-140
Ci EE
E
E
E
Others (specify)
Ci E
E
E
E
Ci E
E
E
E
Ci E
E
E
E
unidentified Ci E
E
E
.
E
1.21-16 BATCH MODE


% of Limit
TABLE 2A
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)
LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES
Unit Quarter Quarter Est. Total Error, %  
A. Fission and activation products
1. Total release (not including tritium, gases, alpha)
Ci E
E
E
2. Average diluted concentration during period
/Ci/ml E
E
3. Percent of applicable limit
%  
%  
E
E E
B. Tritium


Particulates Ci  
===1. Total release Ci E ===
 
E
Average Release Rate Ci/s
2. Average diluted concentration during period
 
/Ci/ml E
% of Limit
E
3. Percent of applicable limit
%  
%  
.7 C. Dissolved and entrained gases


Tritium Ci  
===1. Total release Ci E ===
 
E
Average Release Rate Ci/s
E
 
2. Average diluted concentration during period Ici/mi E
% of Limit
E
3. Percent of appicable limit
%  
%  
E
E
D. Gross alpha radioactivity


Gross Alpha Ci  
===1. Total release Ci E ===
E
E. Volumne of waste released (prior to dilution)T liters I
E
.
E
.
E7 F. Volume of d ilution water used during period Iliters I
E
.
E
.
E
1.21-17


Appendix A to RG 1.21, Page A-2 Table A-1A.  Gaseous EffluentsGround-Level ReleaseBatch Mode Fission and Activation Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Ar-41 Ci  
TABLE 2B
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)
LIQUID EFFLUENTS
CONTINUOUS MODE
Nuclides Released
1*
I
Unit Quarter Quarter BATCH MODE
Quarter Quarter strontium-89 Ci E
E
E
E
strontium-90
Ci E
E
E
E
cesium-134 Ci E
E
E
E
cesium- 137 Ci E
E
E
E
iodine- 131 Ci E
E
.
E
E
cobalt-58 Ci E
E
E
E
cobalt-60
Ci E
EE
E
E
iron-59 Ci E
E
E
E
zinc-65
-
Ci EE
E
E
manganese-54 Ci EE
E
E
E
chromium-51 Ci E
E
E
E
zirconium-niobium-95 Ci E
E
E
E
molybdenum-99 Ci E
E
E
E
technetium-99m Ci E
E
E
E
barium-lanthanum-140
Ci E
E
E
E
cerium-141 Ci E
E
E
E
Other (specify)
Ci E
E
E
E
Ci E
E
E
E
Ci E
E
E
E
Ci E
E
E
E
Ci E
E
E
E
unidentified Ci E
E
E
E
Total for period (above)
Ci E
.
E I
E
E E
xenon-133 I
Ci E
E
El.


Kr-85 Ci  
E
xenon-135 Ci E
E
E
.
E
1.21-18


Kr-85m Ci
TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)
SOLID WASTE AND IRRADIATED FUEL SHIPMENTS
A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)


Kr-87 Ci
===1. Typ of waste Unit ===
6-month Est. Total Period Error, %
a. Spent resins, filter sludges, evaporator n?
E
bottoms, etc.


Kr-88 Ci  
Ci E
E
b. Dry compressible waste, contaminated in E
equip, etc.


Xe-131m Ci  
Ci E
E
C. Irradiated components, control m.


Xe-133 Ci
E
rods, etc.


Xe-133m Ci  
Ci E
E
d. Other (describe)
.
E
Ci E
E
2. Estimate of major nuclide composition (by type of waste)
b.


Xe-135 Ci
C.


Xe-135m Ci
d.
 
Xe-138 Ci
 
(List Others)
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 Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others)
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-3 Table A-1B.  Gaseous EffluentsGround-Level ReleaseContinuous Mode Fission and Activation 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
 
Xe-138 Ci
 
(List Others)
 
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 Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others)
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-4 Table A-1C.  Gaseous EffluentsElevated ReleaseBatch Mode Fission and Activation 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
 
Xe-138 Ci
 
(List Others)
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 Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others) 
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-5 Table A-1D.  Gaseous EffluentsElevated ReleaseContinuous Mode Fission and Activation 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
 
Xe-138 Ci
 
(List Others)
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 Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others)
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-6 Table A-1E.  Gaseous EffluentsMixed Mode ReleaseBatch Mode Fission and Activation 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
- E
 
__
Xe-135m Ci
__
 
__
Xe-138 Ci
__
 
__
(List Others)
__
Ci
__
 
__
Total Ci
__%_
 
E
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
__%_
 
E
Total Ci
 
Particulates Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others)
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-7 Table A-1F.  Gaseous EffluentsMixed Mode ReleaseContinuous Mode Fission and Activation 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
 
Xe-138 Ci
 
(List Others)
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 Co-58 Ci
 
Co-60
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Cs-134 Ci
 
(List Others)
Ci
 
Total Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-8 Table A-2.  Liquid EffluentsSummation of All Releases Summation of All Liquid Releases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Uncertainty
(%)
Fission and Activation Products (excluding tritium, gases, and gross alpha)
Ci
 
Average Concentration Ci/ml
 
% of Limit
%  
%  
 
E
Tritium Ci
 
Average Concentration Ci/ml
 
% of Limit
%  
%  
 
E
Dissolved and Entrained Gases Ci
 
Average Concentration Ci/ml
 
% of Limit
%  
%  
 
E
Gross Alpha Ci
__
 
__
Average Concentration Ci/ml
__
 
__
Volume of Primary System Liquid Effluent (Before Dilution)
__
Liters
__
 
__
Dilution Water Used for Above Liters
__
 
__
Volume of Secondary or Balance-of-Plant Liquid Effluent (e.g., low-activity or unprocessed)
_%_
(Before Dilution)
E
Liters
%_
 
E
Dilution Water Used for Above Liters
_
 
_
Average Stream Flow m3/s
_
 
_%  
Appendix A to RG 1.21, Page A-9 Table A-2A.  Liquid EffluentsBatch Mode Fission and Activation Products Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Cr-51 Ci
E~
 
__
Mn-54 Ci
__
 
__
Fe-55 Ci
__
 
__
Fe-59 Ci
_
 
__
Co-57 Ci
__
 
__
Co-58 Ci
_%_
 
E
Co-60
________________________E
Ci
 
Sr-89 Ci
 
Sr-90
Ci
 
Nb-95 Ci
 
Ag-110m Ci
 
Sn-113 Ci
 
Sb-124 Ci
 
Sb-125 Ci
 
I-131 Ci
 
I-133 Ci
 
I-135 Ci
 
Cs-134 Ci
 
Cs-137 Ci
 
(List Others)
Ci
 
Totals Ci
 
Appendix A to RG 1.21, Page A-10
Table A-2A.  Liquid EffluentsBatch Mode (continued)
Dissolved and Entrained Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Kr-85 Ci
 
Kr-85m Ci
 
Kr-88 Ci
 
Xe-131m Ci
 
Xe-133 Ci
 
Xe-133m Ci
 
Xe-135 Ci
 
Xe-135m Ci
 
(List Others) 
Ci
 
Totals Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-11 Table A-2B.  Liquid EffluentsContinuous 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
 
Nb-95 Ci
 
Ag-110m Ci
 
Sn-113 Ci
 
Sb-124 Ci
 
Sb-125 Ci
 
I-131 Ci
 
I-133 Ci
 
I-135 Ci
 
Cs-134 Ci
 
Cs-137 Ci
 
(List Others)
Ci
 
Totals Ci
 
Appendix A to RG 1.21, Page A-12 Table A-2B.  Liquid EffluentsContinuous Mode (continued)
Dissolved and Entrained Gases Units Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Kr-85 Ci
 
Kr-85m Ci
 
Kr-88 Ci
 
Xe-131m Ci
 
Xe-133 Ci
 
Xe-133m Ci
 
Xe-135 Ci
 
Xe-135m Ci
 
(List Others)
Ci
 
Totals Ci
 
Tritium Ci
 
Gross Alpha Ci
 
Appendix A to RG 1.21, Page A-13 Table A-3.  Low-Level Waste Resins, Filters, and Evaporator Bottoms Volume Curies Shipped Waste Class ft3 m3 Curies A 
 
B 
 
C 
 
ALL
 
Major Nuclides for the Above Table:
 
Dry Active Waste Volume Curies Shipped Waste Class ft3 m3
 
A 
 
B 
 
C 
 
ALL
 
Major Nuclides for the Above Table:
 
Appendix A to RG 1.21, Page A-14 Table A-3.  Low-Level Waste (continued)
Irradiated Components Volume Curies Shipped Waste Class ft3 m3
 
A 
 
B 
 
C 
 
ALL
 
Major Nuclides for the Above Table:
 
Other Waste Volume Curies Shipped WASTE
CLASS
ft3 m3
 
A 
 
B 
 
C 
 
ALL
 
Major Nuclides for the Above Table:
 
Appendix A to RG 1.21, Page A-15 Table A-3.  Low-Level Waste (continued)
Sum of All Low-Level Waste Shipped from Site Volume Curies Shipped Waste Class ft3 m3
 
A 
 
B 
 
C 
 
ALL
 
Major Nuclides for the Above Table:
 
Appendix A to RG 1.21, Page A-16
 
Table A-4.  Dose Assessments, 10 CFR Part 50, Appendix I
 
Quarter 1 Quarter 2 Quarter 3 Quarter 4 Yearly 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
5 mrad
5 mrad
5 mrad
5 mrad
10 mrad Gamma Air Dose
 
% of Limit
 
Gaseous Effluent Dose Limit, Beta Air
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 RG 1.21, Page A-17 Table A-5.  EPA 40 CFR Part 190 Individual in the Unrestricted Area
 
Whole Body Thyroid Any other organ Dose Limit
25 mrem
75 mrem
25 mrem Dose
 
% of Limit
 
Appendix A to RG 1.21, Page A-18 Table A-6.  Supplemental Information
 
1.
 
Abnormal Releases and Abnormal Discharges (e.g., leaks and spills)
 
2.
 
Non routine, 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.
3.


Radioactive Waste Treatment System Changes
Solid Waste Disposition Number of Shipments Mode of Transportation B. IRRADIATED FUEL SHIPMENTS (Disposition)
Number of Shipments Mode of Transportation
1.21-19 Destination Destination a.


4.
TABLE 4A
 
HOURS AT EACH WIND SPEED AND DIRECTION a PERIOD OF RECORD:
Annual Land-Use Census Changes
STABILITY CLASS:
ELEVATION:
Wind Speed (mph) at 1Om Level Wind Direction
1-3
4-7
8-12
13-18
19-24
>24 TOTAL
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
SW
WSW
W
WNW
NW
NNW
VARIABLE
Total Periods of calm (hours):
Hours of missing data:
a In the table, record the total number of hours of each category of wind direction for each calendar quarter. Provide similar tables separately for each atmospheric stability class and elevation.


5.
1.21-20


Effluent Monitor Instrument Inoperability
TABLE 4B
CLASSIFICATION OF ATMOSPHERIC STABILITY
Stability Pasquill oa0
Temperature change Classification Categories (degrees)
with height (VC/lOOm)
Extremely unstable A
25.0
<-1.9 Moderately unstable B
20.0
-1.9 to -1.7 Slightly unstable C
15.0
-1.7 to -1.5 Neutral D
10.0
-1.5 to -0.5 Slightly stable E
5.0
.0.5 to 1.5 Moderately stable F
2.5
1.5 to 4.0
Extremely stable G
1.7
>4.0
a Standard deviation of h6rizontal wind direction fluctuation over a period of 15 minutes to
1 hour. The values shown are average for each stability classification.


6.
*U.S.GOIVERNMENT
 
PRINTING OFFICE:,988-202-292:60349
Offsite Dose Calculation Manual Changes 
1.21-21
 
7.
 
Process Control Program Changes
 
8.
 
Errata/Corrections to Previous ARERRs
 
9.


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Measuring,Evaluating & Reporting Radioactivity in Solid Wastes & Releases of Radioactive Materials in Liquid & Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants
ML003739960
Person / Time
Issue date: 06/30/1974
From:
Office of Nuclear Regulatory Research
To:
References
RG-1.21, Rev 1
Download: ML003739960 (22)
v • d • e


U.S. ATOMIC ENERGY COMMISSION

REGULATORY

DIRECTORATE OF REGULATORY STANDARDS

Revision 1 June 1974 GUIDE

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

A. INTRODUCTION

General Design Criterion 60, "Control of releases of radioactive materials to the environment,"

of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10CFR Part 50, "Licensing of Production and Utilization Facilities," requires that the nuclear power plant design include means to control the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactcr operation, including anticipated operational occurrences.

General Design Criterion 64,

"Monitoring radioactivity releases,"

requires that nuclear powver plant designs provide means for monitoring effluent discharge paths for radioactivity that may be released

_-

from normal operations, including anticipated operational occurrences, and from postulated accidents.

Section 20.106, "Concentrations in effluents to unrestricted areas," of 10 CFR Part 20, "Standards for Protection Against Radiation," provides that a licensee shall not release to an unrestricted area, radioactive materials in concentrations which exceed limits specified in 10 CFR Part 20 or as otherwise authorized in a license issued by the Commission. Section 20.201,

"Surveys," of 10 CFR Part 20 further requires that a licensee conduct surveys of concentrations of radioactive materials as necessary to demonstrate compliance with AEC regulations.

Paragraph (a)(2)

of

§50.36a,

"Technical specifications on effluents from nuclear power reactors,"

of 10 CFR Part 50 provides that technical specifications for each license will include a requirement that the licensee submit a report to the Commission within 60

days after January 1 and July 1 of each year which specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents during the previous 6 months of operation, and such other information as may be required by the Commission to estimate maximum potential annual radiation doses to the public resulting from effluent releases.

Paragraph (c) of §20.1, "Purpose," of 10 CFR Part

20 states that every reasonable effort should be made by AEC licensees to maintain radiation exposure, and releases of radioactive materials in effluents to unrestricted areas, as far below the limits specified in Part 20 as practicable, i.e., as low as is practicably achievable, taking into account the state of technology, and the economics of improvements in relation to benefits to the public health and safety and in relation to the utilization of atomic energy in the public interest.

This guide describes programs acceptable to the Regulatory staff for measuring, reporting, and evaluating releases of radioactive materials in liquid and gaseous effluents and guidelines for classifying and reporting the categories and curie content of solid wastes. Other programs for the reporting of operating information, including abnormal occurrences, are presented in Regulatory Guide

1.16,

"Reporting of Operating Information." In some cases, specific programs should be supplemented because of individual plant design features or other factors. The need for supplemental or modified programs will be determined on a case-by-case basis.

The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.

USAEC REGULATORY GUIDES

Copies of published guld.

may be obtained by request indicating the divslions desired to the U.S. Atomic Energy Commission, Washington, D.C. 20545, Regulatory Guides we issued to describe and make available to the public Attention: Director of Regulatory Standarde. Comments end euggestions for methods acceptable to the AEC Regulatory staff of implementing specific parts of Improvements In these guides are encouraged and should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commission, U.S. Atomic Energy Commission, Washington, D.C. 20545, weluating specfic problems or postulated accidents, or to provide guidance to Attention: Chief, Public Proceedings Staff.

applicants. Regulatory Guides are not substitutes for regulations and compliance with them is not required. Methods and solutions different from those set out in The guides are issued in the following ten broad divisions:

the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission.

1. Power Reactors

6. Products

2. Research and Test Reactors

7. Transportation

3. Fuels and Materials Facilities

8. Occupational Health Published guides will be revised periodically, as appropriate, to accommodate

4. Environmental and Siting

9. Antitrust Review comments ard to reflect new information or experience.

5. Materials and Plant Protection

1

0. General

B. DISCUSSION

Information on the identity and quantity of radionuclides in liquid and gaseous effluents and solid wastes from light-water-cooled nuclear power plants, together with meteorological data representative of principal release points, are needed:

1.

For evaluation by the licensee and the Regulatory staff of the environmental impact of radioactive materials in effluents and solid wastes, including estimates of the potential annual radiation doses to the public;

2.

To ascertain whether AEC regulatory requirements and limiting conditions of operation have been met and whether concentrations of radioactive materials in liquid and gaseous effluents have been kept as low as practicable;

3.

For evaluation by the licensee and the Regulatory staff of the adequacy and performance of containment, waste treatment methods, and effluent controls.

It is essential to have a degree of uniformity in the methods used for measuring, evaluating, recording, and reporting data on radioactive material in effluents and solid wastes. The methods described in this guide provide a uniform basis for comparison of data from different sources and permit the preparation of consistent summaries of data for use by the Regulatory staff as bases for the assessment of a licensee's effluent controls and the potential environmental impact of radioactive materials in effluents and solid wastes.

This guide outlines general guidelines for monitoring and reporting programs. Detailed specifications for sampling and analysis of effluents are not included since they need to be tailored to the requirements of each specific plant. Standardized methods for monitoring, sampling, and analysis should be used to the extent practicable. The following is an example of a standard which is appropriate for these purposes.

The American National Standards Institute (ANSI)

has developed a standard' which includes general prin ciples and guidance for sampling airborne radioactive materials.

To assure uniformity of interpretation, the following definitions of terms used in this guide are provided:

Abnormal releases-unplanned or uncontrolled release of radioactive material from the site boundary.

SANSI N. 13.1-1969, "Guide to Sampling Airborne Radio active Materials in Nuclear Facilities." Copies may be obtained from the American National Standards Institute, Inc., 1430

Broadway, New York, N.Y. 10018.

Batch releases-discontinuous release of gaseous or liquid effluent which takes place over a finite period of time, usually hours or days.

Continuous release-release of gaseous or liquid effluent which is essentially uninterrupted for extended periods during normal operation of the facility.

Determined (or a 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. In some cases it may not be practical to make direct measurements of specific radionuclides in effluent or waste; e.g., the concentrations may be too low for measurement in a reasonable or practical volume of sample, certain nuclides may be masked by other radionuclides in the sample, or as in the case of solid or concentrated wastes, it may be difficult to obtain a representative sample. Under these circumstances, it may be more appropriate to calculate releases using previously estaibli-shed ratios with those nuclides which are readily measurable.

Such a procedure would constitute a determination.

Elevated release point-the point of release of gaseous waste for which credit was given as such in the determination of the technical specification limit for that release point.

Ground-level release point-the point of release for gaseous waste which is treated in the technical specifications as having zero height.

This guide, which is a revised and rewritten version of Regulatory Guide 1.21 (issued as Safety Guide 21 December 29, 1971), describes acceptable programs for measuring, evaluating, and reporting release of radioactive material in liquid and gaseous effluents and solid wastes from nuclear power plants. It also provides guidelines for calculating potential annual radiation doses to individuals and populations using appropriate models and parameters and pertinent recorded effluent and meteorological data. Significant changes from the previous version are identified below:

1. There has been a major change in the format of this guide. The more detailed recommendations concerning radionuclide measurements are presented in Appendix A

and the reporting recommendations are indicated in Appendix B.

2.

In many cases the criteria for sensitivity of effluent measurements have been modified to reflect as low as practicable dose considerations in the offsite environs;

i.e., the sensitivity of effluent measurements should be sufficient to detect concentrations which, when dispersed in the offsite environs, would result in a dose to individuals of a small fraction of natural background radiation.

3.

Some changes have been made in the frequency of analysis for certain radionuclides in several categories of effluents.

1.21-2

4.

Provisions for monitoring and reporting of solid wastes and for reporting of meteorological measurements, categories not considered in the earlier guide, have been included.

5.

Provisions for applying the measured meteorological and effluent data to acceptable dose models 2 in calculating potential doses to individuals and populations, and for reporting of these dose estimates have been included.

C. REGULATORY POSITION

1.

Meteorology A knowledge of meteorological conditions in the vicinity of the nuclear plant is essential to make valid estimates of maximum potential annual radiation doses resulting from radioactive materials released in gaseous effluents. Meteorological measurements should be made in accordance with the guidance set forth in Regulatory Guide 1.23 (Safety Guide 23), "Onsite Meteorological Programs." A summary report of the meteorological measurements taken during each calendar quarter in the

6-month period should be submitted with the semiannual Effluent and Waste Disposal Report as joint frequency distributions of wind direction and wind speed by atmospheric stability class in the format presented in Table 4A of Appendix B to this guide.

Hourly meteorological data for batch releases should be recorded for the periods of actual release, and quarterly summaries should be reported separately from the summaries of all observations taken during each quarter. The batch release data and the quarterly summaries of all observations should each be given in the format presented in Table 4A of Appendix B.

For abnormal releases, hourly meteorological data should be recorded for the periods of actual release and should be included in the quarterly summaries of batch releases.

2.

Location of Monitoring All major and potentially significant paths for release of radioactive material during normal reactor

2 Draft Regulatory Guide 1.AA, "Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Implementing Appendix I," Draft Regulatory Guide 1.DD, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents from Routine Releases," and Draft Regulatory Guide 1.EE, "Analytical Models for Estimating Radioisotope Concentration in Different Water Bodies", in Attachment to Concluding Statement, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactors, Docket RM-50-2, USAEC, February 20, 1974.

3 "Final Environmental Statement-Numerical Guides for Design Objectives and Limiting Conditions for Operation to

-'

Meet the Criterion 'As Low as Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," WASH-1258, Vol. 1, Directorate of Regulatory Standards, USAEC, July 1973.

operation, including anticipated operational occurrences, should be monitored. Measurements of effluent volume, rates of release, and specific radionuclides should be made, insofar as practicable, at the point(s) which would provide data that are the most representative of effluent releases to the plant environs. For those effluent discharge points which have input from two or more contributing sources within the plant, monitoring of the major contributing sources should also be considered from the standpoint of more effective process and effluent control. In many cases, monitoring of each of the major contributing sources may be a preferable or more sensitive alternative to monitoring the total effluent release when dilution with other less concentrated effluent streams makes the resultant effluent concentrations too low for accurate measurements.

3.

Type of Monitoring The type of monitoring selected, including the frequency, duration, and methods of measurement, depends to a large degree on the objectives of the monitoring program. Effluent monitoring is required to (a) demonstrate compliance with technical specification and/or 10 CFR Part 20 effluent limits, (b) allow evaluation of the performance of containment, waste treatment, and effluent controls, and (c)

permit evaluation of environmental impact and estimation of the potential annual radiation doses to the public.

Because radiation dose is dependent on the radionuclide(s) to which the individual is exposed, monitoring programs should provide accurate information on the identity and quantity of specific radionuclides in effluents and wastes.

4.

Gross Radioactivity Measurements Gross radioactivity measurements alone are generally not acceptable for showing compliance with effluent release limits. However, gross radioactivity measurements are often the only practicable means of continuously monitoring effluents and therefore are acceptable under certain specified conditions. Gross radioactivity measurements are acceptable for the purpose of quantifying radioactivity (a) when gross total radioactivity concentrations are a small fraction of the maximum permissible concentrations (MPCs)

for

"unidentified mixtures" as specified in the notes of Appendix B to 10 CFR Part 20 or (b) when gross radioactivity measurements are shown to be truly indicative of the actual quantity and/or concentration of radionuclides released.

5.

Measurements of Specific Radionucfides Measurements should be made to identify specific radionuclides in batch releases prior to their release to the environment. In those cases where analysis of specific radionuclides such as strontium-89 and strontium-90

cannot be made prior to release, representative samples should be collected from each

1.21-3

batch of effluents for the purpose of analysis at some later time. The use of composite samples is acceptable, and analyses of such samples should be performed at scheduled frequencies.

Measurements should be made to quantify specific radionuclides in continuous releases by analyses of grab samples collected at scheduled frequencies.

The frequency of radionuclide analyses should be based on the degree of variance of the concentrations and mixture compositions from an established norm. Continuous monitoring data as well as grab sample data should be the bases for identifying this variance.

SFrequent comparisons should be made between gross r'adioactivity measurements of continuous monitors and analyses of specific radionuclides. These comparisons should be the bases for calibrating cohitinuous monitors to establish relationships between monitor readings and concentrations or release rates of radionuclides in continuous effluent releases.

6.

Representative Samples A sample should be representative of the bulk stream or volume of effluent from which it is taken.

Provisions should be made to assure that representative samples are obtained from well-mixed streams or volumes of effluent by the selection of proper sampling equipment, the proper location of sampling points, and the development and use of vroper sampling procedures.

Prior to sampling, large volumes of liquid waste should be mixed in as short a time interval as practicable to assure that any sediments or particulate solids are distributed uniformly in the waste mixture. Sample points should be located where there is a minimum of disturbance of flow due to fittings and other physical characteristics of the equipment and components.

Sample nozzles should be inserted into the flow or liquid volume to ensure sampling the bulk volume of pipes and tanks. Sample lines should be flushed for a sufficient period of time prior to sample extraction in order to remove sediment deposits and air and gas pockets.

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 assure that individual samples are indeed representative of the effluent mixture.

The general principles for obtaining valid samples of airborne radioactive material, the methods and materials for gas and particle sampling, and the guides for sampling from ducts and stacks contained in ANSI

N13.1-1969 1 are generally acceptable and provide ade quate base, for the design and conduct of monitoring programs ior airborne effluents.

7.

Composite Samples To be representative of the average quantities and concentrations of radioactive materials released in liquid and in particulate form in gaseous effluents, samples for compositing 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. Prior to analysis, the composite should be thoroughly mixed S'

that the sample is representative of the average effluent release.

Periods of collection for composites should be as short as practicable to preclude the loss of radioactive material by deposition on walls of the sample container or volatilization of potentially volatile material. Periodic checks should be performed to identify any such changes in composite samples.

8.

Time between Collection and Analysis Measurements should be made as soon as practicable after collection to minimize loss of short-lived radionuclides by decay. Measurement of longer-lived radionuclides sometimes can be simplified by allowing sufficient time before their analysis for the decay of short-lived radionuclides.

Procedures should be instituted for handling, packaging, and storing samples to assure that loss of radioactive matenals or other factors causing sample deterioration do not invalidate the analysis.

9.

Corrections for Decay Decay corrections should be made as though the effluent were released uniformly throughout the sampling period unless it is shown that most of the effluent was released during a particularly short interval.

The exact time or time intervals of sample collection should be recorded. To estimate radioactive decay in composite or pooled samples, weighting should be applied to the delay time of each portion and to the quantity of each portion in relation to the total quantity of the sample.

10. Sensitivity The sensitivity limits given for radioactivity analyses in Appendix A of this guide are based on the potential significance in the environment of the quantities of radioactive materials released. For some radionuclides, lower detection limits than those given herein may be readily achievable and when measurements below the stated sensitivity limits are attained, the results should be recorded and reported.

For certain mixtures of gamma-emitting nuclides, it may not be possible to measure certain radionuclides at the stated sensitivity limits when other radionuclides are present in the sample in much greater concentrations.

Also, it may not be possible to measure certain radionuclides whose gamma ray yields are low (e.g.,

Kr-8 5, Cr-5 1, etc.) at the stated sensitivity limits. Under these circumstances, and in the case of radionuclides

1.21-4

which have no gamma rays and weak beta radiation (e.g.,

Fe-55, Ni-63, etc.), it may be more appropriate to calculate releases of such radionuclides using measured ratios of these radionuclides to those radionuclides which are routinely identified and measured.

Measurements should be made periodically to establish and assure the continued validity of the ratios used. Any reported data determined by this method should be clearly identified.

11. Accuracy of Measurements a.

Errors in Measurements An estimate should be made of the error associated with measurement of radioactive materials in effluents and solid wastes. Counting statistics can provide an estimate of the minimum error involved in radioactivity analyses.

Counting statistics (e.g.,

one-sigma counting error) should be included in the records of measurements, since they provide a readily calculable estimate of the statistical uncertainty due to counting.

The total or maximum error associated with the effluent measurement will include the cumulative errors resulting from the total operation of sampling and measurement. Because it may be very difficult to assign error terms for each parameter affecting the final measurement, detailed statistical evaluations of error are not suggested. The objective should be to obtain an overall estimate of the error associated with measurements of radioactive materials released in liquid and gaseous effluents and solid waste.

b.

Quality Controls Control checks and tests should be applied to the analytical process by the use of blind duplicate analyses of selected effluent samples and by cross-check analysis of selected samples with an independent laboratory. Quality controls should also be applied to the entire sample-collection procedure to assure that representative samples are obtained and that samples are not changed or affected prior to their analysis because of handling or because of their storage environment.

c.

Calibrations Individual written procedures should be prepared and utilized for specific methods of calibrating radiological monitoring systems and measuring equipment. Calibration practices for ancillary equipment and systems are described in Regulatory Guide 1.23,

"Onsite Meteorological Programs," and elsewhere, I and where appropriate, they should be utilized and included as a part of the 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.

Calibrations of measuring equipment should be performed using reference standards certified by the National Bureau of Standards or standards that have been calibrated against standards certified by the National Bureau of Standards. Calibration standards should have the necessiry accuracy, stability, and range required for their intended use.

Calibrations should generally be performed at regular intervals. Frequency of calibration should be based on the reproducibility and time stability of the system. 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 which gives measurements within a more narrow range. In many cases, it would be more appropriate to calibrate measuring equipment before and after use in addition to or instead of calibration at arbitrarily scheduled intervals. Calibration of measuring equipment before and after use permits detection of any erroneous readings or malfunctions that may have occurred during use.

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. Functional checks, i.e., routine checks performed to demonstrate that a given instrument is in working condition and functioning properly, may be performed using radioactive sources that are not standards.

Continuous radioactivity monitoring systems should be calibrated against appropriate standards and the relationship established between concentration and monitor readings 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 inservice calibrations should also be performed to relate monitor "readings"

to the concentrations and/or release rates of radioactive material in the monitored release path. These calibrations should be based on the results of analyses for specific radionuclides in grab samples from the release path.

12. Expression of Results of Measurements a.

Units The information and data on effluent releases included in reports to the Commission should be expressed in the units given in Appendix B of this guide and reported in the form given in paragraphs b and c below.

b.

Significant Figures To avoid ambiguity, significant figures should be used in recording the results of effluent

1.21-5

measurements. When several numbers are multiplied or divided together, the result should be rounded off to as few significant figures as are present in the factor with the fewest significant figures. When numbers are added or subtracted, the number with the fewest decimal places, not necessarily the fewest significant figures, puts the limit on the number of places that may justifiably be carried in the sum or difference.

For the purpose of reporting in the format of Appendix B of this guide, numerical values should be rounded off to three figures.

c.

Numerical Values Results of measurements, including percentages, should be reported in external floating point form, using the letter "E" to denote the exponent to the base 10. For example: 2% should appear as

2.00E+00; 0.00032 should appear as 3.20E-04; 157.6 should appear as 1.58E+02; 2.67 should appear as

2.67E+00.

The term "not detected" should not be used. If radioactivity in the sample(s) is less than the maximum sensitivity of measurement, the value should be reported as less than the maximum sensitivity. For example, if the maximum sensitivity is 3 x 10-9 uCi/ml, the values should be reported as <3.OOE-09.

13. Radiological Impact on Man Estimations of doses to individuals and populations are necessary for the assessment of the radiological impact on man from the operation of nuclear power plants. Dose calculations should be made using the measured effluent and meteorological data and acceptable dose models such as those provided in draft regulatory guides for implementation of numerical guides. 2 To the extent that they are not inconsistent with the models provided in these draft guides, other dose models such as those given in WASH-1258 3 or those used for calculating the estimated dose values given in the licensee's Environmental Report are also acceptable as bases for making dose calculations.

14. Other Provisions The provisions and principles presented in Appendices A and B of this guide are acceptable to the Regulatory staff as bases for measuring and reporting of radioactive materials in liquid and gaseous effluents and solid wastes from nuclear power plants, as well as for estimating doses to individuals and populations in the offsite environs.

1.21-6

APPENDIX A

MEASURING RADIOACTIVE MATERIALS IN LIQUID

AND GASEOUS EFFLUENTS AND SOLID WASTE

This appendix describes a monitoring program that is acceptable to the Regulatory staff. The frequencies of sampling and analysis and the types of measurements described are considered to be the minimum acceptable.

In some cases, this program should be supplemented with additional measurements because of individual plant design features or other factors. The need for supplemental or modified programs is determined on a case-by-case basis.

A.

GASEOUS EFFLUENTS

Continuous monitoring should be conducted along principal gaseous effluent discharge paths.

The radionuclide composition and quantities and concentrations of radioactive material in gaseous effluents should be determined and recorded. For the periods of release, the records should also show, on an hourly basis, the existing meteorological conditions of wind direction, wind speed, and atmospheric stability which are representative of conditions at the principal points of release (see Regulatory Guide 1.23, "Onsite Meteorological Programs").

The single Poisson (one sigma) error for discrete measurements should be less than 50 percent for release rates at the design objective level, less than 30 percent at twice the design objective release rate, and less than 20

percent at eight times the design objective release rate.

1.

Fission and Activation Gases During the release of gaseous wastes from the primary system waste gas holdup system, the effluent monitor should be operating and set to alarm and to initiate the automatic closure of the waste gas discharge valve before the limits specified in the technical specifications are exceeded.

a.

Continuous Releases For reactors which release gases continuously, a sample of the gaseous effluent should be analyzed within one month after the date of initial criticality of the reactor and at least weekly thereafter to determine the identity and quantity of the principal radionuclides being released. A similar analysis of samples should be performed following each refueling, process change, or other occurrence that could alter the mixture of radionuclides. For those processes or other conditions that change significantly (e.g., when the average daily gross radioactivity release rate equals or exceeds that given in the technical specifications or when the steady-state gross radioactivity release rate increases by

50% over the previous steady-state release rate at the same power level), an analysis should'be done following each change until it is shown that a pattern exists that can be used to predict the isotopic composition of the effluent. In addition, radionuclide analyses should be performed when continuous monitoring shows an unexplained variance from an established norm which may be indicative of a change in the concentration and composition. The norm should be established as a range of readings that may be expected due to normal operating conditions including anticipated operational occurrences.

The calibration of continuous gross radioactivity monitoring systems should be performed by normalizing against the results of specific radionuclide analyses using established ratios of the respective radionuclides to total activity.

When calibrated in this fashion, the gross radioactivity measurements obtained from continuous monitors may be used to determine the total quantity of radioactivity released.

b.

Batch Releases For reactors which release gases intermittently, an analysis should be made of a representative sample of each planned release prior to discharge to determine the identity and quantity of the principal radionuclides released.

Continuous monitoring should also be conducted at appropriate points to obtain information on the quantity and pattern of abnormal releases.

c.

Sensitivity For those discharge points which have input from two or more contributing sources within the plant, separate monitoring of the major sources should be performed as a more sensitive alternative to monitoring the composite effluent stream when bulk dilution results in concentrations too low for accurate measurements.

The sensitivity of gross radioactivity measurements of fission and activation gases, as a minimum, should be sufficient to permit measurement of a small fraction of the activity which would result in

(1) an annual air dose of 10 millirads due to gamma radiation at any location near ground level at or beyond the site boundary and (2) an annual air dose of 20

millirads due to beta radiation at any location near ground level at or beyond the site boundary.

The sensitivity of analysis for each of the principal radioactive gases in representative samples of gaseous effluents should be such that concentrations of

1 e ACi/cc are measurable.

1.21-7

2.

lodines a.

Monitoring A representative sample from the principal discharge paths should be drawn continuously through an iodine sampling device. The sample collected in the device should be analyzed at least weekly for iodine- 131.

An analysis should also be made monthly or more often for iodine-133 and iodine-135.

The results of these analyses should be used as the basis for recording, evaluating, and reporting the quantities of radioiodines released during the sampling period. In estimating releases for periods when analyses were not performed, the average of the two adjacent data points spanning this period should be used. These estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.

b.

Sensitivity The sensitivity of the analysis of radioiodines should be sufficient to permit measurement of a small fraction of the activity which would result in annual exposures of 15 millirems to the thyroid of individuals in unrestricted areas.

3.

Particulates a.

Monitoring A representative sample from the discharge paths should be drawn continuously through a

particulate filter. Measurements should be made on these filters to determine the quantities of radionuclides with half-lives greater than 8 days that are released in particulate form to the environment.

(1) The particulate filters should be changed and analyzed at least weekly for the principal gamma-emitting nuclides (at least for the radionuclides barium-lanthanum-140 and iodine-13 1). When quantities of released radioactive materials are at low levels, precluding accurate measurement of principal radionuclides, gross beta radioactivity measurements should be made as a basis for estimating the quantity Of radioactive material released in the week.

(2) A quarterly analysis for strontium-89 and strontium-90 should be made on a composite of all filters from each sampling location collected during the quarter.

(3)

A

monthly analysis for gross alpha radioactivity should be made on a composite of all filters collected during the month from each sampling location.

The results of these analyses should be used as the basis for recording and reporting the quantities of radioactive material in particulate form released during

,he sampling period. In estimating releases for periods when analyses were not performed, the average of the two adjacent data points spanning this period should be used. These estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.

b.

Sensitivity The sensitivity of analysis for radioactive material in particulate form should be sufficient to permit measurement of a small fraction of the activity which would result in annual exposures of 15 millirems to any organ of an individual in an unrestricted area.

4.

Tritium a.

Monitoring The release of tritium to the atmosphere should be determined for each batch released on an intermittent basis, and at least monthly for continuous releases.

b.

Sensitivity The sensitivity of analysis of tritium released to the atmosphere should be such that a concentration of

10-6 pCi/cc (of air) is measurable.

B.

LIQUID EFFLUENTS

During the release of radioactive wastes, the effluent control monitor should be set to alarm and to initiate automatic closure of the waste discharge valve prior to exceeding the limits specified in the technical specifications.

Continuous monitoring should be provided' for liquid effluent releases. The radionuclide mixture of liquid effluents should be determined and recorded. For the period(s) of release, the records should also show the volume of water used to dilute the liquid effluent and the resultant concentrations at the point(s) of release to unrestricted areas. If the effluent passes into a flowing stream, data on the average flow of the stream during periods of effluent release should be collected and reported in the Supplemental Information section of the report. (See Effluent and Waste Disposal Semiannual Report, Appendix B.)

The single Poisson (one sigma) error for discrete measurements should be less than 50 percent for release rates at the design objective level, less than 30 percent at twice the design objective release rate, and less than 20

percent at eight times the design objective release rate.

1.

Batch Releases a.

A representative sample of each batch of liquid effluent released should be analyzed for the principal gamma-emitting radionuclides.

1.21-8

When operational or other limitations preclude specific gamma radionuclide analysis of each batch, gross radioactivity measurements should be made to estimate the quantity and concentrations of radioactive material released in the batch, and a weekly sample composited from proportional aliquots from each batch released during the week should be analyzed for the principal gamma-emitting radionuclides.

b.

A

monthly sample composited from proportional aliquots from each batch released during the month should be analyzed for tritium and gross alpha radioactivity.

c.

A representative sample from at least one representative batch per month should be analyzed for dissolved and entrained fission and activation gases.

d.

A

quarterly sample composited from proportional aliquots from each batch released during the three-month period should be analyzed for strontium-89 and strontium-90.

The results of these analyses should be used as the basis for recording and reporting the quantities of radioactive material released in liquid effluents during the sampling period. In estimating releases for a period when analyses were not performed, the average of the two adjacent data points spanning this period should be used. Such estimates should be included in the effluent records and reports; however, they should be clearly identified as estimates, and the method used to obtain these data should be described.

2.

Continuous Releases For continuous releases (e.g., secondary plant leakage),

in addition to continuous monitoring, a representative sample of the liquid effluent should be analyzed at least weekly to determine the identity and quantity of the principal gamma-emitting radionuclides being released. Analysis for other specific radionuclides should be conducted in accordance with 1 above.

3.

Sensitivity The sensitivities of analyses of radioactive materials in liquid effluents should be sufficient to permit the measurement of concentrations of IffipCi/mri by gross radioactivity measurements, 5 x 10-7 yCi/ml of 1icTh gamma-emitting radionuclide, 10' uCi/nml of each of the dissolved and entrained gaseous radionuclides,

10-'

puCi/ml of gross alpha radioactivity, Ilff

/Ci/mli of tritium, and 5 x 10-8 pCi/ml of strontium-89 and strontium-90.

C. SOLID WASTE

The total curie quantity and radionuclide composition of the solid waste shipped offsite should be determined. Provisions should be made to monitor and to limit the curie quantity of material and the maximum radiation level of each package of solid waste in order to reduce radiation exposure to personnel and to meet the regulatory requirements of 10 CFR Part 71, "Packaging of Radioactive Material for Transport and Transportation of Radioactive Material under Certain Conditions," and of the Department of Transportation.

Monitoring of solid wastes in storage and preparatory to shipment should be performed to provide assurance that the radiation levels from waste in storage and in transport do not exceed regulatory limits.

1.21-9

APPENDIX B

EFFLUENT AND WASTE DISPOSAL REPORT

This appendix describes the data and information that should be included in effluent and waste disposal reports. The data and information should be reported in a format similar to that given in Tables 1 through 4 and the Supplemental Information sheet. Except as noted, effluent and solid waste data should be summarized on a quarterly basis, although in some cases more detailed data may be needed. The need fcr reporting of additional data to the Commission will be determined on a case-by-case basis.

The reporting method includes the use of uniform notation for numerical values and generally defined guidance for reporting certain supplemental information.

Data from licensee's effluent and waste disposal reports are compiled, and summary reports of nuclear power plant effluents are prepared by the Commission. The supplemental information reduces errors in processing and compiling of report data.

In the report, a separate section should contain a discussion of the radiological impact of facility operation on man. Calculations and estimates of potential doses to individuals and population doses should be summarized for the report (6-month) period, although in some cases more detailed data may be needed. The need for these additional data to be reported to the Commission is determined on a case-by-case basis.

Meteorological data during continuous releases should be submitted in the format presented in Table

4A. (Also see Regulatory Guide 1.23.) Data on meteoro logical conditions during batch releases should be reported separately in the same format. For the purpose of this guide, abnormal releases should be treated as batch releases, and the meteorological data obtained during abnormal releases should be included in the batch release report.

A.

SUPPLEMENTAL INFORMATION

1.

Regulatory Limits The technical specification limits for radioactive materials released in liquid and gaseous effluents should be included in each report. If changes are made in limiting conditions of operation during the report period, the appropriate limits and dates should be included.

2.

Maximum Permissible Concentrations The maximum permissible concentrations (MPC)

used to calculate permissible release rates and concentrations for air and water should be included in each report (if appropriate), i.e., the MPC used in accordance with technical specifications and/or derived from the use of Notes to Appendix B, 10 CFR Part 20.

3.

Average Energy The release rate limits for fission and activation gases in gaseous effluents are usually based on the average energy (FE) of the radionuclide mixture in the effluent. The E value for the gamma and beta energies per disintegration that is used should be included in the report.

4.

Measurements and Approximations of Total Radioactivity A summary description should be provided of the method(s)

used to determine or measure total radioactivity in effluent releases (total here means the overall gross curie quantity). For example, gross radioactivity measurements (gross beta and/or gross gamma) may be used to approximate total radioactivity

'n effluents, and/or analyses of specific radionuclides in selected or composited samples may be used to determine the radionuclide composition of the effluent.

A summary description of the methods used for estimating overall errors associated with radioactivity measurements should also be provided.

5.

Batch Releases The report should provide information relating to batch releases of liquid and gaseous effluents which are discharged to the environment. This information should include the number of releases, total time period for batch releases, and the maximum, mean, and minimum time period of release.

6.

Abnormal Releases The number of abnormal releases of radioactive material to the environment should be reported. The total curies of radioactive materials released as a result of abnormal releases should be included.

This information should be reported separately for liquid and gaseous releases. The activity values should also be included, as appropriate, in Tables 1 and 2.

Hourly meteorological data should be recorded for the periods of actual release and included in the quarterly summaries for batch releases in the format given in Table

4A.

B. GASEOUS EFFLUENTS

Summary information should be reported in the formats of Tables IA through IC. Table IA values

1.21-10

should include the sums of all sources of release, i.e.,

routine and abnormal releases, continuous and batch, elevated and ground level. The reported percent of technical specification limits should be based on the combined releases from multiple sources as given in the technical specifications. This also applies to the releases from multireactor sites.

For reactors that have technical specification limits for more than one principal point of release, separate radionuclide data should be reported for each of these release points. Data should be separated by release height, i.e., elevated or ground level, and these data should be further subdivided by release mode, i.e.,

continuous or batch mode. (See Tables lB and IC.)

Estimates of the total error associated with certain total values should be provided in each report. (See Table 1A.) These error values should be the best effort at an overall estimate of the errors associated with the totals in the report.

Report the following information as indicated by Tables IA through IC.

1. Gases a.

Quarterly sums of total curies of fission and activation gases released.

b.

Average release rates (guCi/sec) of fission and activation gases for the quarterly periods covered by the report.

c.

Percent of technical specification limit for releases of fission and activation gases. This should be calculated in accordance with technical specification limits.

d.

Quarterly sums of total curies for each of the radionuclides determined to be released, based on analyses of fission and activation gases. The data should be categorized by (1) elevated releases, batch and continuous modes, and (2) ground-level releases, batch and continuous modes. (See Tables lB and IC.)

2.

lodines a.

Quarterly sums of total curies of iodine-131 released.

b.

Average release rate (pCi/sec) of iodine-131.

c.

Percent of technical specification limit for iodine-131.

d.

Quarterly sums of total curies of each of the isotopes, iodine-131, iodine-133, and iodine-135 determined to be released. (See B.1.d above and Tables lB and 1C.)

3.

Particulates a.

Quarterly sums of total curies of radioactive material in particulate form with half-lives greater than 8 days determined to be released.

b.

Average release rate (pCi/sec) of radioactive material in particulate form with half-lives greater than 8 days.

c.

Percent of technical specification limit for radioactive material in particulate form with half-lives greater than 8 days.

d.

Quarterly sums of total curies for each of the radionuclides in particulate form determined to be released based on analyses performed. (See B.l.d above and Tables IB and IC.)

e.

Quarterly sums of total curies of gross alpha radioactivity determined to be released.

4.

Tritium a.

Quarterly sums of total curies of tritium determined to be released in gaseous effluents.

b.

Average release rate (pCi/sec) of tritium.

c.

Percent of appropriate technical specification or MPC limits for tritium.

C. LIQUID EFFLUENTS

Summary information should be reported in the formats of Tables 2A and 2B. Table 2A values should include the quarterly sums of all releases of radioactive materials in liquid effluents, i.e., routine and abnormal occurrences, continuous and batch. The reported percent of technical specification limits should be based on the combined releases from multiple sources as given in the technical specifications. This also applies to the releases from multireactor sites.

Estimates of the total error associated with certain total values should be provided in each report. (See Table 2A.) These error values should be the best effort at an overall estimate of the errors associated with the totals in the report.

Report the following information, as indicated by Tables 2A and 2B.

1.

Mixed Fission and Activation Products a.

Quarterly sums of total curies of radioactive material determined to be released in liquid effluents (not including tritium, dissolved and entrained gases, and alpha-emitting material). (See Table 2A.)

b.

Average concentrations (juCi/ml) of mixed fission and activation products (C.1 .a above) released to unrestricted areas, averaged over the quarterly periods covered by the report.

c.

Percent of applicable limit of average concentrations released to unrestricted areas (C.l.b above). Include the limit used and the bases in the supplemental report information.

d.

Quarterly sums of total curies for each of the radionuclides determined to be released in liquid effluents, based on analyses performed. Data should be separated by type of release mode, i.e., continuous or batch. (See Table 2B.)

1.21-11

2.

Tritium a.

Quarterly sums of total curies of tritium determined to be released in liquid effluents.

b.

Average concentrations (pCi/ml) of tritium released in liquid effluents to unrestricted areas, averaged over the quarterly periods covered by the report.

c.

Percent of applicable limit of average concentrations released to unrestricted areas (C.2.b above), i.e., percent of 3 x 10-3 pCi/mi. Include the limit and the bases in the supplemental report information.

3.

Dissolved and Entrained Gases a.

Quarterly sums of total curies of gaseous radioactive material determined to be released in liquid effluents.

b.

Average concentrations (pCi/ml) of dissolved and entrained gaseous radioactive material released to unrestricted areas, averaged over the quarterly periods covered by the report.

c.

Percent of technical specification limit of average concentrations released to unrestricted areas (C.3.b above). Include the limit used and the bases in the supplemental report information.

d.

Quarterly sums of total curies for each of the radionuclides determined to be released as dissolved and entrained gases in liquid effluents.

4.

Alpha Radioactivity Quarterly sums of total curies of gross alpha-emitting material determined to be released in liquid effluents.

5.

Volumes a.

Quarterly sums, in liters, of total measured volume, prior to dilution, of liquid effluent released.

b.

Quarterly sums of total determined volume, in liters, of dilution water used during the period of the report.

6.

Stream Flow Where the effluent passes into a flowing stream, data on the average flow of the stream during periods of effluent release should be collected and reported in the Supplemental Information section of the report.

D. SOLID WASTE

The following information should be reported for shipments of solid waste and irradiated fuel transported from the site during the report period:

1. The semiannual total quantity in cubic meters and the semiannual total radioactivity in curies for the categories or types of waste. (See Table 3.)

a.

Spent resins, filter sludges, evaporator bottoms;

b.

Dry compressible waste, contaminated equipment, etc.;

c.

Irradiated components, control rods, etc.;

d.

Other (furnish description).

2.

An estimate of the major nuclide composition in the categories of waste in D.1 above.

3.

The disposition of solid waste shipments. (Identify the number of shipments, the mode of transport, and the destination.)

4.

The disposition of irradiated fuel shipments.

(Identify the number of shipments, the mode of transport, and the destination.)

Estimates of the total error associated with certain total values should be provided in each report. (See Table 3.) These error values should be the best effort of an overall estimate of the errors associated with the totals in the report.

E. RADIOLOGICAL IMPACT ON MAN

Potential doses to individuals and populations should be calculated using measured effluent and meteorological data. A semiannual summary report should be submitted containing the following information:

1. Total body and significant organ doses to individuals in unrestricted areas from receiving water-related exposure pathways.

2.

Total body and skin doses to individuals exposed at the point of maximum offsite ground-level concentrations of radioactive materials in gaseous effluents.

3.

Organ doses to individuals in unrestricted areas from radioactive iodine and radioactive material in particulate form from all pathways of exposure.

4.

Total body doses to individuals and populations in unrestricted areas from direct radiation from the facility.

5.

Total body doses to the population and average doses to individuals in the population from all receiving-water-related pathways.

6.

Total body doses to the population and average doses to individuals in the population from gaseous effluents to a distance of 50 miles from the site. If a significantly large population area is located just beyond

50 miles from the site, the dose to this population group should be considered.

F. METEOROLOGICAL DATA

The report should include the cumulative joint frequency distribution of wind speed, wind direction, and atmospheric stability for the quarterly periods.

Similar data should be reported separately for the meteorological conditions during batch releases. (See Regulatory Guide 1.23 and Tables 4A and 4B in this appendix.)

1.21-12

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT

Supplemental Information Facility licensee

1.

Regulatory Limits a.

Fission and activation gases:

b.

lodines:

c.

Particulates, half-lives >8 days:

d.

Liquid effluents:

2.

Maximum Permissible Concentrations Provide the MPCs used in determining allowable release rates or concentrations.

a.

Fission and activation gases:

b.

lodines:

c.

Particulates, half-lives >8 days:

d.

Liquid effluents:

3.

Average Energy Provide the average energy (E) of the radionuclide mixture in releases of fission and activation gases, if applicable.

4.

Measurements and Approximations of Total Radioactivity Provide the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.

a.

Fission and activation gases:

b.

lodines:

c.

Particulates:

d.

Liquid effluents:

5.

Batch Releases Provide the following information relating to batch releases of radioactive materials in liquid and gaseous effluents.

a.

Liquid

1. Number of batch releases:

2.

Total time period for batch releases:

3. Maximum time period for a batch release:

4.

Average time period for batch releases:

5. Minimum time period for a batch release:

6.

Average stream flow during periods of release of effluent into a flowing st ream:

b.

Gaseous

1. Number of batch releases:

2.

Total time period for batch releases:

3.

Maximum time period for a batch release:

4.

Average time period for batch releases:

5.

Minimum time period for a batch release:

6.

Abnormal Releases a.

Liquid

1. Number of releases:

2.

Total activity released:

b.

Gaseous I.

Number of releases:

2.

Total activity released:

1.21-13

TABLE 1A

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES

Unit Quarter Quarter Es tal A. Fission & activation gases

1. Total release Ci E

EE

E

2. Average release rate for period pCi/sec E

E

3. Percent of Technical specification limit

%

E

E

B. lodines

1. Total iodine-131 Ci E

E

.

E

2. Average release rate for period IpCi/sec E

E

3. Percent of technical specification limit

%

E

.

E

C. Particulates

1. Particulates with half-lives >8 days Ci E

E

E

2. Average release rate for period pACi/sec E

E

3. Percent of technical specification limit

%

E

E

4. Gross alpha radioactivity Ci E

E

D. Tritium

1. Total release Ci E

E.

EJI

2. Average release rate for period

3. Percent of technical specificati(

pCi/sec E

E

1.21-14 E

i

°.

TABLE 1B

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

GASEOUS EFFLUENTS-ELEVATED RELEASE

CONTINUOUS MODE

BATCH MODE

Nuclides Released Unit Quarter Quarter Quarter Quarter

1. Fission gases krypton-85 Ci E

E

E

E

krypton-85m Ci E

E

E

E

krypton-87 Ci E

E

E

E

krypton-88 Ci E

E

E

E

xenon-133 Ci E

E

E

E

xenon-135 Ci E

E

E

E

xenon-135m Ci E

E

E

E

xenon-138 Ci E

E

E

E

Others (specify)

Ci E

E

E

E

Ci E

E

E

E

Ci E

I

E

.

E

E

unidentified Ci E

E

E

E

Total for period Ci I

E

E

E

.

E

2. Iodines iodine-131 Ci E

E

E

E

iodine-133 Ci E

E

E

E

iodine-135 Ci E

E

E

E

Total for period Ci E

E

E

E

3. Particulates strontium-89 Ci E

E

.

E

E

strontium-90

Ci E

E

E

E

cesium-134 Ci E

E

E

E

cesium-137 Ci E

E

E

E

barium-lanthanum-140

Ci E

E

E

E

Others (specify)

Ci E

E

E

E

Ci_.

E

E

E

E

Ci E

E

E

E

unidentified Ci E

E

E

E

1.21-15

TABLE 1C

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

GASEOUS EFFLUENTS-GROUND-LEVEL RELEASES

CONTINUOUS MODE

Nuclides Released Unit Quarter Quarter Quarter Quater

1. Fission gases krypton-85 Ci E

E

E

E

krypton-85m Ci E

E

E

E

krypton-87 Ci E

E

E

E

krypton-88 Ci E

E

E

E

xenon-133 Ci E

E

E

E

xenon-135 Ci E

E

E

E

xenon-135m Ci E

E

E

E

xenon-138 Ci E

E

E

E

Others (specify)

Ci E

E

E

E

Ci E

E

E

E

Ci E

E

E

E

unidentified Ci E

E

EE

.

E

Total for period Ci E

E

E

E

2. Iodines iodine-131 Ci E

E

E

.

E

iodine. 133 Ci E

E

E

E

iodine-135 Ci E

E

E

E

Total for period Ci E

E

E

E

3. Particulates strontium-89 Ci E

E

E

E

strontium-90

Ci E

E

E

E

cesium-134 Ci E

E

E

E

cesium-137 Ci E

E

E

E

barium-lanthanum-140

Ci EE

E

E

E

Others (specify)

Ci E

E

E

E

Ci E

E

E

E

Ci E

E

E

E

unidentified Ci E

E

E

.

E

1.21-16 BATCH MODE

TABLE 2A

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES

Unit Quarter Quarter Est. Total Error, %

A. Fission and activation products

1. Total release (not including tritium, gases, alpha)

Ci E

E

E

2. Average diluted concentration during period

/Ci/ml E

E

3. Percent of applicable limit

%

E

E E

B. Tritium

1. Total release Ci E

E

2. Average diluted concentration during period

/Ci/ml E

E

3. Percent of applicable limit

%

.7 C. Dissolved and entrained gases

1. Total release Ci E

E

E

2. Average diluted concentration during period Ici/mi E

E

3. Percent of appicable limit

%

E

E

D. Gross alpha radioactivity

1. Total release Ci E

E

E. Volumne of waste released (prior to dilution)T liters I

E

.

E

.

E7 F. Volume of d ilution water used during period Iliters I

E

.

E

.

E

1.21-17

TABLE 2B

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

LIQUID EFFLUENTS

CONTINUOUS MODE

Nuclides Released

1*

I

Unit Quarter Quarter BATCH MODE

Quarter Quarter strontium-89 Ci E

E

E

E

strontium-90

Ci E

E

E

E

cesium-134 Ci E

E

E

E

cesium- 137 Ci E

E

E

E

iodine- 131 Ci E

E

.

E

E

cobalt-58 Ci E

E

E

E

cobalt-60

Ci E

EE

E

E

iron-59 Ci E

E

E

E

zinc-65

-

Ci EE

E

E

manganese-54 Ci EE

E

E

E

chromium-51 Ci E

E

E

E

zirconium-niobium-95 Ci E

E

E

E

molybdenum-99 Ci E

E

E

E

technetium-99m Ci E

E

E

E

barium-lanthanum-140

Ci E

E

E

E

cerium-141 Ci E

E

E

E

Other (specify)

Ci E

E

E

E

Ci E

E

E

E

Ci E

E

E

E

Ci E

E

E

E

Ci E

E

E

E

unidentified Ci E

E

E

E

Total for period (above)

Ci E

.

E I

E

E E

xenon-133 I

Ci E

E

El.

E

xenon-135 Ci E

E

E

.

E

1.21-18

TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS

A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

1. Typ of waste Unit

6-month Est. Total Period Error, %

a. Spent resins, filter sludges, evaporator n?

E

bottoms, etc.

Ci E

E

b. Dry compressible waste, contaminated in E

equip, etc.

Ci E

E

C. Irradiated components, control m.

E

rods, etc.

Ci E

E

d. Other (describe)

.

E

Ci E

E

2. Estimate of major nuclide composition (by type of waste)

b.

C.

d.

- E

__

__

__

__

__

__

__

__

__%_

E

__

__

__

__

__

__

__

__

__%_

E

%

E

%

E

%

E

__

__

__

__

__

__

__

__

__

_%_

E

%_

E

_

_

_

_%

E~

__

__

__

__

__

_

__

__

__

_%_

E

________________________E

3.

Solid Waste Disposition Number of Shipments Mode of Transportation B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation

1.21-19 Destination Destination a.

TABLE 4A

HOURS AT EACH WIND SPEED AND DIRECTION a PERIOD OF RECORD:

STABILITY CLASS:

ELEVATION:

Wind Speed (mph) at 1Om Level Wind Direction

1-3

4-7

8-12

13-18

19-24

>24 TOTAL

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW

VARIABLE

Total Periods of calm (hours):

Hours of missing data:

a In the table, record the total number of hours of each category of wind direction for each calendar quarter. Provide similar tables separately for each atmospheric stability class and elevation.

1.21-20

TABLE 4B

CLASSIFICATION OF ATMOSPHERIC STABILITY

Stability Pasquill oa0

Temperature change Classification Categories (degrees)

with height (VC/lOOm)

Extremely unstable A

25.0

<-1.9 Moderately unstable B

20.0

-1.9 to -1.7 Slightly unstable C

15.0

-1.7 to -1.5 Neutral D

10.0

-1.5 to -0.5 Slightly stable E

5.0

.0.5 to 1.5 Moderately stable F

2.5

1.5 to 4.0

Extremely stable G

1.7

>4.0

a Standard deviation of h6rizontal wind direction fluctuation over a period of 15 minutes to

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The values shown are average for each stability classification.

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1.21-21

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