UN

TI ED STATES NUCLEAR REGULATORY COMM SI S

OI N

Occupational Radiation Exposure

at Commercial Nuclear Power

Reactors and other Facilities

1997

NUREG-0713

Vol. 19

Thirtieth Annual Report

U.S. Nuclear Regulatory Commission

 Office of Nuclear Regulatory Research

 M.L. Thomas, D.A. Hagemeyer
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Publicly released documents include, to name a

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NUREG-0713 ii

PREVIOUS REPORTS IN SERIES

WASH-1311 A Compilation of Occupational Radiation Exposure from Light Water Cooled Nuclear Power Plants, 1969-1973, U.S.

Atomic Energy Commission, May 1974.

NUREG-75/032 Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1974, U.S. Nuclear Regulatory

Commission, June 1975.

NUREG-0109 Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1975, U.S. Nuclear Regulatory

Commission, August 1976.

NUREG-0323 Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1976, U.S. Nuclear Regulatory

Commission, March 1978.

NUREG-0482 Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1977, U.S. Nuclear Regulatory

Commission, May 1979.

NUREG-0594 Occupational Radiation Exposure at Commercial Nuclear Power Reactors, 1978, U.S. Nuclear Regulatory

Commission, November 1979.

NUREG-0713 Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1979, Vol. 1, U.S. Nuclear Regulatory

Commission, March 1981.

NUREG-0713 Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1980, Vol. 2, U.S. Nuclear Regulatory

Commission, December 1981.

NUREG-0713 Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1981, Vol. 3, U.S. Nuclear Regulatory

Commission, November 1982.

NUREG-0713 Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1982, Vol. 4, U.S. Nuclear Regulatory

Commission, December 1983.

NUREG-0713 Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1983, Vol. 5, U.S. Nuclear Regulatory

Commission, March 1985.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1984, Vol. 6, U.S.

Nuclear Regulatory Commission, October 1986.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1985, Vol. 7, U.S.

Nuclear Regulatory Commission, April 1988.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1986, Vol. 8, U.S.

Nuclear Regulatory Commission, August 1989.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1987, Vol. 9, U.S.

Nuclear Regulatory Commission, November 1990.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1988, Vol. 10, U.S.

Nuclear Regulatory Commission, July 1991.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1989, Vol. 11, U.S.

Nuclear Regulatory Commission, April 1992.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1990, Vol. 12, U.S.

Nuclear Regulatory Commission, January 1993.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1991, Vol. 13, U.S.

Nuclear Regulatory Commission, July 1993.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1992, Vol. 14, U.S.

Nuclear Regulatory Commission, December 1993.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1993, Vol. 15, U.S.

Nuclear Regulatory Commission, January 1995.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1994, Vol. 16, U.S.

Nuclear Regulatory Commission, January 1996.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1995, Vol. 17, U.S.

Nuclear Regulatory Commission, January 1997.

NUREG-0713 Occupational Radiation Exposure At Commercial Nuclear Power Reactors and Other Facilities 1996, Vol. 18, U.S.

Nuclear Regulatory Commission, February 1998.

Previous reports in the NUREG-0714 series, which are now combined with NUREG-0713, are as follows:

WASH-1350-R1 First through Sixth Annual Reports of the Operation of the U.S. AEC’s Centralized Ionizing Radiation Exposure

 through Records and Reporting System, U.S. Atomic Energy Commission.

WASH-1350-R6

NUREG-75/108 Seventh Annual Occupational Radiation Exposure Report for Certain NRC Licensees - 1974, U.S. Nuclear

Regulatory Commission, October 1975.

NUREG-0119 Eighth Annual Occupational Radiation Exposure Report for 1975, U.S. Nuclear Regulatory Commission, October

1976.

NUREG-0322 Ninth Annual Occupational Radiation Exposure Report for 1976, U.S. Nuclear Regulatory Commission, October

1977.

NUREG-0463 Tenth Annual Occupational Radiation Exposure Report for 1977, U.S. Nuclear Regulatory Commission, October

1978.

NUREG-0593 Eleventh Annual Occupational Radiation Exposure Report for 1978, U.S. Nuclear Regulatory Commission, January

1981.

NUREG-0714 Twelfth Annual Occupational Radiation Exposure Report for 1979, Vol. 1, U.S. Nuclear Regulatory Commission,

August 1982.

NUREG-0714 Occupational Radiation Exposure, Thirteenth and Fourteenth Annual Reports, 1980 and 1981, Vols. 2 and 3, U.S.

Nuclear Regulatory Commission, October 1983.

NUREG-0714 Occupational Radiation Exposure, Fifteenth and Sixteenth Annual Reports, 1982 and 1983, Vols. 4 and 5, U.S.

Nuclear Regulatory Commission, October 1985.
iii NUREG-0713

ABSTRACT

This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear

Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System

(REIRS). The bulk of the information contained in the report was compiled from the 1997

annual reports submitted by six of the seven categories1 of NRC licensees subject to the

reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high

level waste currently licensed, only six categories will be considered in this report.

Annual reports for 1997 were received from a total of 296 NRC licensees, of which 109 were

operators of nuclear power reactors in commercial operation. Compilations of the reports

submitted by the 296 licensees indicated that 142,730 individuals were monitored, 75,291 of

whom received a measurable dose (Table 3.1). The collective dose incurred by these

individuals was 19,841 person-rem which represents a 9% decrease from the 1996 value. The

number of workers receiving a measurable dose also decreased, resulting in the average

measurable dose of 0.26 rem for 1997. The average measurable dose is defined to be the total

collective dose (TEDE) divided by the number of workers receiving a measurable dose. These

figures have been adjusted to account for transient reactor workers.

In 1997, the annual collective dose per reactor for light water reactor licensees (LWRs) was 157

person-rem. This represents a 9% decrease from the value reported for 1996. The annual

collective dose per reactor for boiling water reactors (BWRs) was 205 person-rem and, for

pressurized water reactors (PWRs), it was 132 person-rem.

Analyses of transient worker data indicate that 31,065 individuals completed work assignments

at two or more licensees during the monitoring year. The dose distributions are adjusted each

year to account for the duplicate reporting of transient workers by multiple licensees. In 1997,

the average measurable dose calculated from reported data was 0.22 rem. The corrected dose

distribution resulted in an average measurable dose of 0.26rem.

1 Commercial nuclear power reactors; industrial radiographers; fuel processors (including uranium enrichment), fabricators,

and reprocessors; manufacturers and distributors of byproduct material; independent spent fuel storage installations;

facilities for land disposal of low-level waste; and geologic repositories for high-level waste.
NUREG-0713 iv

EDITOR’S NOTE

The NRC currently has a five-year contract with Science Applications

International Corporation (SAIC) to assist the NRC Staff in the

preparation of the NUREG-0713 series. Mr. Charles Hinson (NRR)

assisted in the preparation of this NUREG, serving as the NRC

Technical reviewer. SAIC will be suggesting changes in the

presentation of certain data in these reports. Readers should be

alert to these changes, and the NRC welcomes responses, especially

where these changes can be improved upon.

Comments should be directed to:

Mary L. Thomas: (301) 415-6230

E-Mail Address: mlt1@nrc.gov

REIRS Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555
v NUREG-0713

TABLE OF CONTENTS

Page

PREVIOUS REPORTS IN SERIES .............................................................................................. ii

ABSTRACT .................................................................................................................................. iii

EDITOR’S NOTE ......................................................................................................................... iv

FOREWORD ................................................................................................................................xi

PREFACE .................................................................................................................................. xiii

1 INTRODUCTION ................................................................................................................. 1-1

1.1 Radiation Exposure Information on the Internet ...................................................... 1-3

2 LIMITATIONS OF THE DATA .............................................................................................. 2-1

3 ANNUAL PERSONNEL MONITORING REPORTS - 10 CFR 20.2206 .............................. 3-1

3.1 Definition of Terms and Sources of Data ................................................................. 3-1

3.1.1 Statistical Summary Reports ........................................................................... 3-1

3.1.2 Number of Monitored Workers......................................................................... 3-1

3.1.3 Number of Workers with Measurable Doses ................................................... 3-2

3.1.4 Collective Dose ................................................................................................ 3-2

3.1.5 Average Individual Dose .................................................................................. 3-3

3.1.6 Average Measurable Dose .............................................................................. 3-3

3.1.7 Number of Licensees Reporting ...................................................................... 3-3

3.1.8 CR ................................................................................................................... 3-3

3.2 Annual TEDE Dose Distributions............................................................................. 3-5

3.3 Summary of Occupational Exposure Data By License Category ............................ 3-8

3.3.1 Industrial Radiography Licenses, Single and

Multiple Locations ............................................................................................ 3-8

3.3.2 Manufacturing and Distribution Licenses, Type “A” Broad and

Limited ............................................................................................................. 3-9

3.3.3 Low-Level Waste Disposal Licenses ............................................................. 3-13

3.3.4 Independent Spent Fuel Storage Installation Licenses ................................. 3-13

3.3.5 Fuel Cycle Licenses ...................................................................................... 3-15

3.3.6 Light-Water-Cooled Power Reactor (LWR) Licenses .................................... 3-18

3.4 Summary of Intake Data by License Category ............................................................ 3-18
NUREG-0713 vi

TABLE OF CONTENTS (Continued)

Page

4 COMMERCIAL LIGHT WATER REACTORS - FURTHER ANALYSIS ............................... 4-1

4.1 Introduction.............................................................................................................. 4-1

4.2 Definition of Terms and Sources of Data ................................................................. 4-1

4.2.1 Number of Reactors ........................................................................................ 4-1

4.2.2 Electric Energy Generated .............................................................................. 4-5

4.2.3 Collective Dose per Megawatt-Year ................................................................ 4-5

4.2.4 Average Maximum Dependable Capacity ....................................................... 4-5

4.2.5 Percent of Maximum Dependable Capacity Achieved..................................... 4-6

4.3 Annual TEDE Distributions ...................................................................................... 4-6

4.4 Average Annual TEDE Doses ................................................................................. 4-8

4.5 Plant Rankings by Collective Dose per Reactor.................................................... 4-13

4.6 Collective Dose by Work Function and Employee Type ........................................ 4-18

4.7 Number of Personnel by Work Function and Employee Type ............................... 4-25

4.8 Graphical Representation of Dose Trends in Appendix E ..................................... 4-29

5 TRANSIENT WORKERS AT NRC LICENSED FACILITIES ............................................... 5-1

5.1 Termination Reports ................................................................................................ 5-1

5.2 Transient Workers at NRC Facilities........................................................................ 5-1

6 EXPOSURES TO PERSONNEL IN EXCESS OF REGULATORY LIMITS ......................... 6-1

6.1 Control Levels.......................................................................................................... 6-1

6.2 Limitations of the Data ............................................................................................. 6-2

6.3 Summary of Exposures in Excess of Regulatory Limits .......................................... 6-3

6.4 Maximum Exposures Below the NRC Limits ........................................................... 6-4

7 REFERENCES.................................................................................................................... 7-1

APPENDIX A - LISTING OF ANNUAL EXPOSURE DATA COMPILED FOR CERTAIN NRC

LICENSEES, 1997 ......................................................................................... A-1

APPENDIX B - ANNUAL WHOLE BODY DOSES AT LICENSED NUCLEAR POWER

FACILITIES, 1997............................................................................................ B-1

APPENDIX C - PERSONNEL, DOSE, AND POWER GENERATION SUMMARY,

1969-1997 .......................................................................................................C-1
vii NUREG-0713

TABLE OF CONTENTS (Continued)

Page

APPENDIX D - NUMBER OF PERSONNEL AND PERSON-REM BY WORK AND JOB

FUNCTION, 1997 ............................................................................................D-1

APPENDIX E - GRAPHICAL REPRESENTATION OF COLLECTIVE DOSE TRENDS

BY YEAR AND JOB FUNCTION FOR EACH SITE,

1973-1997 ....................................................................................................... E-1

APPENDIX F - SUMMARY OF ANNUAL WHOLE BODY DOSE DISTRIBUTIONS BY YEAR

AND REACTOR TYPE, 1988-1997 ................................................................. F-1

LIST OF TABLES

Table 3.1 Annual Exposure Data for Certain Categories of

Licensees 1988-1997 .................................................................................... 3-4

Table 3.2 Distribution of Annual Collective TEDE by License

Category 1997 ............................................................................................... 3-6

Table 3.3 Summary of Annual Dose Distributions for Certain NRC

Licensees 1968-1997 .................................................................................... 3-7

Table 3.4 Annual Exposure Information for Industrial

Radiographers 1995-1997 ............................................................................. 3-8

Table 3.5 Annual Exposure Information for Manufacturers and

Distributors 1995-1997 ................................................................................ 3-11

Table 3.6 Annual Exposure Information for Fuel Cycle Licenses 1995-1997.................. 3-15

Table 3.7 Intakes by Licensee Type and Radionuclide

Mode of Intake - Ingestion 1997 ................................................................. 3-19

Table 3.8 Intakes by Licensee Type and Radionuclide

Mode of Intake - Inhalation 1997 ................................................................ 3-20

Table 3.9 Collective and Average CEDE by Licensee 1997 ........................................... 3-23

Table 3.10 Internal Dose (CEDE) Distribution, 1994 - 1997.............................................. 3-24

Table 4.1 Summary of Information Reported by Commercial

Boiling Water Reactors 1973-1997 ................................................................ 4-2

Table 4.2 Summary of Information Reported by Commercial

Pressurized Water Reactors 1973-1997........................................................ 4-3
NUREG-0713 viii

LIST OF TABLES (Continued) Page

Table 4.3 Summary of Information Reported by Commercial

Light Water Reactors 1973-1997 ................................................................... 4-4

Table 4.4 Summary Distribution of Annual Whole Body Doses at

Commercial Light Water Reactors 1977-1997............................................... 4-7

Table 4.5 Boiling Water Reactors Listed in Ascending Order of

Collective Dose per Reactor 1993-1997...................................................... 4-14

Table 4.6 Pressurized Water Reactors Listed in Ascending Order

of Collective Dose per Reactor 1993-1997 .................................................. 4-15

Table 4.7a Five-year Totals and Averages Listed in Ascending

Order of Collective Dose per BWR 1993-1997............................................ 4-16

Table 4.7b Five-year Totals and Averages Listed in Ascending

Order of Collective Dose per PWR 1993-1997............................................ 4-17

Table 4.8 Annual Collective Dose by Work Function and Personnel Type 1997............. 4-20

Table 4.9 Percentages of Annual Collective Dose at LWRs by Work

Function 1986-1997 ..................................................................................... 4-21

Table 4.10 Annual Collective Dose by Occupation and Personnel Type 1997 ................. 4-24

Table 4.11 Number of Personnel by Work Function and Personnel Type 1997................ 4-26

Table 4.12 Number of Personnel by Occupation and Personnel Type 1997..................... 4-27

Table 4.13 Average Doses by Occupation and Personnel Type 1997 .............................. 4-28

Table 5.1 Effects of Transient Workers on Annual Statistical Compilations, 1997 ............ 5-2

Table 6.1 Occupational Exposures in Excess of Regulatory Limits, 1994-1997 ............... 6-5

Table 6.2 Occupational Exposures in Excess of Regulatory Limits, 1985-1993 ............... 6-6

Table 6.3 Maximum Occupational Exposures for Each Exposure Category 1997 ............ 6-7
ix NUREG-0713

LIST OF FIGURES Page

Figure 3.1 Average Annual Values at Industrial Radiography Facilities

1973-1997 ................................................................................................... 3-10

Figure 3.2 Average Annual Values at Manufacturing and Distribution

Facilities 1973-1997 .................................................................................... 3-12

Figure 3.3 Average Annual Values at Low-Level Waste Disposal

Facilities 1982-1997 .................................................................................... 3-14

Figure 3.4 Average Annual Values at Independent Spent Fuel Storage

Facilities 1982-1997 .................................................................................... 3-16

Figure 3.5 Average Annual Values at Fuel Cycle Licenses 1973-1997 ............................ 3-17

Figure 4.1 Average Collective Dose and Number of Workers per

Reactor 1973-1997 ........................................................................................ 4-9

Figure 4.2 Number of Operating Reactors and Gross Electricity

Generated 1973-1997 ................................................................................. 4-10

Figure 4.3 Average Measurable Dose per Worker and Collective Dose per

Megawatt-Year 1973-1997 .......................................................................... 4-11

Figure 4.4 Average, Median, and Extreme Values of the Collective

Dose per Reactor 1973-1997 ...................................................................... 4-12

Figure 4.5 Outage Days, Average Dose, and Collective Dose ......................................... 4-19

Figure 4.6 Collective Dose by Work Function and Personnel Type

1990-1997 ................................................................................................... 4-22
NUREG-0713 x

Intentionally left blank.
xi NUREG-0713

Foreword

NUREG-0713, Volume 19, summarizes the 1997 occupational radiation exposure data

maintained in the U.S. Nuclear Regulatory Commission’s Radiation Exposures Information

Reporting System (REIRS). Certain classes of licensees are required to annually report

individual exposures in accordance with 10 CFR 20.2206.

The occupational radiation exposure data contained in this volume of NUREG-0713 is a

compilation of the annual reports received from 300 licensees required to submit annual reports.

The annual collective dose decreased by 9% overall from 1996 to 1997. This decrease is

partially a result of a decrease in the total number of workers who received a measurable dose.

This decrease is shown in Table 3.1 of this NUREG.

John W. Craig, Director

Division of Regulatory Applications

Office of Nuclear Regulatory Research
NUREG-0713 xii

Intentionally left blank.
xiii NUREG-0713

PREFACE

A number of NRC licensees have inquired as to how the occupational radiation exposure data

that are compiled from the individual exposure reports required by § 20.2206 and the annual

dose data reported by work function in accordance with Subsection 6.9.1.5 of the standard

technical specifications for nuclear power plants are used by the NRC staff. This is a very

appropriate inquiry that may be of importance to many affected licensees. In combination with

other sources of information, the principal uses of the data are to provide facts regarding routine

occupational exposures to radiation and radioactive material that occur in connection with

certain NRC-licensed activities. These facts are used by the NRC staff as indicated below:

1. The data permit evaluation, from the viewpoint of trends, of the effectiveness of the overall

NRC/licensee radiation protection and ALARA efforts by certain licensees. They also

provide for the identification (and subsequent correction) of unfavorable trends.

2. The external dose data assist in the evaluation of the radiological risk associated with

certain categories of NRC-licensed activities and are used for comparative analyses of

radiation protection performance: US/foreign, BWRs/PWRs, civilian/military, facility/facility,

nuclear industry/other industries, etc.

3. The data provide for the monitoring of transient workers who may affect dose distribution

statistics through multiple counting.

4. The data help provide facts for evaluating the adequacy of the current risk limitation system

(e.g., are individual lifetime dose limits, worker population collective dose limits, and

requirements for optimization needed?).

5. The data permit comparisons of occupational radiation risks with potential public risks when

action for additional protection of the public involves worker exposures.

6. The data are used in the establishment of priorities for the utilization of NRC health physics

resources: research, standards development, and regulatory program development.

7. The data provide facts for answering Congressional and Administration inquiries and for

responding to questions raised by the public.

8. The data provide information that may be used in the planning of epidemiological studies.
1-1 NUREG-0713

Occupational Radiation Exposure

at Commercial Nuclear Power Reactors and Other Facilities

Thirtieth Annual Report, 1997

1 INTRODUCTION

One of the basic purposes of the Atomic Energy Act and the implementing regulations in Title

10, Code of Federal Regulations, Chapter I, Part 20, is to protect the health and safety of the

public, including the employees of the licensees conducting operations under those

regulations. Among the regulations designed to ensure that the standards for protection

against radiation set out in 10 CFR 20 are met is a requirement that licensees provide

individuals likely to be exposed to radiation with devices to monitor their exposure. Each

licensee is also required to maintain indefinitely records of the results of such monitoring.

However, there was no initial provision that these records or any summary of them be

transmitted to a central location where the data could be retrieved and analyzed.

On November 4, 1968, the U.S. Atomic Energy Commission (AEC) published an amendment

to 10 CFR 20 requiring the reporting of certain occupational radiation exposure information to

a central repository at AEC Headquarters. This information was required of the four

categories3 of AEC licensees that were considered to involve the greatest potential for

significant occupational doses and of AEC facilities and contractors exempt from licensing. A

procedure was established whereby the appropriate occupational exposure data were

extracted from these reports and entered into the Commission’s Radiation Exposure

Information Reporting System (REIRS), a computer system that was maintained at the Oak

Ridge National Laboratory Computer Technology Center in Oak Ridge, Tennessee, until May

1990. At that time, the data were transferred to a database management system at Science

Applications International Corporation (SAIC) at Oak Ridge, Tennessee. The computerization

of these data ensures that they are kept indefinitely and facilitates their retrieval and analysis.

The data maintained in REIRS have been summarized and published in a report every year

since 1969. Annual reports for each of the years 1969 through 1973 presented the data

reported by both AEC licensees and contractors and were published in six documents

designated as WASH-1350-R1 through WASH-1350-R6.

In January 1975, with the separation of the AEC into the Energy Research and Development

Administration (ERDA) and the U.S. Nuclear Regulatory Commission (NRC), each agency

assumed responsibility for collecting and maintaining occupational radiation exposure

information reported by the facilities under its jurisdiction. The annual reports published by the

3 Commercial nuclear power reactors; industrial radiographers; fuel processors (including uranium enrichment as of 1997),

fabricators, and reprocessors; manufacturers and distributors of specified quantities of byproduct material.
NUREG-0713 1-2

NRC on occupational exposure for calendar year 1974 and subsequent years do not contain

information pertaining to ERDA facilities or contractors. Comparable information for facilities

and contractors under ERDA, now the Department of Energy (DOE), is collected and

published by DOE’s Office of Health, a division of Environment, Safety and Health, in

Germantown, Maryland.

In 1982 and 1983, paragraph 20.408(a) of Title 10 of the Code of Federal Regulations was

amended to require three additional categories of NRC licensees to submit annual statistical

exposure reports and individual termination exposure reports. The new categories are (1)

geologic repositories for high-level radioactive waste, (2) independent spent fuel storage

installations, and (3) facilities for the land disposal of low-level radioactive waste. Therefore,

this document presents the exposure information that was reported by NRC licensees

representing two of these new categories. (There are no geologic repositories for high-level

waste currently licensed.)

This report and each of the predecessors summarizes information reported for both the current

year and for previous years. More licensee-specific data for previous years, such as the

annual reports submitted by each commercial power reactor pursuant to 10 CFR 20.407 and

their technical specifications, may be found in those documents listed on the inside of the front

cover of this report for the specific year desired. Additional operating data and statistics for

each power reactor for the years 1973 through 1982 may be found in a series of reports,

“Nuclear Power Plant Operating Experience” [Refs. 1-9]. These documents are available for

viewing at all NRC public document rooms, or they may be purchased from the National

Technical Information Service, as shown in the Reference section.

In May of 1991, the revised 10 CFR 20 “Standards for Protection Against Radiation; Final

Rule” was published in the Federal Register. The revision redefined the radiation monitoring

and reporting requirements of NRC licensees. Instead of summary annual reports (§ 20.407)

and termination reports (§ 20.408), licensees are now required to submit an annual report of

the dose received by each monitored worker (§ 20.2206). Licensees were required to

implement the new requirements on or before January of 1994. This report is the fourth

compilation of radiation exposure information collected under the revised 10 CFR 20. Certain

sections of the report have been modified to account for the change in the reporting of

exposure information. Readers are encouraged to comment on these changes.

Recommendations for further analysis or for different presentation of information are welcome.
1-3 NUREG-0713

1.1 Radiation Exposure Information on the Internet

In May of 1995, the NRC began pursuing the dissemination of radiation exposure information

via a World Wide Web site on the Internet. This allows interested parties with the appropriate

equipment to access the data electronically rather than through the published NUREG-0713

document. A web site was created for radiation exposure and linked into the main NRC web

page. The web site contains up-to-date information on radiation exposure, as well as

information and guidance on reporting radiation exposure information to the NRC. Interested

parties may read the documents on-line or down-load information to their systems for further

analysis. Software, such as REMIT, is also available for downloading via the web site. There

are also links to other web sites dealing with the topics of radiation and health physics. The

NRC intends to continue pursuing the dissemination of radiation exposure information via the

World Wide Web and will focus more resources on the electronic distribution of information

rather than the published hard copy reports.

The main web URL address for the NRC is:

http://www.nrc.gov

The NRC radiation exposure information web URL address is:

http://www.saic.com/home/nrc_rad

Comments on this report or the NRC’s web page should be directed to:

REIRS Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555
2-1 NUREG-0713

2 LIMITATIONS OF THE DATA

All of the figures compiled in this report relating to exposures and doses are based on the

results and interpretations of the readings of various types of personnel monitoring devices

employed by each licensee. This information, obtained from routine personnel monitoring

programs, is sufficient to characterize the radiation environment in which individuals work and

is used in evaluating the radiation protection program.

Monitoring requirements are specified in 10 CFR § 20.1502, which requires licensees to

monitor individuals who receive or are likely to receive a dose in a year in excess of 10% of the

applicable limits. For most adults, the annual limit for the whole body is 5 rem, so 0.5 rem per

year is the level above which monitoring is required. Separate dose limits have been

established for minors and pregnant workers. Monitoring is required for any individual entering

a high or very high radiation area. Depending on the administrative policy of each licensee,

persons such as visitors and clerical workers may also be provided with monitoring devices,

although the probability of their being exposed to measurable levels of radiation is extremely

small. Licensees must report the dose records of those individuals for whom monitoring is

required. Many licensees elect to report the doses for every individual for whom they provided

monitoring. This practice tends to increase the number of individuals that one could consider

to be radiation workers. In an effort to account for this, the number of individuals reported as

having “no measurable exposure” has been subtracted from the total number of individuals

monitored in order to calculate an average dose per individual receiving a measurable dose,

as well as the average dose per monitored individual (for example, see Table 3.1).

The Revised 10 CFR § 20 was published in the Federal Register on May 21, 1991. With the

revision of Part 20, licensees report the monitoring results for each individual. This has

eliminated the need for the staff to calculate collective dose from the statistical distributions

and has improved the accuracy of the collective dose information presented in this report.

Although licensees were required to implement the new reporting requirements as of

January 1, 1994, certain licensees began reporting under these new requirements during

1993, and that data has been included in the analyses presented here.

Another impact of the Revised Part 20 is the change from whole body dose to total effective

dose equivalent (TEDE). The TEDE includes both external and internal dose. The TEDE is

determined by summing the deep dose equivalent (DDE) from external radiation exposure and

the committed effective dose equivalent (CEDE) from internal exposures. For reports prior to

1994, only the whole body dose (equivalent to the DDE) was reported and analyzed. In

subsequent reports, the TEDE is presented and analyzed in all graphs and tables unless

otherwise noted. Readers should be aware of this change from external whole body dose to
NUREG-0713 2-2

the TEDE. For most licensed activities, the internal dose is not a significant contributor to the

TEDE. However, workers at Fuel Fabrication facilities receive significant exposures from

internal exposure. This change in reporting requirements can be seen in the 1994 through

1997 data for this licensee category. (See Section 3.3.5)

The average dose per individual, as well as the dose distributions shown for groups of

licensees, also can be affected by the multiple reporting of individuals who were monitored by

two or more licensees during the year. Licensees are only required to report the doses

received by individuals at their licensed facility. A dose distribution for a single licensee does

not consider that some of the individuals may have received doses at other facilities. When

the data are summed to determine the total number of individuals monitored by a group of

licensees, individuals may be counted more than once. This can also affect the distribution of

doses because individuals may be counted multiple times in the lower dose ranges rather than

one time in the higher range corresponding to the actual accumulated dose for the year (the

sum of the individual’s dose accrued at all facilities). This source of error has the greatest

potential impact on the data reported by power reactor facilities since they employ many

short-term workers. Further discussion of this point is provided in Section 5.

Another fact that should be kept in mind when examining the annual statistical data is that all

of the personnel included in the report may not have been monitored throughout the entire

year. Many licensees, such as radiography firms and nuclear power facilities, may monitor

numerous individuals for periods much less than a year. The average doses calculated from

these data, therefore, are less than the average dose that an individual would receive if

involved in that activity for the full year.

Considerable attention should also be given when referencing the collective totals presented in

this report. The differences between the totals presented for all licensees that reported versus

only those licensees that are required to report should be noted. Likewise, one should

distinguish between the doses attributed to the pressurized water reactors (PWRs), and boiling

water reactors (BWRs). The totals may be inclusive or exclusive of those licensees that were in

commercial operation for less than one full year. These parameters vary throughout the tables

and appendices of this report in order to provide the most comprehensive analysis of all the

data available. The apparent discrepancies among the various tables are a necessary

side-effect of this endeavor.

Also, it should again be pointed out that this report contains information reported by NRC

licensees only. Since the NRC licenses all commercial nuclear power reactors, fuel processors

and fabricators, and independent spent fuel storage facilities, information shown for these

categories reflects the U.S. experience. This is not the case, however, for the remaining

categories of industrial radiography, manufacturing and distribution of specified quantities of
2-3 NUREG-0713

by-product material, and low-level waste disposal. Companies that conduct these types of

activities in Agreement States4 are licensed by the state and are not required to submit

occupational exposure reports to the NRC. Approximately twice as many facilities are licensed

to Agreement States than the number licensed by the NRC. This report also does not include

non-occupational exposure such as exposure due to medical x-rays, fluoroscopy, and

accelerators received as a patient. Information shown for these categories does not reflect the

total U.S. experience.

All dose equivalent values in this report are given in units of rem in accordance with the

general provisions for records, 10 CFR 20.2101(a). In order to convert rem into the SI unit of

Sieverts (Sv), one should divide the value in rem by 100. Therefore 1 rem = 0.01 Sv. In order

to convert rem into millisieverts (mSv), multiply the value in rem by 10. Therefore 1 rem = 10

mSv.

4 States that have entered into an agreement with the NRC that allows each state to license organizations using radioactive

materials for certain purposes. As of 12/31/97, there are 30 Agreement States.
3-1 NUREG-0713

3 ANNUAL PERSONNEL MONITORING REPORTS - 10 CFR 20.2206

3.1 Definition of Terms and Sources of Data

3.1.1 Statistical Summary Reports

On February 4, 1974, 10 CFR 20.407 was amended to require certain categories4 of licensees

to submit an annual statistical report indicating the distribution of the whole body doses

incurred by workers whom they monitored for exposure to radiation. Since the regulations did

not require these licensees to report the collective dose incurred by the workers shown on the

statistical reports, the dose distributions were used as the basis for the staff’s calculation of the

collective dose (see Section 3.1.4).

The revised 10 CFR 20 was published in the Federal Register on May 21, 1991. Section

20.2206 of the revised rule requires licensees to report the radiation exposure monitoring

results for each individual for the monitoring year. All licensees were required to implement the

new reporting requirements on or before January 1, 1994.

Under the revised requirements, the individual’s total effective dose equivalent (TEDE, as

defined in §20.1003) is reported, so that the dose distributions may be determined directly

from the individual’s exposure. The TEDE is summed per individual and tabulated into the

appropriate dose range to generate the dose distribution for each licensee. The total collective

dose is more accurate using this method, since the licensee reported the dose to each

individual and the total collective dose was calculated from the sum of these doses and not

statistically derived from the distribution (see Section 3.1.4). The TEDE includes the dose

contribution from the committed effective dose equivalent (CEDE) for those workers who had

intakes that required monitoring and reporting of internal dose. Reports submitted under

formerly applicable 10 CFR 20.407 did not include the whole body contribution from internal

dose.

3.1.2 Number of Monitored Workers

The number of monitored workers refers to the total number of workers that the NRC

licensees, who are covered by 10CFR 20.1502, reported as being monitored for exposure to

external and internal radiation during the year. This number includes all workers for whom

monitoring is required, and may include visitors, service representatives, contract workers,

clerical workers, and any other workers for whom the licensee feels that monitoring devices

should be provided.

4 Commercial nuclear power reactors; industrial radiographers; fuel processors (inclding uranium enrichment as of 1997),

fabricators and reprocessors; and manufacturers and distributors of by-product material. Independent spent fuel storage

installations; and facilities for land disposal of low-level radioactive waste were added to this list in 1983.
NUREG-0713 3-2

For licensees submitting under the revised 10 CFR 20.2206, the total number of workers was

determined from the number of unique personal identification numbers submitted per licensee.

Uniqueness is defined by the combination of identification number and identification type.

[Ref. 18]

3.1.3 Number of Workers with Measurable Dose

Under the revised 10 CFR 20.2206, the number of workers with measurable dose includes any

individual with a TEDE greater than zero rem. This does not include workers with a TEDE

reported as zero, not detectable (ND), or not required to be reported (NR). [Ref. 18]

3.1.4 Collective Dose

The concept of collective dose is used in this report to denote the summation of the TEDE

received by all monitored workers and has the units person-rem. The revised 10 CFR 20.2206

requires that the TEDE be reported, so the collective dose is calculated by summing the TEDE

for all monitored workers. The phrase “collective dose” is used throughout this report to mean

the collective TEDE, unless otherwise specified.

It should be noted that prior to the implementation of the revised dose reporting requirements

of 10 CFR 20.2206 in 1994, the collective dose was, in some cases, calculated from the dose

distributions by summing the products obtained from multiplying the number of workers

reported in each of the dose ranges by the midpoint of the corresponding dose range. This

assumes that the midpoint of the range is equal to the arithmetic mean of the individual doses

in the range. Past experience has shown that the actual mean dose of workers reported in

each dose range is less than the midpoint of the range, and therefore the resultant calculated

collective doses shown in this report for these licensees may be about 10% higher than the

sum of the actual individual doses. Care should be taken when comparing the actual collective

dose calculated for 1997 with the collective dose for years prior to 1994 because of this change

in methodology. In addition, prior to 1994, doses only included the external whole body dose.

Although the contribution of internal dose to the TEDE is minimal for most licensees, it should

be taken into consideration when comparing the 1997 collective dose with the collective dose

for prior years. One noted exception is for fuel fabrication licensees where the CEDE in some

cases contributes the majority of the TEDE (see Section 3.3.5.).
3-3 NUREG-0713

3.1.5 Average Individual Dose

The average individual dose is obtained by dividing the collective dose by the total number of

workers reported as being monitored. This figure is usually less than the average measurable

dose (see below) because it includes the number of those workers who received zero or less

than measurable doses.

3.1.6 Average Measurable Dose

The average measurable dose is obtained by dividing the collective TEDE by the number of

workers who received a measurable dose. This is the average most commonly used in this

and other reports when examining trends and comparing doses received by workers in various

segments of the nuclear industry because it deletes those workers receiving zero or minimal

doses, many of whom were monitored for convenience or identification purposes.

3.1.7 Number of Licensees Reporting

The number of licensees refers to the NRC licenses issued to companies to use radioactive

material for certain activities that would place them in one of the six categories that are

required to report pursuant to 10 CFR 20.2206. The third column in Table 3.1 shows the

number of licensees that have filed such reports during the last 10 years. Agreement State

licensees do not submit such reports to the NRC and are not included in this report.

3.1.8 CR

One of the parameters that the United Nations Scientific Committee on the Effects of Atomic

Radiation (UNSCEAR) recommends be calculated for occupational dose distributions to aid in

the comparison of exposure data is a ratio “CR.” CR is defined to be the ratio of the annual

collective dose incurred by workers whose annual doses exceed 1.5 rem to the total annual

collective dose. One UNSCEAR report [Ref. 10] states that normal values of CR should be

between 0.05 and 0.50. A CR of 0.50 means that 50% of the collective dose is due to

individual doses that exceed 1.5 rem.

Prior to 1994, the value of CR was calculated from the statistical distributions that were

submitted under 10 CFR 20.407. For this calculation, it was assumed that the doses were

uniformly distributed between each dose range interval. The number of people in each dose

range above 1.5 rem was multiplied by the midpoint of the dose range to estimate the collective

dose attributed to each dose range. The collective dose of workers with doses exceeding 1.5

rem in the 1 to 2 rem range was calculated by assuming that half of the collective dose incurred

by workers with doses between 1 and 2 rem was because of doses greater than 1.5 rem. This

value was then added to the collective dose incurred by workers in the higher ranges. This

was known to yield a conservative CR value, but was a useful indicator when consistently

applied to the data from year to year.
NUREG-0713 3-4 NUREG-0713 3-4
3-5 NUREG-0713

The last column in Table 3.1 shows the values of CR for the different types of licensees. With

the implementation of the revised 10 CFR 20 in 1994, licensees were required to submit dose

records for each individual. This allowed the NRC to determine the CR value directly by

summing the collective dose for individuals with a total TEDE greater than or equal to 1.5 rem

and divide it by the collective TEDE for the licensee. This method yielded a large reduction in

the CR for Reactors. The CR value for Reactors dropped 64% from 0.22 in 1993 to 0.08 in

1994 and to 0.04 in 1997. Using the previous methodology, the CR value would have been

calculated to be 0.08 for 1997. One of the contributing factors for this difference is the

administrative controls imposed at nuclear power facilities for individuals who exceed 1 rem.

This causes the dose distribution to drop off sharply above 1 rem with fewer exposures

exceeding 1.5 rem. Therefore, the actual CR is significantly less than the value that is

calculated by assuming a uniform dose distribution.

The Manufacturing and Distribution licensees have experienced an increase in the CR value

and exceed the 0.50 value recommended by UNSCEAR. Fuel Fabrication doses, including the

CR value, have increased primarily because of the inclusion of internal exposure in the TEDE

for 1994 through 1997. However, the overall average CR for all licensees remained well below

0.50, primarily because of the low CR values at power reactor licensees. The overall average

CR remained at a value of 0.09 in 1997.

3.2 Annual TEDE Dose Distributions

Table 3.2 is a statistical compilation of the exposure reports submitted by six categories of

licensees (see Section 3.3 for a description of each licensee category). The dose distributions

are generated by summing the TEDE for each individual and counting the number of

individuals in each dose range. In nearly every category a large number of workers receive

doses that are less than measurable, and very few doses exceed 4 or 5rem. About 90% of the

reported workers continue to be monitored by nuclear power facilities where they receive

approximately 90% of the total collective dose.

Under the regulatory limits of the revised 10 CFR 20.1201, annual TEDE in excess of 5 rem for

occupationally exposed adults is, by definition, an exposure in excess of regulatory limits (see

Section 6).

Table 3.3 gives a summary of the annual exposures reported to the Commission by certain

categories of NRC licensees as required by 10CFR 20.2206. Table 3.3 shows that ~ 95% of

the exposures consistently remained <2 rem between 1968 and 1984. For the past 12 years

the percentage of workers with <2 rem has been ≥98%. The number of workers receiving an

annual exposure in excess of 5 rem had been <0.01% since 1985. 1997 is the first year

recorded where no individual received a TEDE or whole body dose in excess of 5 rem.
NUREG-0713 3-6
3-7 NUREG-0713

TABLE 3.3

SUMMARY OF ANNUAL DOSE DISTRIBUTIONS FOR CERTAIN* NRC LICENSEES

1968-1996

Total Number of Percent of Percent of Number of

Monitored Persons Individuals Individuals Individuals

Year Reported Corrected With Doses With Doses With Doses

Number Number < 2 cSv** < 5 cSv** >12 cSv**

1968 36,836 97.2% 99.5% 3

1969 31,176 96.5% 99.5% 7

1970 36,164 96.1% 99.4% 0

1971 36,311 96.3% 99.3% 1

1972 44,690 95.7% 99.5% 8

1973 67,862 95.0% 99.5% 1

1974 85,097 96.4% 99.7% 1

1975 78,713 94.8% 99.5% 1

1976 92,773 95.0% 99.6% 3

1977 98,212 93,438 93.8% 99.6% 1

1978 105,893 100,818 94.6% 99.8% 3

1979 131,027 125,316 95.2% 99.8% 1

1980 159,177 150,675 94.6% 99.7% 0

1981 157,874 149,314 94.6% 99.8% 1

1982 162,456 154,117 94.9% 99.9% 0

1983 172,927 164,239 94.6% 99.9% 0

1984 181,627 168,899 95.1% 99.9% 0

1985 212,217 201,339 97.5% >99.99% (15) 2

1986 225,582 213,017 98.0% >99.99% (8) 0

1987 243,562 227,997 98.7% >99.99% (4) 1

1988 231,234 215,662 98.6% >99.99% (8) 0

1989 229,353 212,474 98.9% >99.99% (7) 1

1990 234,045 214,781 98.9% >99.99% (3) 0

1991 219,229 206,732 99.4% >99.99% (2) 0

1992 222,728 205,009 99.4% >99.99% (1) 0

1993 209,386 189,711 99.5% >99.99% (2) 0

1994 179,803 152,834 99.5% >99.99% (1) 0

1995 179,176 143,684 99.5% >99.99% (1) 0

1996 173,536 137,968 99.5% >99.99% (1) 0

1997 180,677 128,466 99.5% 100% (0) 0

* Licensees required to submit radiation exposure reports to the NRC under 10 CFR 20.2206.

** Data for 1977-1997 are based on the distribution of individual doses after adjusting for the

multiple counting of transient reactor workers (see Section 5). The number of people exceeding 5

rem is shown in parentheses from 1985-1997.
NUREG-0713 3-8

3.3 Summary of Occupational Exposure Data by License Category

3.3.1 Industrial Radiography Licenses, Single and Multiple Locations

Industrial Radiography licenses are issued to allow the use of sealed radioactive materials,

usually in exposure devices or “cameras,” that primarily emit gamma rays for nondestructive

testing of pipeline weld joints, steel structures, boilers, aircraft and ship parts, and other

high-stress alloy parts. Some firms are licensed to conduct such activities in one location,

usually in a permanent facility that was designed and shielded for radiography, and others

perform radiography at multiple, temporary sites in the field. The radioisotopes most commonly

used are cobalt-60 and iridium-192. As shown in Table 3.1, annual reports were received for

143 radiography licensees in 1997. Table 3.4 summarizes the reported data for the two types

of radiography licenses for 1997 and for the previous 2 years for comparison purposes.

The average measurable dose for workers performing radiography at a single location ranges

from 20 to 30% of the average measurable dose of workers at multiple location facilities. This

is because it is more difficult for workers to avoid exposure to radiation in the field, where

conditions are not optimal and may change daily. To see the contribution that each

radiography licensee made to the total collective dose, a summary of the information reported

by each of these licensees in 1997 is presented in Appendix A.

Year Number of Type of License

Licenses

Number of

Monitored

Workers

Workers with

Measurable

Dose

Collective

Dose

(person-rem)

Average

Measurable

Dose (rem)

 TABLE 3.4

ANNUAL EXPOSURE INFORMATION FOR INDUSTRIAL RADIOGRAPHERS

1995 - 1997

Single Location 27 296 84 10 0.12

1997 Multiple Locations 116 3,140 2,370 1,281 0.54

Total 143 3,436 2,454 1,291 0.53

Single Location 27 291 60 10 0.17

1996 Multiple Locations 117 3,340 2,477 1,375 0.56

Total 144 3,631 2,537 1,385 0.55

Single Location 27 285 61 6 0.10

1995 Multiple Locations 112 3,245 2,404 1,332 0.55

Total 139 3,530 2,465 1,338 0.54
3-9 NUREG-0713

High exposures in radiography can be directly attributable to the type and location of the

radiography field work. For example, locations such as oil drilling platforms and aerial tanks

offer the radiographer little available shielding. In these situations, there may not be an

opportunity to use distance as a means of minimizing exposure and achieving ALARA.

Although these licensed activities usually result in average measurable doses that are higher

than other licensees, they involve a relatively small number of exposed workers.

Figure 3.1 shows the number of workers with measurable dose per licensee, the total collective

dose per licensee, and the average measurable dose per worker for both types of Industrial

Radiography facilities from 1973 through 1997.

3.3.2 Manufacturing and Distribution Licenses, Type “A” Broad and Limited

Manufacturing and Distribution licenses are issued to allow the manufacture and distribution of

radionuclides in various forms for a number of diverse purposes. The products are usually

distributed to persons specifically licensed by the NRC or an Agreement State. Type “A” Broad

licenses are issued to larger organizations that may use many different radionuclides in many

different ways and that have a comprehensive radiation protection program. The Limited

licenses are usually issued to smaller firms requiring a more restrictive license. Some firms are

medical suppliers that process, package, or distribute such products as diagnostic test kits,

radioactive surgical implants, and tagged radiochemicals for use in medical research,

diagnosis, and therapy. Limited firms are suppliers of industrial radionuclides and are involved

in the processing, encapsulation, packaging, and distribution of the radionuclides that they

have purchased in bulk quantities from production reactors and cyclotrons. Major products

include gamma radiography sources, cobalt irradiation sources, well-logging sources, sealed

sources for gauges and smoke detectors, and radiochemicals for nonmedical research.

However, only those NRC licensees that possess or use at any one time specified quantities of

the nuclides listed in paragraph 20.2206(a)(7) are required to submit reports to the NRC.

Table 3.5 presents the annual data that were reported by the two types of licensees for 1997

and the previous 2 years. Looking at the information shown separately for the Type “A” Broad

and Limited licensees, it can be seen that the values of all of the parameters remain higher for

the Broad licensees. However, when attempting to examine trends in the data presented for

this category of licensees, it should be noted that the types and quantities of radionuclides

may fluctuate from year to year, and even during the year, so that some licensees may report

dose data one year and not the next and may be included as a Broad licensee one year and a

Limited licensee at other times. Because the number of reporting licensees is quite small,

these fluctuations may have a significant impact on the values of the parameters.
NUREG-0713 3-10
3-11 NUREG-0713

M & D-”A”-Broad 5 496 386 364 0.94

1997 M & D-Limited 26 655 279 33 0.12

Total 31 1,151 665 397 0.60

M & D-”A”-Broad 7 2,018 987 522 0.53

1996 M & D-Limited 29 610 252 34 0.13

Total 36 2,628 1,239 556 0.45

M & D-”A”-Broad 7 2,016 909 557 0.61

1995 M & D-Limited 29 650 313 38 0.12

Total 36 2,666 1,222 595 0.49

Figure 3.2 shows the number of workers with measurable dose per licensee, the total collective

dose per licensee, and the average measurable dose per worker for both Type “A” Broad and

Limited Manufacturing and Distribution facilities. The average measurable dose for Type “A”

Broad licensees increased by 77% from 1996 to 1997 primarily due to the increase in average

measurable dose at Mallinckrodt Medical, Inc.. In addition, three Type “A” Broad licensees that

have reported significant dose in prior years, were transferred to Agreement State licensees in

the Commonwealth of Massachusetts.

For the contribution that each of these licensees made toward the total values of the number of

workers monitored, number of workers, and collective dose, see Appendix A, which lists the

values of these parameters for each licensee for 1997.

 TABLE 3.5

ANNUAL EXPOSURE INFORMATION FOR MANUFACTURERS AND DISTRIBUTORS

1995 - 1997

Year Number of Type of License

Licenses

Number of

Monitored

Workers

Workers with

Measurable

Dose

Collective

Dose

(person-rem)

Average

Measurable

Dose (rem)
NUREG-0713 3-12
3-13 NUREG-0713

3.3.3 Low-Level Waste Disposal Licenses

Low-Level Waste Disposal licenses are issued to allow the receipt, possession, and disposal

of low-level radioactive wastes at a land disposal facility. The licensee has the appropriate

facilities to receive wastes from such places as hospitals and laboratories, store them for a

short time, and dispose of them in a properly prepared burial ground. The licensees in this

category are located in and licensed by Agreement States which have primary regulatory

authority over its activity. However, these licensees also have an NRC license that covers

certain special nuclear material they might receive. The annual dose reports submitted by

these licensees include all doses received during the year regardless of whether they were the

result of NRC or Agreement State licensed material.

The requirement for this category of NRC licensee to file annual reports became effective in

January 1983. There was only one licensee in this category in 1982 and 1983 and two

licensees in this category from 1984 to 1997. Table 3.1 summarizes the data reported for

1988 through 1997. Appendix A summarizes the exposure information reported by this

licensee in 1997.

Figure 3.3 shows the number of workers with measurable dose per licensee, the total collective

dose per licensee, and the average measurable dose per worker for Low-Level Waste

Disposal facilities from 1982 through 1997. Because only two licensees have been involved in

this activity over the past 10 years, the numbers have remained fairly stable from 1984

through 1997 with the exception of the average measurable TEDE, which peaked in 1992 and

has decreased by 75% since then.

3.3.4 Independent Spent Fuel Storage Installation Licenses

Independent Spent Fuel Storage Installation (ISFSI) licenses are issued to allow the

possession of power reactor spent fuel and other associated radioactive materials for the

purpose of storage of such fuel in an ISFSI. Here, the spent fuel, which has undergone at

least 1 year of decay since being used as a source of energy in a power reactor, is provided

interim storage, protection, and safeguarding for a limited time pending its ultimate disposal.

Eighteen licenses have been issued for these activities. Eleven are at nuclear power plants,

allowing on-site temporary storage of fuel. These licensees report the dose from fuel storage

activities along with the dose from reactor operations at these sites. Out of the seven

remaining licenses, only one is active and is located at a facility that is independent of a

reactor site. Only this licensee is included in this analysis of ISFSI facilities for 1997. Appendix

A summarizes the exposure information reported by this installation.

Figure 3.4 shows the number of workers with measurable dose per licensee, the total collective

dose per licensee, and the average measurable dose per worker for Independent Spent Fuel

Storage facilities. The large increase in the collective dose per licensee and number of
NUREG-0713 3-14
3-15 NUREG-0713

workers per licensee in 1994 was mainly because only one licensee reported separately for

1994 through 1997, rather than the two licensees that reported in prior years. All parameters

have decreased significantly from 1996 to 1997.

3.3.5 Fuel Cycle Licenses

Fuel cycle licenses are issued to allow the processing, enrichment, and fabrication of reactor

fuels. In most uranium facilities where light water reactor fuels are fabricated enriched uranium

hexafluoride is converted to solid uranium dioxide pellets and inserted into zirconium alloy

tubes. The tubes are fabricated into fuel assemblies that are shipped to nuclear power plants.

Some facilities also perform chemical operations to recover the uranium from scrap and other

off-specification materials prior to disposal of these materials. For 1997, this category also

includes the two uranium enrichment facilities at Portsmouth, Ohio, and Paducah, Kentucky.

The regulatory oversight for these facilities was transferred from the U. S. Department of

Energy to the NRC in 1997.

Figure 3.5 shows the number of workers with measurable dose per licensee, the total collective

dose per licensee, and the average measurable dose per worker for Fuel Cycle licensees. In

addition to the TEDE collective and average measurable dose, the Deep Dose Equivalent

(DDE) collective dose and DDE average measurable dose are shown. Both doses are shown

since the CEDE is a significant contribution to the TEDE for Fuel Fabrication facilities.

Appendix A lists each of the licensees reporting in 1997, with the number of workers

monitored, the number of workers receiving measurable external doses, and the collective

dose for each licensee. Table 3.6 shows that there were 10 licensed Fuel Cycle (Fabrication

and Enrichment) facilities in 1997.

1997 Fuel Cycle 10 11,214 3,910 1,006 0.26 197 0.08 800 0.30

1996 Fuel Cycle 8 4,369 3,061 878 0.29 161 0.08 711 0.32

1995 Fuel Cycle 8 4,106 2,959 1,217 0.41 131 0.10 990 0.33

 TABLE 3.6

ANNUAL EXPOSURE INFORMATION FOR FUEL CYCLE LICENSES

1995 - 1997

Type of

License

Year Number

of

Licenses

Number of

Monitored

Workers

Workers

with

Meas.

TEDE

Collective

TEDE

(personrem)

Average

Meas.

TEDE

(rem)

Collective

DDE

(personrem)

Average

Meas.

DDE

(rem)

Collective

CEDE

(personrem)

Average

Meas.

CEDE

(rem)
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3-17 NUREG-0713
NUREG-0713 3-18

3.3.6 Light-Water-Cooled Power Reactor (LWR) Licenses

LWR licenses are issued to utilities to allow them to use special nuclear material in a reactor

that produces heat to generate electricity to be sold to consumers. There are two major types

of commercial LWRs in the United States - pressurized water reactors (PWRs) and boiling

water reactors (BWRs) - each of which uses water as the primary coolant.

Table 3.1 shows the number of licensees, total number of monitored workers, the number of

workers with measurable dose, the total collective dose, and average dose per worker for all

reports received from reactor facilities that were in commercial operation for the years 1988

through 1997. This table includes reactors that may not have been in commercial operation for

a full year. Data for 1988 includes all reactors that reported, even though some of them were

shut down. Data for 1989 through 1997 do not include reactors that have been shut down.

These figures have been adjusted for the multiple counting of transient workers (see

Section5). The reported dose distribution of workers monitored at each plant site is presented

in alphabetical order by site name in AppendixB.

More detailed presentations and analyses of the annual exposure information reported by

nuclear power facilities can be found in Sections 4 and 5.

3.4 Summary of Intake Data by License Category

With the revision of 10 CFR 20 in 1994, licensees were required to report additional data to the

NRC concerning intakes of radioactive material. Licensees were required to list for each intake

the radionuclide that was taken into the body, the pulmonary clearance class, intake mode,

and amount of the intake in microcuries. An NRC Form 5 report containing this information is

required to be completed and submitted to the NRC under 10 CFR 20.2206.

Tables 3.7 and 3.8 summarize the intake data reported to the NRC during 1997. The data are

categorized by licensee type and are listed in order of radionuclide and pulmonary clearance

class. Table 3.7 lists the intakes where the mode of intake into the body was recorded as

ingestion. In 1997, one record was reported as an ‘absorption’ of tritium and is included in

Table 3.7. Table 3.8 lists the intakes where the mode of intake was inhalation from ambient

airborne radioactive material in the workplace. The pulmonary clearance class is recorded as

D, W, or Y corresponding to its clearance half-time in the order of days, weeks, or years from

the pulmonary region of the lung into the blood and gastrointestinal tract. The amount of

material taken into the body is given in microcuries, a unit of measure of the quantity of

radioactive material. For each category of licensee, the maximum number of intake records

and the maximum intake is highlighted in the table in bold for ease of reference.
3-19 NUREG-0713
NUREG-0713 3-20
3-21 NUREG-0713
NUREG-0713 3-22

Table 3.9 lists the number of individuals with measurable CEDE, the collective CEDE and the

average measurable CEDE in descending order of the number with measurable CEDE. Fuel

fabrication facilities have the majority of internal dose (99%) and the highest average CEDE

per individual. This is due to the worker’s exposure to uranium during the processing and

fabrication of the uranium fuel.

Table 3.10 shows the distribution of internal dose (CEDE) from 1994 to 1997 for licensees

required to report under 10 CFR 20.2206. For the purposes of this table, the definition of a

‘measurable CEDE’ is any reported value greater than zero. As noted above, the vast majority

of the internal doses are received by individuals working at fuel fabrication facilities. The table

shows that the number with measurable CEDE remained at nearly 3,000 from 1994 to 1996,

and then increased to 3,739 in 1997. However, the average measurable CEDE has

decreased 31% in the past four years, from 0.316 rem in 1994 to 0.217 rem in 1997.
3-23 NUREG-0713
NUREG-0713 3-24
4-1 NUREG-0713

4 COMMERCIAL LIGHT WATER REACTORS - FURTHER ANALYSIS

4.1 Introduction

General trends in occupational radiation exposures at nuclear power reactors are best

evaluated within the context of other pertinent information. In this chapter, some of the tables

and appendices that summarize exposure data also show the type, capacity, and age of the

reactor; the amount of electricity generated; the types of workers being exposed; and the sort

of tasks being performed. Exposure data are then presented as a function of these data.

4.2 Definition of Terms and Sources of Data

4.2.1 Number of Reactors

The number of reactors shown in Tables 4.1, 4.2, and 4.3 is the number of BWRs, PWRs, and

LWRs, respectively, that had been in commercial operation for at least 1 full year as of

December 31 of each of the indicated years. This is the number of reactors on which the

average number of workers with measurable dose and average collective dose per reactor is

based. Excluded are those reactors that had been in commercial operation for less than 12

months during the first year and reactors that have been permanently defueled. This yields

conservative values for many of the averages shown in the tables. The date that each reactor

was declared to be in commercial operation was taken from Reference 14.

Three Mile Island (TMI) 2 had been included in the compilation of data for commercially

operating reactors through 1988 even though the reactor was shut down following the 1979

accident and has been in the process of defueling and decommissioning since that time. TMI

2 has not been included in the data analysis since 1988. Data for this reactor, however, will be

listed in Appendices B, C, D and E for reference purposes.

In 1997, Haddam Neck (a PWR) was removed from the count of operating reactors and Watts

Bar Unit 1 (also a PWR) was added to the count, keeping the total count of operating reactors

at 109. Three sites permanently ceased operation during 1997. These plants were kept in the

count of operating reactors for 1997 even though they were not in commercial operation for

the entire year. These plants are Maine Yankee (a PWR shut down 8/97), Big Rock Point (a

BWR shut down 9/97) and Zion 1,2 (two PWRs shut down 12/97). Maine Yankee and Zion 2

produced no power during all of 1997 and Zion 1 generated only 12% of the unit’s maximum

dependable capacity. Big Rock Point is a relatively small plant, and only generated 45% of the

unit’s maximum dependable capacity prior to shut down. Therefore, the inclusion of these

plants in the count of operating reactors for 1997 resulted in a decrease in the overall average

electricity generated per reactor and an increase in the collective dose per megawatt-year for

PWRs.
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4-5 NUREG-0713

4.2.2 Electric Energy Generated

The electric energy generated in megawatt-years (MW-yr) each year by each facility is shown

in Appendix C and graphically represented in AppendixE. This number was obtained by

dividing the megawatt-hours of electricity annually produced by each facility by 8,760, the

number of hours in the year, except for leap years when the number is 8,784 hours. For the

years 1973 to 1996, the electricity generated is the gross electricity output of the reactor. In

1997, the number reflects the net electricity produced which is the gross electricity minus the

amount the plant uses for operations. This change is the result of a change in the NRC power

generation reporting requirements. The electricity generated (in megawatt-years) that is

presented in Tables 4.1, 4.2, and 4.3 is the summation of electricity generated by the number

of reactors included in each year. These sums are divided by the number of reactors included

in each year to yield the average amount of electric energy generated per reactor, which is

also shown in Tables 4.1, 4.2, and 4.3. The number of megawatt-hours of electricity produced

each year was found in Reference 14.

As shown in Table 4.3, there was an 11% decrease in the electricity generated in 1997.

Approximately 4% of this decrease is due to the change in reporting requirements from the

gross electricity generated to the net electricity generated. The remaining 7% of the decrease

is due to reductions in the power generated. Three BWR sites (Clinton, Lasalle 1 and 2, and

Millstone Point 1) and four PWR sites (Crystal River, Maine Yankee, Millstone Point 2 and 3,

and Zion 1 and 2) generated little or no power in 1997. Reasons for the outages at these sites

included maintenance, refueling, and regulatory restrictions.

4.2.3 Collective Dose per Megawatt-Year

The number of megawatt-years of electricity generated was used in determining the ratio of

the average value of the annual collective dose (TEDE) to the number of megawatt-years of

electricity generated. The ratio was calculated by dividing the total collective dose in

person-rem by the electric energy generated in megawatt-years and is a measure of the dose

incurred by workers at power plants in relation to the electric energy produced. For the years

1973 to 1996, the electricity generated is the gross electricity output of the reactor. In 1997,

the number reflects the net electricity produced. This ratio was also calculated for each reactor

site and is presented in Tables 4.1, 4.2, and 4.3 and Appendix C.

4.2.4 Average Maximum Dependable Capacity

Average maximum dependable capacity, shown in Tables 4.1, 4.2, and 4.3, was found by

dividing the sum of the net maximum dependable capacities of the reactors in megawatts (net

MWe) by the number of reactors included each year. The net maximum dependable capacity

is defined as the gross electrical output as measured at the output terminals of the turbine

generator during the most restrictive seasonal conditions, less the normal station service loads.
NUREG-0713 4-6

This “capacity” of each plant was found in Reference 14, and it is shown for each site in

Appendix C.

4.2.5 Percent of Maximum Dependable Capacity Achieved

The percent of maximum dependable capacity achieved is shown for all LWRs in Table 4.3.

This parameter gives an indication of the overall power generation performance of LWRs as

compared to the maximum capacity that could be obtained in a given year. It is calculated by

dividing the average electricity generated per reactor by the average maximum dependable

capacity for each year.

From 1973 to 1978 this indicator exhibited an increasing trend as a number of new reactors

began producing power at higher efficiencies. Following the accident at Three Mile Island,

reactor operations personnel concentrated on improving safety systems and complying with

the new regulations for these systems. During this time period, from 1979 to 1987, the percent

of maximum dependable capacity remained around 61%. Following the completion of most of

these mandated repairs, reactors have increased the percent of maximum dependable

capacity from 62% in 1987 to 81% in 1996, a gain of nearly 20% in 10 years. For 1997, the

number dropped down to 72% due to the change from measuring the gross electricity

generated to the net electricity generated, and a decrease in the electricity generated due to

outages for maintenance, refueling, and regulatory restrictions (see Section 4.2.2).

4.3 Annual TEDE Distributions

Table 4.4 summarizes the distribution of the annual TEDE doses received by workers at all

commercial LWRs during each of the years 1977 through 1997. This distribution is the sum of

the annual dose distributions reported by each licensed LWR each year. As previously noted,

the distribution reported by each LWR site for 1997 is shown in Appendix B. Table 4.4 shows

the reported dose distributions corrected for the number of transient workers that were

reported by more than one site (see Section 5). The total collective dose decreased by 9% to

a value of 17,136 person-rem in 1997. The value of CR decreased to a value of 0.04. The

large decrease in the value of CR from 1993 to 1994 is primarily because of the change in

methodology by which the CR value is determined. CR is defined to be the ratio of the annual

collective dose. CR is one of the parameters that the United Nations Scientific Committee on

the Effects of Atomic Radiation (UNSCEAR) recommends be calculated for occupational dose

distributions to aid in the comparison of exposure data. Once UNSCEAR report [Ref. 10]

states that normal values of CR should be between 0.05 and 0.50. A CR of 0.50 means that

50% of the collective dose is due to individual doses that exceed 1.5 rem. For the years 1994

to 1997, the CR value was determined directly from the individual radiation exposure records

submitted under 10 CFR 20.2206 (Form 5) rather than calculating the value indirectly from the

statistical dose distribution summary as in prior years. This is the thirteenth consecutive year

that the value of CR has been <0.50.
4-7 NUREG-0713
NUREG-0713 4-8

4.4 Average Annual TEDE Doses

Some of the data presented in Tables 4.1, 4.2, and 4.3 are graphically displayed in Figure 4.1,

where it can be seen that the average collective dose and average number of workers per

BWR have been higher than those for PWRs since 1974 and that the values of both

parameters, in general, continued to rise at both types of facilities until 1983. Between 1983

and 1997, the average collective dose per reactor dropped by 79%. In 1997, the collective

dose per reactor for PWRs increased by 1% to 132 person-rem. The collective dose per

reactor for BWRs decreased by 20% to 205 person-rem in 1997. The overall collective dose

per reactor for LWRs decreased by 9% to 157 person-rem in 1997. The number of workers

with measurable dose per reactor decreased to 915 for BWRs but increased to 703 for PWRs

in 1997. However, the overall decreasing trend in average reactor collective doses since 1983

indicates that licensees are continuing to successfully implement ALARA dose reduction

features at their facilities.

Figures 4.2 and 4.3 are plots of most of the other information that is given in Tables 4.1, 4.2,

and 4.3. The value for the total collective dose for all LWRs decreased by 9% from a value of

18,874 person-rem in 1996 to 17,136 person-rem in 1997. Together with the slight increase in

the number of workers with measurable dose, this resulted in the average measurable dose

per worker decreasing to 0.20 rem in 1997. Figure 4.2 shows that in 1997 the net electricity

generated was 71,851 megawatt-years.

The fluctuations in the parameters for the years following the accident at the TMI plant in 1979

may reflect some of the impact that this incident had on the nuclear power industry. The

decrease seen in dose trends since 1983 may be attributable to several factors. Utilities have

completed most of the tasks initiated as a result of the lessons learned from the Three Mile

Island accident, and they are increasing efforts to avoid and reduce exposure. The importance

of exposure control and the concept of keeping exposures to ALARA levels is continually being

stressed, and most utilities have established programs to collect and share information relative

to tasks, techniques, and exposures.

To further assist in the identification of any trends that might exist, Figure4.4 displays the

average and median5 values of the collective dose per reactor for BWRs and for PWRs for the

years 1973 through 1997. The ranges of the values reported each year are shown by the

vertical lines with a small bar at each end marking the two extreme values. The rectangles

indicate the range of values of the collective dose exhibited by those plants ranked in the

twenty-fifth through the seventy-fifth percentiles. Since the median values usually are not as

greatly affected by the extreme values of the collective doses, they do not normally fluctuate as

much from year to year as do the average values. The median collective dose for PWRs

5 The value at which 50% of the reactors reported greater collective doses and the other 50% reported smaller collective doses.
4-9 NUREG-0713

197319741975197619771978197919801981198219831984198519861987198819891990199119921993199419951996199

0

200

400

600

800

1000

1200

Collective Dose per Reactor (Person-rem)

Year

BWR

PWR

LWR

Figure 4.1

Average Collective Dose and Number of Workers per Reactor 1973 – 1997

Average Annual Collective Dose

1973197419751976197719781979198019811982198319841985198619871988198919901991199219931994199519961997

0

200

400

600

800

1000

1200

1400

1600

Workers per Reactor

Year

BWR

PWR

LWR

Average Annual Number of Workers with Measurable Dose
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1973197419751976197719781979198019811982198319841985198619871988198919901991199219931994199519961997

0

10

20

30

40

50

60

70

80

90

100

110

120

Number of Reactors

Year

BWR

PWR

Figure 4.2

Number of Operating Reactors and Gross Electricity Generated 1973 – 1997

Number of Operating Reactors

1973 1974 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997

0

10

20

30

40

50

60

70

80

90

Megawatt-Years (Thousands)

Year

BWR

PWR

LWR

Electricity Generated*

*Gross electricity 1973-1996, Net electricity for 1997.
4-11 NUREG-0713

1973197419751976197719781979198019811982198319841985198619871988198919901991199219931994199519961997

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

1.1

Average Measurable Dose (rem)

Year

BWR

PWR

LWR

Figure 4.3

Average Measurable Dose per Worker and Collective Dose per Megawatt-Year 1973 – 1997

Average Measurable Dose per Worker

1973 1974 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1996 1997

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

2.4

2.6

2.8

3.0

Collective Dose (Person-rem) per MW-YR

Year

BWR

PWR

LWR

Collective Dose per Megawatt-Year*

*Gross electricity 1973-1996, Net electricity for 1997.
NUREG-0713 4-12 Figure 4.4 Average, Median, and Extreme Values of the Collective Dose Per Reactor 1973 – 1997

M

Collective Doses Per Reactor (person-rem)

500

5500500045004000350030002500200015001000

0

1973 1974 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992

Year

A

M

A

M

A

M A

M

M

A

A

M

A

M A

M

A

M

A

M

A

M

A

M

A

M

A

M

A

M

A

M

A

M

A

M

MA

A

M

M

A

A

M

A

M

A

M

A

M

A

M

A

M

A

M

A

M A

M

A

M A

M

A

M

A

M

MA

MA A

M A

M

A

M

A

M

Middle 50% of BWRsMiddle 50% of PWRsAverage Collective DoseMedian Collective Dose

A

M

1993

A

M

1994

A

A

M

M

1995

A

A

M A

M

1996

A

M

A

M

1997

A

M

MA
4-13 NUREG-0713

experienced a slight increase from 120 person-rem in 1996 to 121 person-rem in 1997. At

BWRs, the median fluctuates more from year to year, and in 1997 the median collective dose

decreased to 206 person-rem. Figure 4.4 also shows that, in 1997, 50% of the PWRs

reported collective doses between 74 and 163 person-rem while 50% of the BWRs reported

collective doses between 117 and 245 person-rem. Nearly every year, the median collective

dose is less than the average, which indicates that the collective dose for most plants is less

than the average collective dose per reactor (the value that is widely quoted).

4.5 Plant Rankings by Collective Dose per Reactor

Because the number of reactors from which data have been collected is statistically rather

small, the information reported by a few reactors where unusual conditions or problems may

have occurred could have a large impact on some of the statistics presented in this report. In

an effort to identify those plants, Tables 4.5 and 4.6 list the BWRs and PWRs in ascending

order of collective dose per reactor for each of the 5 years from 1993 through 1997. The total

collective dose per site is listed in the tables even though the dose per reactor was used for

the ranking. Two other parameters, average measurable dose per worker and collective dose

per megawatt-year, are also given for each plant. Also shown is a parameter CR, which is

defined as the ratio of the annual collective dose delivered at individual doses exceeding 1.5

rem to the total annual collective dose. The value of CR has continued to decline for most

plants, and in 1997, the CR for all the U.S. LWRs fell to 0.04 which is the first time that the CR

has been below the 0.05 to 0.50 range recommended by the UNSCEAR [Ref. 10]. Note that

for 1994 through 1997, the CR value was determined directly from the individual radiation

exposure records submitted under 10 CFR 20.2206 (Form 5) rather than calculating the value

from the statistical dose distribution summary (see Section 3.1.8).

Tables 4.7a and b list the sites that had been in commercial operation for at least 5 years as of

December 31, 1997, and show the values of several parameters for each of the sites. They

also give averages for the two types of reactors. Based on the 185 reactor-years of operation

accumulated by the 37 BWRs listed, the average annual collective dose per reactor was found

to be 275 person-rem, the average measurable dose per worker was 0.27 rem, and the

average collective dose per megawatt-year was 0.4.

Based on the 345 reactor-years of operation at the 69 PWRs listed, the average annual

collective dose per reactor, average measurable dose per worker, and average collective dose

per megawatt-year were found to be 153 person-rem, 0.22 rem, and 0.2 person-rem/MW-yr,

respectively. All of these values, at both types of facilities, are lower than those found for the 5

year period ending in 1996, with the exception of the average collective dose per megawattyear at PWRs, which remained the same.
NUREG-0713 4-14
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NUREG-0713 4-16
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NUREG-0713 4-18

In some cases, the plants having the lower values for most of the parameters shown in Tables

4.7a and b are the newer plants. Some of the older, smaller plants, such as Big Rock Point,

also appear near the top of the listings because they report small collective doses. However,

the ratio of collective dose to megawatt-years is generally higher for these plants because of

their limited power generation capability.

The largest contributor to the collective dose is usually associated with outages at a site. In

analyzing collective dose trends, it is useful to examine the outage data for reactors to look for

a relationship between the collective dose and the outage information for the reactors. Figure

4.5 displays the total number of outage days for BWRs and PWRs respectively. The collective

dose and average measurable dose are also plotted to allow for the comparison of outage

duration to collective dose.

4.6 Collective Dose by Work Function and Employee Type

Each plant is required by its Technical Specifications to submit an annual report in accordance

with Regulatory Guide 1.16 that provides the collective dose of workers monitored at each

plant site by employee type (plant, utility, or contractor) and by work and job functions. A copy

of the report submitted for each reactor site is provided in Appendix D, and much of the data

are graphically represented for each site in Appendix E. Tables 4.8 through 4.13 summarize

the 1997 data for BWRs, PWRs, and LWRs. Table 4.8 shows that, at both BWRs and PWRs,

about 67% of the collective dose is incurred during routine and special maintenance activities.

Also, the portion of the collective dose incurred during most of the other activities is similar at

the two types of plants.

One should note that the collective doses obtained from these reports are not used in any

other tables in this document. This is because the Technical Specifications of each plant

require only 80% of the plant’s collective dose be accounted for, and some utilities may use

the results of self-reading pocket dosimeters instead of the results of the dosimeter of record

(usually thermoluminescent dosimeters) in compiling the data. Also, when examining the

number of personnel shown on these reports, it should be remembered that individuals who

perform tasks in more than one category may be counted more than once.

Table 4.9 shows that for the past 10 years, the percentage of collective dose attributed to

routine maintenance has been greater than that of special maintenance. This may be

indicative of a trend showing a reduction in TMI-related activities and a greater emphasis on

steady-state routine maintenance. Overall, values have been fairly stable over the years with

these two categories, special maintenance and routine maintenance, always accounting for

the majority of the collective dose. Some of the fluctuations shown in the percentage of the

dose incurred during refueling activities (particularly in 1992 through 1995, when it increased to

over 11%) is due to the fact that some sites include doses other than those directly associated

with fuel movement in this category. Figure 4.6 graphically shows the trends in the collective
4-19 NUREG-0713

1987 1988 1989 1990 1990 1992 1993 1994 1995 1996 1997

0

2,000

4,000

6,000

8,000

10,000

12,000

14,000

16,000

18,000

0.00

0.10

0.20

0.30

0.40

0.50

0.60

Collective Dose (person-rem)

Average Measurable Dose per Worker (rem)

Year

Collective Dose (person-rem)

Total Outage Days, All Reasons

Refueling Outage Days

Average Measurable Dose per Worker (rem)

Figure 4.5

 Outage Days, Average Dose, and Collective Dose

PWR

1987 1988 1989 1990 1990 1992 1993 1994 1995 1996 1997

0

5,000

10,000

15,000

20,000

25,000

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0.40

0.45

0.50

Collective Dose (Person-rem)

Average Measurable Dose per Worker (rem)

Year

Collective Dose (person-rem)

Total Outage Days, All Reasons

Refueling Outage Days

Average Measurable Dose per Worker (rem)

BWR

Total Outage Days, Refueling Days Total Outage Days, Refueling Days
NUREG-0713 4-20
4-21 NUREG-0713
NUREG-0713 4-22

Figure 4.6

Collective Dose by Work Function and Personnel Type 1990 – 1997

Boiling Water Reactors

Pressurized Water Reactors

1997

1996

1995

1994

1993

1992

1991

1990

1997

1996

1995

1994

1993

1992

1991

1990

6

4

2

0

Year

Special Maintena

Routine Maintenan

n

Reactor Ops. & Surv.

c

In-Service Inspection

Waste Processing

Refueling

Plant

Contract

DOSE (Person-rem)

(Thousands)

1997

1996

1995

1994

1993

1992

1991

1990

1997

1996

1995

1994

1993

1992

1991

1990

6

4

2

0

Year

Special Maintena

Routine Maintenan

n

Reactor Ops. & Surv.

ce

In-Service Inspection

Waste Processing

Refueling

Plant

Contract

DOSE (Person-rem)

(Thousands)
4-23 NUREG-0713

dose by work function and type of personnel for the years 1990 through 1997 for BWRs and

PWRs separately. The general decrease in collective dose is also apparent among most of

these activities.

Table 4.10 presents the distribution of the collective dose for 1997 at all LWRs among five

occupational categories. As in past years, maintenance personnel incurred the majority (65%)

of the collective dose with contractor maintenance personnel receiving about twice as much as

the station maintenance employees combined. None of the values listed changed significantly

from those found for 1987 through 1996. The collective doses shown in Tables 4.8 and 4.10

do not equal those shown in other tables in the report because they are the sum of the doses

taken from the type of annual reports shown in Appendix D rather than the collective dose that

was calculated from the annual reports submitted pursuant to 10 CFR 20.2206.

Another use made of the reports submitted under Regulatory Guide 1.16 shown in Appendix D

is in proportioning the collective dose obtained from the § 20.2206 annual reports into the work

functions and personnel types shown in Appendix C. This was done in the following way:

(1) The collective dose incurred by workers in the work function “Reactor Operations and

Surveillance” on each plant’s annual report submitted pursuant to their technical

specifications (the first number in the last column in Appendix D) was determined.

(2) The ratio of this dose to the total collective dose (the last number in the last column in

Appendix D) was calculated and multiplied by the total collective dose that had been

obtained from the § 20.2206 annual reports. This product is the collective dose shown

in the column headed “Operations” in Appendix C.

 (3) The collective dose shown in the column headed “Maintenance and Others” in

Appendix C was determined by first summing the collective doses incurred by workers

in the five remaining functions given in Appendix D and then calculating the fraction

that this dose is of the total collective dose. This fraction was multiplied by the total

collective dose calculated from the § 20.2206 annual reports to yield the collective dose

shown in this column of Appendix C.

(4) A similar procedure was followed in determining the collective dose for the columns

headed “Contractor” and “Station & Utility” in Appendix C.
NUREG-0713 4-24
4-25 NUREG-0713

4.7 Number of Personnel by Work Function and Employee Type

Half of the information presented in the statistical annual reports shown in Appendix D

concerns the number of various types of personnel that performed certain work functions.

Tables 4.11 and 4.12 sum this information to show the percentage of personnel by work

function and occupation. The major problem in interpreting the numbers shown in these tables

is that the same person may perform several work functions during the year so that the total

number of personnel obtained by summing those shown in the various work functions would

be inflated. However, Table 4.12 is still useful in showing the percentage of personnel

associated with each of the six work functions shown. About 51% of the personnel performed

routine or special maintenance functions, 28% were involved with reactor operations and

surveillance, and the remaining 21% were divided among the other three work functions.

Table 4.12 shows the percentage of personnel in each of five occupational categories at

BWRs, PWRs, and LWRs. The workers were similarly distributed at BWRs and PWRs. The

largest difference occurred in the health physics personnel for 1997. Health physics personnel

at PWRs received about twice the percentage of the collective dose than for BWRs. Overall,

54% of the personnel were contractors, 38% were station employees, and 8% were utility

employees in 1997.

Table 4.13 presents the average annual dose incurred by workers in the five occupational

categories in 1997. These averages were calculated by dividing the collective dose reported

for these groups (see Table 4.11) by the number of individuals shown in Table 4.12. It shows

that, in most instances, the maintenance personnel incur the highest average doses.

Examination of the values of the averages given in Table 4.13 is subject to several sources of

error: (1) the number of individuals may be inflated because the same plant contractor

employee may work at several plants so that the employee would be counted more than once

in a summary such as Table 4.13; (2) the occupations are not clearly defined so that workers

performing certain tasks in one plant may be classified as being in one occupation and be

included in a different one at another plant; and (3) some plants count only those workers

whose doses exceed 0.10 rem while other plants count all workers regardless of the dose

received. Because of these factors, the usefulness of the numbers of individuals obtained

from the reports provided in Appendix D is limited; therefore, they are not used to develop any

other statistics in this document.
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4.8 Graphical Representation of Dose Trends in Appendix E

Each page of AppendixE presents two types of graphs for one site. One graph plots selected

dose-performance indicators from 1973 through 1997, and the other indicates the collective

dose by job function for 1978 through 1997. The dose and performance indicators shown in

the top graph illustrate the history of the collective dose per reactor for the site, the rolling 3-

year average collective dose per reactor, and the electricity generated at the site. These data

are plotted, beginning with the plant’s first full year of commercial operation, and continuing

through 1997. Data for years when the plant was not in commercial operation have been

included when available. However, any data reported prior to 1973 are not included. The

3-year average collective dose per reactor data is included because it provides a better overall

indication of the plant’s general trend in collective dose. This average is determined by

summing the collective dose for the current year and the previous 2 years and then dividing

this sum by the number of reactors reporting during those years. Depicting dose trends using

a 3-year average reduces the sporadic effects on annual doses of refueling operations (usually

a 2- to 3-year cycle) and occasional high-dose maintenance activities, and gives a better idea

of collective dose trends over the life of the plant. The annual average collective dose per

reactor for all reactors of the same type is also shown on the graph.

The second type of graph at the bottom of each page in Appendix E displays the breakdown of

collective dose by job function and employee type for the years 1978 through 1997. The

horizontal axis lists the six job functions of reactor operations, routine maintenance, in-service

inspection, special maintenance, waste management, and refueling operations, and the

vertical axis indicates collective dose at each site. This representation shows the job functions

where most of the dose was accumulated as well as the division of the collective dose

between plant and contract workers. The data are taken from the submittals presented in

Appendix D and therefore represent at least 80% of the collective dose at each site. Only

those reactors that have completed at least 1 full year of commercial operation are presented

in Appendix E.
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5 TRANSIENT WORKERS AT NRC LICENSED FACILITIES

5.1 Termination Reports

Under the revised 10 CFR 20, licensees are required to submit NRC Form 5s to the

Commission for each individual who is required to be monitored at the end of the monitoring

year or upon the individual’s termination of employment at the facility. The “termination reports”

submitted in accordance with the old § 20.408, listing the individual’s complete dose history

during employment at the facility, are no longer required.

However, the Form 5s submitted to the NRC upon an individual’s termination of employment

serve the same function as the previous requirements with regard to the analysis of transient

workers at NRC-licensed facilities. The following analysis examines the workers who had more

than one Form 5 dose record at more than one NRC-licensed facility during the monitoring

year. These workers are defined to be transient in that they worked at more than one facility

during the monitoring year.

The term “monitoring year” is used here in accordance with the definition of a year given in

§ 20.1003, which defines a year as “the period of time beginning in January used to determine

compliance with the provisions of this part. The licensee may change the start date of the

monitoring year used to determine compliance provided that the change is made at the

beginning of the monitoring/calendar year and that no day is omitted or duplicated in

consecutive years”.

5.2 Transient Workers at NRC Facilities

Examination of the data reported for workers who began and terminated two or more periods of

employment with two or more different facilities within one monitoring year is useful in many

ways. For example, the number and average dose for these “annual transients” can be

determined from examining these data.

Additionally, the distribution of the doses received by transient workers can be useful in

determining the impact that the inclusion of these individuals in each of two or more licensees’

annual reports has on the annual summary (as reported in Appendices B and F) for all nuclear

power facilities, and all NRC licensees combined (one of the problems mentioned in Section 2).

Table 5.1 shows the “actual distribution” of transient worker doses as determined from the

above-mentioned Form 5 termination reports and compares it with the “reported distribution” of

the doses of these workers as they would have appeared in a summation of the annual reports

submitted by each of the licensees.
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Because >95% of these transients are reported by nuclear power facilities, these data were

considered separately. Table 5.1 shows that the power reactor transient data constitute the

vast majority of the transient worker exposure. The nonreactor licensees account for only 2%

of the transient workforce.

The following definitions apply to Table 5.1:

Form 5 Summation The summation of the TEDE from each of the Form 5s submitted for

the monitoring year. This is the summation of each dose record

grouped by licensee and individual. This distribution takes into

account multiple Form 5s for an individual at one NRC-licensed facility

but not multiple exposures at multiple licensees.

Transients - As This distribution represents the population of transient workers as they

Reported were reported by each licensee. This distribution is the subset of all

Form 5s where individuals were monitored at more than one licensee

during the monitoring year. This is the summation of dose records

grouped by individual and by licensee, so the distribution represents

how the transient worker population would appear within the total

distribution of all workers. This distribution takes into account multiple

Form 5s for an individual at one NRC-licensed facility but not multiple

exposures at multiple licensees.

Transients - Actual This is the actual distribution for transient workers summed per

individual. This represents the true number of individuals and places

each individual in the correct dose range. This distribution accounts

for multiple records per individual and multiple licensees.

Corrected Distribution This distribution represents the correction of the reported distribution

by subtracting the difference in the reported and actual distribution for

transient workers. This represents the most accurate dose distribution

for the licensee category and accounts for the multiple reporting of

individuals.

Table 5.1 illustrates the impact that the multiple reporting of these transient individuals had on

the summation of the exposure reports for 1997. Because each licensee reports the doses

received by workers while monitored by the particular licensee during the year, one would

expect that a summation of these reports would result in individuals being counted several

times in dose ranges lower than the range in which their total accumulated dose (the sum of the

personnel monitoring results incurred at each facility during the year) would actually place them.

Thus, while the total collective dose would remain the same, the number of workers, their dose

distribution, and average dose would be affected by this multiple reporting. This was found to
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be true because too few workers were reported in the higher dose ranges. For example, in

1997, Table 5.1 shows that the summation of annual reports for reactor licensees indicated that

62 individuals received doses greater than 2 rem. After accounting for those individuals who

were reported more than once, the corrected distribution indicated that there were really 343

workers who received doses greater than 2 rem. Correcting for the multiple counting of

individuals also has a significant effect on the average measurable dose for these workers. The

corrected average measurable dose for transient workers is twice as high as the value

calculated by the summation of licensee records. The transient workers represent 31% of the

workforce that receives measurable dose and increases the average measurable dose for all

licensees by 18% from 0.22 rem to 0.26 rem. It should be noted that this analysis of transient

workers does not include workers who may have been exposed at facilities that are not required

to report to the NRC REIRS database (see Section 1), Agreement State licensees, or

Department of Energy facilities.

One purpose of the REIRS database, which tracks occupational radiation exposures at NRClicensed facilities, is to identify individuals who may have exceeded the occupational radiation

exposure limits because of multiple exposures at different facilities throughout the year. The

REIRS database stores the radiation exposure information for an individual by their unique

identification number and identification type [Ref. 18, Section 1.5] and sums the exposure for all

facilities during the monitoring year. An individual exceeding the TEDE 5 rem per year

regulatory limit would be identified in Table 5.1 in one of the dose ranges >5 rem. In 1997, no

individual was discovered to have exceeded the limit as a result of the correction for transient

workers. Since 1985, there have been no additional transient workers identified as having

received a dose of >5 rem that have not appeared in the annual reports received by the

Commission. This reflects the industry’s continuing concerted efforts to keep the total annual

doses of all workers under 5 rem and shows that such reductions can be accomplished without

increasing the collective dose because the collective dose has decreased during this same time

period.
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6 EXPOSURES TO PERSONNEL IN EXCESS OF REGULATORY LIMITS

6.1 Control Levels

Exposures in excess of regulatory limits are sometimes referred to as “overexposures.” The

phrase “exposures in excess of regulatory limits” is preferred to “overexposures” because the

latter suggests that a worker has been subjected to an unacceptable biological risk, which may,

or may not, be the case.

The implementation date for the revised 10 CFR 20 was January 1, 1994. The separate limits

on internal and external exposure in the old 10 CFR 20 are no longer applicable. The revised

10 CFR 20 now includes requirements for summing internal and external dose equivalents to

yield TEDE and to implement a similar limitation system for organs and tissues (such as the

gonads, red bone marrow, bone surfaces, lung, thyroid, and breast). The dose equivalent limits

for the skin of the whole body and for the extremities have been revised, and a new limit for

dose equivalent to the lens of the eye has been added. The revised 10 CFR 20.1201 limits the

TEDE of workers to ionizing radiation from licensed material and other sources of radiation

within the licensee’s control. The revised 10 CFR 20 no longer contains quarterly exposure

limits but has reporting requirements for planned special exposures (PSEs)* . The annual TEDE

limit for adult workers is 5 rem.

The revised 10 CFR 20.2202 and 10 CFR 20.2203 require that all persons licensed by the NRC

submit reports of all occurrences involving personnel radiation exposures that exceed certain

control levels, thus providing for investigations and corrective actions as necessary. Based on

the magnitude of the exposure, the occurrence may be placed into one of three categories:

(1) Category A

10 CFR 20.2202(a)(1) - a TEDE to any individual of 25rem or more; an eye

dose equivalent of 75 rem or more; or a shallow-dose equivalent to the skin or

extremities of 250 rad or more. The Commission must be notified immediately of

these events.

(2) Category B

10 CFR 20.2202(b)(1) - a TEDE to any individual of 5rem or more; an eye dose

equivalent of 15 rem or more; or a shallow-dose equivalent to the skin or

extremities of 50 rem or more in a 24-hour period. The Commission must be

notified within 24 hours of these events.

* See 10 CFR 20.1206, 20.2204 and Regulatory Guide 8.35 for more information on PSEs and their reporting requirements.
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(3) Category C

10 CFR 20.2203 - In addition to the notification required by 20.2202 (category A

and B occurrences), each licensee must submit a written report within 30 days

after learning of any of the following occurrences: (1) Any incident for which

notification is required by 20.2202; or (2) Doses that exceed the limits in

20.1201, 20.1207, 20.1208, 20.1301 (for adults, minors, the embryo/fetus of a

declared pregnant worker, and the public, respectively), or any applicable limit in

the license; or (3) Levels of radiation or concentrations of radioactive material

that exceed any applicable license limit for restricted areas or that, for

unrestricted areas, are in excess of 10 times any applicable limit set forth in this

part or in the license (whether or not involving exposure of any individual in

excess of the limits in 20.1301); or (4) For licensees subject to the provisions of

the Environmental Protection Agency’s generally applicable environmental

radiation standards in 40 CFR 190, levels of radiation or releases of radioactive

material in excess of those standards, or of license conditions related to those

standards.

6.2 Limitations of the Data

It is important to note that this summary of events includes only:

• Occupational radiation exposures in excess of regulatory limits

• Events at NRC-licensed facilities

• Final dose of record assigned to an individual

It does not include:

• Medical misadministrations to medical patients

• Exposures in excess of regulatory limits to the general public

• Agreement State-licensed activities or Department of Energy facilities

• Other radiation-related violations, such as high dose rate areas or effluent limits

• Exposures to dosimeters that, upon evaluation, have been determined to be high

dosimeter readings only and are not assigned to an individual as the dose of record

by the NRC

Care should be taken when comparing the summary information presented here with other

reports and analyses published by the NRC or other agencies. Various reports may include

other types of “overexposure” events; therefore, the distinctions should be noted.
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The analysis and summary of incidents presented here involving exposures in excess of

regulatory limits represent the status of events as of the publication of this report. Exposure

events of this type typically undergo a long review and evaluation process by the licensee, the

NRC inspector for the regional office, and NRC headquarters. Preliminary dose estimates

submitted by licensees are often conservatively high and do not represent the final (record)

dose assigned for the event. It is therefore not uncommon for an “overexposure” event to be

reassessed and the final assigned dose to be categorized as not having been in excess of the

regulatory limits. In other cases, the exposure may not be identified until a later date, such as

during the next scheduled audit or inspection of the licensee’s exposure records.

For these reasons, an attempt is made to keep current the exposure events summary presented

here. An event that has been reassessed and determined not to be an exposure in excess of

the limits is not included in this report. In addition, events that occurred in prior years are added

to the summary in the appropriate year of occurrence. The reader should note that the

summary presented here represents a “snapshot” of the status of events as of the publication

date of this report. Previous or future reports may not correlate in the exact number of events

because of the review cycle and reassessment of the events.

6.3 Summary of Exposures in Excess of Regulatory Limits

Table 6.1 summarizes the occupational exposures in excess of regulatory limits as reported by

Commission licensees pursuant to 10 CFR 20.2202 and 10 CFR 20.2203 from 1994 to 1997.

Table 6.2 shows the data reported under 10 CFR 20.403 and 10 CFR 20.405 for the period

1985-1993. Note that the categorization criteria changed effective with the revised

10 CFR 20. The dose reporting thresholds have been revised — the skin of the whole body and

the extremities now have the same dose limits, and a new set of dose limits has been added for

the lens of the eye.

For the period 1990-1993, Table 6.2 shows the number of individuals who exceeded various

limits while employed by one of several types of licensees. For the period 1985-1989, only the

exposures in excess of regulatory limits reported by licensed industrial radiography firms are

shown separately. Most of the occurrences included in the “Others” category come from

research facilities, universities, and measuring and well-logging activities.

In 1997, two workers received doses that exceeded the regulatory limit for extremity dose.

There were no occurrences where an individual exceeded the regulatory limit for TEDE. One of

the exposures in excess of the extremity limit was a “Category A” occurrence, and was reported

immediately to the NRC upon discovery as required. There were no occurrences in which

individuals received a “Category B” exposure.
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The largest of the extremity exposures in excess of the regulatory limit occurred in May of 1997.

A “Type A” Broad radiopharmaceutical licensee reported that an employee went home without

properly frisking himself for radioactive contamination. When he returned to work the next day,

he performed a contamination survey of himself and detected significant levels of contamination

on his left thumb. The isotope was determined to be Re-186. The licensee believes that the

employee became contaminated while handling contaminated materials with a faulty glove. The

licensee reported that the individual received a shallow dose of 534 rem to the palm side of the

thumb. This dose is based on the licensee’s measurement and calculation of 4.3 uCi. Lower

levels of contamination were found on the back of his right hand and fingers. The root cause

was attributed to a failure to follow procedure.

The second extremity exposure in excess of the regulatory limit was reported by a reactor

license and occurred in October of 1997. The reactor licensee reported an extremity dose of

51.090 rem from a hot particle. The report stated that the dose was received from licensed

activities off-site. The NRC is following up on this report to determine the nature of this

exposure occurrence.

6.4 Maximum Exposures Below the NRC Limits

Because few exposures exceed the NRC occupational exposure limits, certain researchers

have expressed an interest in a listing of the maximum exposures received at NRC licensees

that do not exceed the limits. This would allow an examination of exposures that approach, but

do not exceed the limits. Table 6.3 shows the maximum exposures for each dose category

required to be reported to the NRC. In addition, the number of exposures in certain dose

ranges is shown to reflect the number of exposures that approach the NRC limits.

As can be seen from Table 6.3, few exposures exceed half of the NRC occupational annual

limits. The only doses to come within 5% of the limit were the two extremity exposures that

exceeded the limit.
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TABLE 6.3

MAXIMUM OCCUPATIONAL EXPOSURES FOR EACH EXPOSURE CATEGORY

1997

Exposure Annual Maximum Max Dose Number of Number of Number of Number of Number of

Category Dose Limit Exposure Percent of Individuals with Individuals Individuals Individuals Individuals

10CFR20* Reported the Limit Measurable ≥ 25% of ≥ 50% of ≥ 75% of ≥ 95% of

(rem) Dose the Limit the Limit the Limit the Limit

SDE-ME 50 rem 533.870 > limit 60,967 96 31 9 2 (>limit)

SDE-WB 50 rem 42.000 84% 75,008 5 2 1 0

LDE 15 rem 10.509 70% 74,323 12 1 0 0

CEDE 2.881 4,105

CDE 29.648 3,376

DDE 4.465 75,561

TEDE 5 rem 4.481 90% 77,094 2,238 280 12 0

TODE 50 rem 29.648 59% 62,984 99 4 0 0

* Shaded boxes represent dose categories that do not have specific dose limits defined in 10 CFR 20.