NUREG-0713

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.

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

 through Radiation Exposure 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.

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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

1995 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 1995 were received from a total of 295 NRC licensees, of which 109 were

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

submitted by the 295 licensees indicated that 143,684 individuals were monitored, 77,737 of

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

individuals was 24,884 person-cSv (person-rem)2

 which represents a <0.1% decrease from

the 1994 value. The number of workers receiving a measurable dose also decreased,

resulting in the average measurable dose of 0.32 cSv (rem) for 1995. 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 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was

199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual

collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem)

and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem).

Analyses of transient worker data indicate that 17,153 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 1995,

the average measurable dose calculated from reported data was 0.26 cSv (rem). The

corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem).

1 Commercial nuclear power reactors; industrial radiographers; fuel processors, fabricators, and reprocessores; 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.

2 In the International System of Units the sievert (Sv) is the name given to the units for dose equivalent. One centisievert

(cSv) equals one rem; therefore, person-rem becomes person-cSv.

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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

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TABLE OF CONTENTS

Page

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

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

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

FOREWORD .................................................................................................................................x

PREFACE .................................................................................................................................... xi

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 Manufacturer 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 Fabrication and Reprocessing Licenses................................................ 3-15

3.3.6 Light-Water-Cooled Power Reactor (LWR) Licenses .................................... 3-19

3.3.7 High-Temperature Gas-Cooled Power Reactor (HTGR) Licenses ................ 3-19

3.4 Summary of Intake Data by License Category ...................................................... 3-21

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NUREG-0713

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 .............................................................................. 4-1

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..................................... 4-5

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

4.4 Average Annual TEDE Doses ................................................................................. 4-6

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

4.6 Collective Dose by Work Function and Employee Type ........................................ 4-23

4.7 Number of Personnel by Work Function and Employee Type ............................... 4-30

4.8 Graphical Representation of Dose Trends in Appendix E ..................................... 4-34

4.9 Health Implications of Average Annual Doses ....................................................... 4-34

4.10 Estimation of Future Occupational Radiation Exposure at

Commercial Reactor Sites ..................................................................................... 4-36

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

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

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

5.3 Career Dose Analysis .............................................................................................. 5-4

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

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

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

LICENSEES IN DESCENDING ORDER OF AVERAGE MEASURABLE DOSE,

1995................................................................................................................. A-1

APPENDIX B - ANNUAL WHOLE BODY DOSES AT LICENSED NUCLEAR POWER

FACILITIES, 1995............................................................................................ B-1

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TABLE OF CONTENTS (Continued)

Page

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

1969-1995 .......................................................................................................C-1

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

AND JOB FUNCTION, 1995 ...........................................................................D-1

APPENDIX E - GRAPHICAL REPRESENTATION OF COLLECTIVE DOSE

TRENDS BY YEAR AND JOB FUNCTION FOR EACH SITE,

1973-1995 ....................................................................................................... E-1

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

AND REACTOR TYPE, 1987-1995 ................................................................. F-1

LIST OF TABLES

Table 3.1 Annual Exposure Data for Certain Categories of

Licensees 1986-1995 .................................................................................... 3-4

Table 3.2 Distribution of Annual Collective TEDE by License

Category 1995 ............................................................................................... 3-6

Table 3.3 Summary of Annual Dose Distributions for Certain NRC

Licensees 1968-1995 .................................................................................... 3-7

Table 3.4 Annual Exposure Information for Industrial

Radiographers 1993-1995 ............................................................................. 3-8

Table 3.5 Annual Exposure Information for Manufacturers and

Distributors 1993-1995 ................................................................................ 3-11

Table 3.6 Annual Exposure Information for Fuel Fabricators

1993-1995 ................................................................................................... 3-16

Table 3.7 Annual Exposure Information for Fort St. Vrain

1974-1995 ................................................................................................... 3-20

Table 3.8 Intakes by Licensee Type and Radionuclide

Mode of Intake - Ingestion 1995 ................................................................. 3-22

Table 3.9 Intakes by Licensee Type and Radionuclide

Mode of Intake - Inhalation 1995 ................................................................ 3-23

Table 4.1 Summary of Information Reported by Commercial

Boiling Water Reactors 1973-1995 ................................................................ 4-2

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LIST OF TABLES (Continued) Page

Table 4.2 Summary of Information Reported by Commercial

Pressurized Water Reactors 1973-1995........................................................ 4-3

Table 4.3 Summary of Information Reported by Commercial

Light Water Reactors 1973-1995 ................................................................... 4-4

Table 4.4 Summary Distribution of Annual Whole Body Doses at

Commercial Light Water Reactors 1977-1995............................................... 4-7

Table 4.5 Boiling Water Reactors Listed in Ascending Order of

Collective Dose per Reactor 1991-1995...................................................... 4-14

Table 4.6 Pressurized Water Reactors Listed in Ascending Order

of Collective Dose per Reactor 1991-1995 .................................................. 4-15

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

Order of Collective Dose per BWR 1991-1995............................................ 4-17

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

Order of Collective Dose per PWR 1991-1995............................................ 4-18

Table 4.8a Activities Contributing to High Collective Doses at Selected

 Plants in 1995 - BWRs with High Collective Doses ....................................... 4-19

Table 4.8b Activities Contributing to High Collective Doses at Selected

 Plants in 1995 - PWRs with High Collective Doses ...................................... 4-21

Table 4.9 Annual Collective Dose by Work Function and Personnel Type 1995............. 4-25

Table 4.10 Percentages of Annual Collective Dose at LWRs by Work

Function 1984-1995..................................................................................... 4-26

Table 4.11 Annual Collective Dose by Occupation and Personnel Type 1995 ................. 4-29

Table 4.12 Number of Personnel by Work Function and Personnel Type 1995 ................ 4-31

Table 4.13 Number of Personnel by Occupation and Personnel Type 1995..................... 4-32

Table 4.14 Average Doses by Occupation and Personnel Type 1995 .............................. 4-33

Table 4.15 Parameters Used in Collective Dose vs. Plant Age Data Model ..................... 4-38

Table 4.16 Projected Collective Dose, 1996 - 2000 .......................................................... 4-40

Table 5.1 Effects of Transient Workers on Annual Statistical Compilations, 1995 ............ 5-2

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LIST OF TABLES (Continued) Page

Table 6.1 Occupational Exposures in Excess of Regulatory Limits, 1994-1995 ............... 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 1995 ............ 6-7

LIST OF FIGURES

Figure 3.1 Average Annual Values at Industrial Radiography Facilities

1973-1995 ................................................................................................... 3-10

Figure 3.2 Average Annual Values at Manufacturing and Distribution

Facilities 1973-1995 .................................................................................... 3-12

Figure 3.3 Average Annual Values at Low-Level Waste Disposal

Facilities 1982-1995 .................................................................................... 3-14

Figure 3.4 Average Annual Values at Independent Spent Fuel Storage

Facilities 1982-1995 .................................................................................... 3-17

Figure 3.5 Average Annual Values at Fuel Fabrication and Processing

Facilities 1973-1995 .................................................................................... 3-18

Figure 4.1 Average Collective Dose and Number of Workers per

Reactor 1973-1995 ........................................................................................ 4-8

Figure 4.2 Number of Operating Reactors and Gross Electricity

Generated 1973-1995 ................................................................................... 4-9

Figure 4.3 Average Measurable Dose per Worker and Collective Dose per

Megawatt-Year 1973-1995 .......................................................................... 4-10

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

Dose per Reactor 1973-1995 ...................................................................... 4-12

Figure 4.5 Outage Days, Average Dose, and Collective Dose ......................................... 4-24

Figure 4.6 Collective Dose by Work Function and Personnel Type

1990-1995 ................................................................................................... 4-28

Figure 4.7 Average Collective Dose by Site Age .............................................................. 4-37

Figure 4.8 Reactor Collective Dose Projections ............................................................... 4-39

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FOREWORD goes here.

NRC supplies this text. Intentionally left blank.

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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, or who may exceed regulatory limits on radiation

exposure due to the accumulation of exposure at multiple sites per calendar quarter or

calendar year.

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 public interest groups, special interest groups, labor

unions, etc.

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

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NUREG-0713

Occupational Radiation Exposure

at Commercial Nuclear Power Reactors and Other Facilities

Twenty-eighth Annual Report, 1995

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

categories1

 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

3 Commercial nuclear power reactors; industrial radiographers; fuel processors, fabricators, and reprocessors; manufacturers

and distributors of specified quantities of byproduct material.

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NUREG-0713

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

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 second

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.

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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

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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 cSv (rem), so 0.5 cSv

(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 for

identification or convenience, although the probability of their being exposed to measurable

levels of radiation is extremely small. Licensees are given the option of reporting the doses of

only those individuals for whom monitoring is required, or the dose distribution of all those for

whom monitoring is provided. Many licensees elect to report the latter; however, this may

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.

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. In previous reports,

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

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report, 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 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 and 1995 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 pay

attention to the differences between all power reactors [including the high temperature gas

reactor (HTGR), all pressurized water reactors (PWRs), and all 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.

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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,

fabricators and reprocessors, 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 by-product material, and low-level waste disposal. Companies that conduct these

types of activities in Agreement States1

 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. Information shown for these categories does not reflect the total U.S. experience.

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 12/31/94, there are 29 Agreement States.

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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 categories1

 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 new 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 10 CFR 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.

5 Commercial nuclear power reactors; industrial radiographers; fuel processors, fabricators, and reprocessores; manufacturers

and distributors of byproduct material; independent spent fuel storage installations; and facilities for land disposal of low-level

radioactive waste.

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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 Doses

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

individual with a TEDE greater than zero cSv (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-cSv (person-rem).2

 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 the collective dose in past years 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 1995 with the collective dose for previous years 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 1995 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.).

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6 In the International System of Units, the sievert (Sv) is the name given to the units for dose equivalent. One centisievert (cSv)

equals one rem; therefore person-rem becomes person-cSv.
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 cSv 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 cSv (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 cSv 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 cSv in the 1 to 2 cSv range was calculated by assuming that half of the

collective dose incurred by workers with doses between 1 and 2 cSv was because of doses

greater than 1.5 cSv. 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

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indicator when consistently applied to the data from year to year.

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 cSv

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 0.06 in 1995. Using the previous methodology, the CR value would have been

calculated to be 0.23 in 1994 and 0.19 for 1995. One of the contributing factors for this

difference is the administrative controls imposed at nuclear power facilities for individuals who

exceed 1 cSv. This causes the dose distribution to drop off sharply above 1 cSv with fewer

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

is calculated by assuming a uniform dose distribution.

Other licensees, such as Manufacturing and Distribution and Independent Spent Fuel Storage,

have experienced increases 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 and 1995. However, the overall

average CR for all licensees remained below 0.50, and decreased to a value of 0.10 in 1995

primarily because of the decrease in CR at power reactor licensees.

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 5 cSv (rem). 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 cSv

(rem) for occupationally exposed adults is, by definition, exposures 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 10 CFR 20.2206. Table 3.3 shows that ~ 95% of

the exposures consistently remained <2 cSv (rem) between 1968 and 1984. For the past 10

years the percentage of workers with <2 cSv (rem) has been ≥98%. The number of workers

receiving an annual exposure in excess of 5 cSv (rem) has been <0.01% since 1985.

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TABLE 3.3

SUMMARY OF ANNUAL DOSE DISTRIBUTIONS FOR CERTAIN NRC LICENSEES

1968-1995

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 99.3% >99.99% (1) 0

* Data for 1977-1995 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

cSv is shown in parentheses from 1985-1995.
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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

139 radiography licensees in 1995. Table 3.4 summarizes the reported data for the two types

of radiography licenses for 1995 and for the previous 2 years for comparison purposes.

For the years prior to 1994, the average measurable dose for workers performing radiography

at a single location ranged from 20 to 40% 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. In 1994, the

average measurable dose for single location radiographers was much closer to the value for

multiple location licensees because of high average doses at one licensee, Buckeye Steel

Castings. For 1995, the average measurable dose for single location licensees

 TABLE 3.4

ANNUAL EXPOSURE INFORMATION FOR INDUSTRIAL RADIOGRAPHERS

1993 - 1995

 Year Type of License Number of Number of Workers Collective Average

Licenses Monitored with Dose Measurable

Workers Measurable (person- Dose (cSv

Dose cSv, rem) or rem)

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

Single Location 29 330 89 44 0.50

1994 Multiple Locations 111 2,900 2,262 1,371 0.61

Total 139 3,230 2,351 1,415 0.60

Single Location 39 673 183 23 0.13

1993 Multiple Locations 137 4,046 2,824 1,572 0.56

Total 176 4,721 3,007 1,596 0.53

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is back down to ~ 20% of the average dose for multi-location licensees. 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 1995 is presented in Appendix A in

descending order of average measurable dose.

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 1995.

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

Manufacturer and Distributor 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 1995

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

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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.

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.

To see the contribution that each of these licensees made toward the total values of the

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

values of these parameters for each licensee in descending order of average measurable dose

for 1995.

 TABLE 3.5

ANNUAL EXPOSURE INFORMATION FOR MANUFACTURERS AND DISTRIBUTORS

1993 - 1995

 Year Type of License Number of Number of Workers Collective Average

Licenses Monitored with Dose Measurable

Workers Measurable (person- Dose (cSv

Dose cSv, rem) or rem)

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

M & D-”A”-Broad 8 2,133 877 544 0.62

1994 M & D-Limited 36 808 374 36 0.10

Total 44 2,941 1,251 580 0.46

M & D-”A”-Broad 8 2,455 925 512 0.55

1993 M & D-Limited 50 2,458 1,329 168 0.13

Total 58 4,913 2,254 680 0.30

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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 that have primary regulatory

authority over its activity. However, they 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; however, there

have been two licensees in this category since 1984. Table 3.1 summarizes the data reported

for 1984 through 1995. Appendix A summarizes the exposure information reported by these

two licensees in 1995.

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 1995. Because only two licensees have been involved in

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

1995.

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 1995. Appendix

A summarizes the exposure information reported by this installation.

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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

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

1995, rather than the two licensees that reported in prior years. The average measurable dose

parameter is not based on the number of licensees and has also experienced a significant

increase since 1993.

3.3.5 Fuel Fabrication and Processing Licenses

The Fuel Fabrication and Processing licenses are issued to allow the processing and

fabrication of reactor fuels. In most uranium facilities where light water reactor fuels are

processed, uranium hexafluoride enriched in the isotope U-235 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. On a much

smaller scale, fuel assemblies containing plutonium oxide pellets can be similarly fabricated

and used in reactors for experimental purposes. However, there are no NRC licensees

engaged in this activity at this time.

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 Fabrication and

Processing 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.

Prior to 1994, only the “whole body” dose values were given, which were equivalent to the

DDE. In 1994, the revised 10 CFR 20 went into effect, requiring the calculation of the CEDE

and the summation of the DDE and CEDE into the TEDE. For Fuel Fabrication facilities, the

CEDE is a significant contribution to the TEDE. To accurately reflect the exposure history for

these facilities, it was necessary to continue to plot the old “whole body” external dose, now

called DDE, in addition to the TEDE, which includes the CEDE contribution. The difference

between the DDE and TEDE plots represents the CEDE contribution.

Appendix A lists each of the licensees reporting in 1995, with the number of workers

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

dose for each licensee in descending order of average measurable dose.

Table 3.6 shows that there were eight licensed Fuel Fabrication facilities in 1995. Several

licensees were involved in decontamination and decommissioning of their plutonium facilities,

and for several years the data for these licensees were shown in the “Decommissioning”

category in Table 3.1. Because these facilities have ceased to fabricate plutonium fuel, they

are not required to file annual reports and are no longer shown in the tables.

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Fuel Reprocessing licenses are issued to allow the separation of useable uranium and

plutonium from spent nuclear fuel. There was only one commercial facility that was ever

licensed to reprocess fuel, and it has been shut down since 1972. However, the licensee did

some decontamination work and stored radioactive waste at the facility for several years, and

the annual report that was submitted each year was usually grouped with those of the Fuel

Fabricators. In February 1982, the Department of Energy assumed possession and control of

the reprocessing facility to conduct waste solidification activities necessary for final

decommissioning. Therefore, since 1982 the NRC license has been suspended, and no reports

have been filed with the NRC.

 TABLE 3.6

ANNUAL EXPOSURE INFORMATION FOR FUEL FABRICATORS

1993 - 1995

Year Type of License Number Number of Workers Collective Average Collective Average

of Monitored with TEDE Measurable CEDE CEDE

Licenses Workers Measurable (person- Dose (cSv (person-cSv, (cSv or rem)

Dose cSv, rem) or rem) rem)

1995 Uranium Fuel Fab 8 4,106 2,959 1,217 0.41 990 0.33

1994 Uranium Fuel Fab 8 3,596 2,847 1,147 0.40 867 0.30

1993 Uranium Fuel Fab 8 9,649 2,611 339 0.13 NA NA

 NA - Not applicable prior to the revised 10 CFR20 implementation in 1994.

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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 1986

through 1995. This table includes reactors that may not have been in commercial operation for

a full year. Data for 1986 through 1988 included all reactors that reported, even though some

of them were shut down. Data for 1989 through 1995 do not include reactors that have been

shut down. These figures have been adjusted for the multiple counting of transient workers

(see Section 5). The reported dose distribution of workers monitored at each plant site is

presented in alphabetical order by site name in Appendix B.

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

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

3.3.7 High-Temperature Gas-Cooled Power Reactor (HTGR) Licenses

A license to operate a power reactor is issued to utilities to allow them to use special nuclear

material in a reactor to produce heat to generate electricity to be sold to consumers. In the

HTGR, a gas, usually helium, is used as the primary coolant. Fort St. Vrain, near Greeley,

Colorado, was the only such reactor in operation in the United States. Fort St. Vrain shut

down permanently in 1989. Table 3.7 shows the annual whole body doses incurred by workers

at the plant. Since 1992, the doses have increased significantly because of decontamination

and decommissioning operations.

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TABLE 3.7

ANNUAL EXPOSURE INFORMATION FOR FORT ST. VRAIN

1974 - 1995

No Meas’ble Number of Dose Electricity Measurable

Meas’ble Dose 0.10 - 0.25 - Monitored (person-cSv Generated Dose (cSv or

Year Dose <0.10 0.25 2.00 >2.0 Workers person-rem) (MW-yr) rem)

1974 1,597 63 1 0 0 1,661 3.3 0.0 0.05

1975 1,263 0 0 0 0 1,263 0.0 0.0 0.00

1976 1,362 25 0 0 0 1,387 1.3 2.8 0.05

1977 946 55 1 0 0 1,002 2.9 29.8 0.05

1978 896 34 0 0 0 930 1.7 75.7 0.05

1979 1,149 120 2 0 0 1,271 6.4 28.6 0.05

1980 902 57 1 0 0 960 3.0 83.2 0.05

1981 1,096 31 0 0 0 1,127 1.0 93.6 0.03

1982 978 22 0 0 0 1,000 0.4 72.6 0.02

1983 965 48 0 0 0 1,013 1.0 94.4 0.02

1984 1,616 62 8 0 0 1,686 3.0 10.9 0.04

1985 1,929 370 40 33 0 2,372 35.0 3.8 0.08

1986 221 66 4 0 0 291 1.8 9.7 0.03

1987 155 52 2 0 0 209 1.2 23.8 0.02

1988 238 24 0 0 0 262 0.7 81.8 0.03

1989 316 47 6 2 0 371 2.7 0.0 0.05

1990 226 30 0 0 0 256 0.6 0.0 0.02

1991 525 63 9 4 0 601 5.4 0.0 0.07

1992 520 144 36 34 0 734 25.4 0.0 0.12

1993 657 51 37 78 1 823 75.2 0.0 0.45

1994 390 89 33 79 4 591 78.0 0.0 0.39

1995 460 62 52 127 37 738 210.3 0.0 0.75

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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 3.8 and 3.9 summarize the intake data reported to the NRC during 1995. The data are

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

class. Table 3.8 lists the intakes where the mode of intake into the body was recorded as

ingestion. Table 3.9 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.

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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 has been shut down since 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.

4.2.2 Electric Energy Generated

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

shown in Appendix C and graphically represented in Appendix E. This number was obtained

by dividing the gross 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. The

gross 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 gross megawatt-hours of electricity produced each year was found in Reference 14.

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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-cSv (person-rem) by the gross electric energy generated in megawatt-years and is a

measure of the dose incurred by workers at power plants in relation to the gross electric

energy 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.

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 82% in 1995, a gain of 20% in 8 years.

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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 1995. This distribution is the sum of

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

mentioned, the distribution reported by each LWR site for 1995 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

<1% to a value of 21,674 person-cSv (person-rem) in 1995. The value of CR decreased to a

value of 0.06. The large decrease from 1993 to 1994 is primarily because of the change in

methodology by which the CR value is determined (see Section 3.1.8) . In 1994 and 1995, 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 eleventh consecutive year that the

value of CR has been <0.50.

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 1995, the average collective dose per reactor dropped by 74%. In 1995, the collective

dose per reactor for PWRs increased by 28% to 170 person-cSv (person-rem). The collective

dose per reactor for BWRs decreased by 22% from 327 person-cSv (person-rem) in 1994, to

256 person-cSv (person-rem) in 1995. The overall collective dose per reactor for LWRs

remained the same at 199 person-cSv (person-rem) in 1995. The number of workers with

measurable dose per reactor has decreased to 964 for BWRs but increased to 720 for PWRs

in 1995. 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 <1% from a value

of 21,695 person-cSv (person-rem) in 1994 to 21,674 person-cSv (person-rem) in 1995.

Together with the increase in the number of workers with measurable dose, this resulted in the

average measurable dose per worker decreasing to 0.25 cSv (rem) in 1995. Figure 4.2 shows

that in 1995 the gross electricity generated increased to an all-time high of 80,562 MW-yr.

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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, Figure 4.4 displays the

average and median1

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

years 1973 through 1995. 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

experienced an increase from 135 person-cSv (person-rem) in 1994 to 146 person-cSv

(person-rem) in 1995. At BWRs, the median fluctuates more from year to year, and in 1995

the median collective dose decreased to 244 person-cSv (person-rem). Figure 4.4 also shows

that, in 1995, 50% of the PWRs reported collective doses between 102 and 207 person-cSv

(person-rem) while 50% of the BWRs reported collective doses between 136 and 357

person-cSv (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).

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9 The value at which 50% of the reactors reported greater collective doses and the other 50% reported smaller collective doses.
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4.5 Plant Rankings by Collective Dose per Reactor

Because the number of reactors from which data have been collected is still 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 1991 through 1995. The total

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

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 cSv

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

plants, and in 1995, the CR for all the U.S. LWRs fell between 0.05 and 0.50, the range

recommended by the UNSCEAR [Ref. 10]. Note that in 1994 and 1995, 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).

In 1995, the five BWR sites with the highest collective doses all exceeded 379 person-cSv

(person-rem) per reactor (Table 4.5). These reactors were Nine Mile Point 1 and 2, Dresden 2

and 3, Washington Nuclear 2, Pilgrim, and Millstone Point 1. Although the seven reactors at

these five sites represented only 19% of the 37 BWRs, they contributed 34% of the total

collective dose incurred at BWRs in 1995.

Some of the activities that contributed to the collective dose accumulated at the BWR site with

the highest collective dose per reactor [Millstone Point 1 with 620 person-cSv (person-rem)]

were weld repair, in-service inspection, hanger work, insulation removal and replacement,

staging work, and refueling activities.

In 1995, the five PWR sites with the highest collective doses all exceeded 398 person-cSv

(person-rem) per reactor (Table 4.6). These reactors were Zion 1 and 2, Haddam Neck,

Palisades, Indian Point 2, and Maine Yankee. Although representing 8% of the 72 PWRs

included in 1995, they contributed 24% of the total collective dose at PWRs. Much of the

collective dose accumulated at the plant with the highest dose per reactor in 1995 [Maine

Yankee with 653 person-cSv (person-rem)] was attributed to steam generator related work

(including tube sleeving, eddy current testing, and sludge lancing), reactor coolant pump work,

outage support, valve work, decontamination, refueling activities, and in-service inspection.

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

December 31, 1995, 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 319 person-cSv (person-rem), the average measurable dose per worker was 0.30 cSv

(rem), and the average collective dose per megawatt-year was 0.5.

Based on the 353 reactor-years of operation at the 71 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 190 person-cSv (person-rem), 0.25 cSv (rem), and

0.3 person-cSv/MW-yr, respectively. All of these values, at both types of facilities, are lower

than those found for the 5 year period ending in 1994, with the exception of the average

collective dose per site and average collective dose per megawatt-year at PWRs, which

remained the same.

In some cases, the plants having the lower values for most of the parameters shown in Tables

4.7a&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.

Usually, the combination of a large annual collective dose and a large collective dose to

megawatt-year ratio for a plant indicates that extensive maintenance or modifications were

undertaken during the year. Jobs that were large contributors to BWR doses in 1995 included

in-service inspections, valve maintenance work, refueling activities, shielding installation and

removal, and area and system decontamination. At PWR facilities, the major contributors to

the collective dose were steam generator related work, valve maintenance work, refueling

activities, scaffolding and insulation, in-service inspections, health physics coverage, and

reactor coolant pump maintenance.

A complete breakdown of the activities contributing to the collective dose at the ten sites with

the highest dose per reactor ranking in 1995 (from Tables 4.5 and 4.6) is given in Tables 4.8a

and 4.8b for BWRs and PWRs respectively. The outage dose and duration are shown as well

as the collective dose for each activity.

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Even with the use of better techniques and robotics, these tasks continue to be responsible for

a major percentage of the collective dose. It should be noted that the differences in nuclear

plant designs and the ages of the plants, even between plants of a given type, affect the

nature of these parameters [Ref. 15]. Therefore, care should be exercised when attempting to

draw conclusions from these data.

From the above analysis, one can see that 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 statistical report 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 4.9 through 4.14 summarize the 1995 data for BWRs, PWRs,

and LWRs. Table 4.9 shows that, at both BWRs and PWRs, about 62% 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 4.10 shows that workers performing special maintenance prior to 1987 incurred the

largest portion (35%-45%) of the collective dose and that workers performing routine

maintenance activities usually incurred between 25% and 35% of the total. For the past 9

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

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and routine maintenance, always accounting for the majority of the collective dose. Some of

the fluctuations shown inthe 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 dose by work function and type of

personnel for the years 1990 through 1995 for BWRs and PWRs separately. The general

decrease in collective dose is also apparent among most of these activities.

Table 4.11 presents the distribution of the collective dose for 1995 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 1994. The collective doses shown in Tables 4.9 and 4.11

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 obtained or calculated from the annual reports that had been required to be submitted

pursuant to 10 CFR 20.2206.

Another use made of the reports given in Appendix D is in proportioning the collective dose

obtained from the § 20.407 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.

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(4) A similar procedure was followed in determining the collective dose for the columns

headed “Contractor” and “Station & Utility” in Appendix C.

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 4.12 and 4.13 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 55% of the personnel performed

routine or special maintenance functions, 26% were involved with reactor operations and

surveillance, and the remaining 19% were divided among the other three work functions.

Table 4.13 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 differences occurred in the maintenance and supervisory percentages for 1995.

Overall, 56% of the personnel were contractors, 36% were station employees, and 8% were

utility employees in 1995.

Table 4.14 presents the average annual dose incurred by workers in the five occupational

categories in 1995. 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 4.13. It shows

that, in most instances, the maintenance and health physics personnel incur the highest

average doses. Examination of the values of the averages given in Table 4.14 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 4.14; (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 cSv (rem) while other plants count all workers regardless of the

dose received. Because of these mitigating 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 Appendix E presents two types of graphs for one site. One graph plots selected

dose-performance indicators from 1973 through 1995, and the other indicates the collective

dose by job function for 1978 through 1995. The dose and performance indicators shown in

the top graph illustrate the history of the collective dose for the site, the rolling 3-year average

collective dose per reactor, and the gross electricity generated at the site. These data are

plotted, beginning with the plant’s first full year of commercial operation, and continuing

through 1995. 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. Data for years when the plant was not in commercial

operation have been included when available. This 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. For sites with more than one reactor, the plot of the 3-year rolling average will lie below

that of the plot of the annual collective dose for the site because it is calculated on a

per-reactor basis.

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 1995. 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 among

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.

4.9 Health Implications of Average Annual Doses

Studies of populations chronically exposed to low levels of radiation delivered over protracted

periods have not shown consistent or conclusive evidence of an associated increase in the risk

of cancer. Thus, there is no evidence that the doses to workers recorded here cause harm.

The risk estimates presented below are based on extensive studies of Japanese Atomic bomb

survivors and other populations exposed to large doses of radiation delivered in short periods

of time. This information is supplemented by animal and in vitro studies, such as irradiation of

cell cultures. These studies have confirmed that human cells have mechanisms that repair

damaged chromosomes. The existence of this repair helps to explain the finding that lower

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doses of radiation delivered at lower dose rates produce less of an effect on a cell per unit

dose than high-dose, high-dose-rate irradiations. Thus the estimates of risks to radiation

workers are likely to be conservative.

Health effects due to radiation exposure fall into three groups: carcinogenic effects, genetic

effects, and mental retardation. Mental retardation has been observed only in Japanese

A-bomb survivors exposed at 8-15 weeks gestational age, and is consequently not applicable

to the workplace except in the case of a pregnant female worker. Genetic effects have never

been observed in man, though they have been observed in mice.

Risk of cancer induction is known to increase with increasing dose, but is hard to quantify as

the risk varies with the site of the cancer, the age and sex of the exposed individual, the

energy and nature of the radiation, the magnitude and duration of the dose, and exposure to

other carcinogens. Since nearly 20% of all deaths in the United States occur from cancer, the

estimated number of cancers attributable to occupational radiation exposure is a small fraction

of the total number that occur. (Those who do not succumb to cancer will, perforce, succumb

to some other cause and in essentially the same time frame.)

The Committee on the Biological Effects of Ionizing Radiations (BEIR) of the National

Academy of Sciences (NAS) National Research Council has been conducting an ongoing

study of the health effects of ionizing radiation. Its latest report, BEIR V, was published in

1990. Based on this report, the 76,822 workers receiving the average dose of 0.32 cSv (rem)

continuously during an entire working career (working from age 18 until age 65) or the

maximum accidental dose of 5.1 cSv (rem) to the whole body during 1995 (see Section 6)

might expect an increased cancer death risk of about 9 chances in 1000 for the average dose

and 4 chances per 1000 for the maximum dose.2 Should a worker receive 0.32 cSv (rem)

continuously during an entire working career (working from age 18 until age 65), his/her

lifetime risk of dying from cancer is estimated to increase by approximately 4%. Since the

American Cancer Society estimates that an individual’s risk of dying of cancer is about 20%

(one in five), the risk to an individual receiving 0.32 cSv (rem) would be approximately 21%.

The potential genetic effects from a worker population receiving 24,536 person-cSv

(person-rem) (Table 3.1) are small compared to genetic damages that normally occur

spontaneously in a population of this size. Approximately 100,000 serious genetic defects

occur normally in one million live births, i.e., an average of about one serious defect in every

ten live births. Theoretically, the total genetic damage in the first generation children of the

76,822 exposed workers would, according to NUREG/CR-4214 [Ref. 17], be an increase of

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10 These estimates were calculated from Table 4-2 of Ref. 16. The average dose risk estimate assumes continuous lifetime

exposure (ages 18-65), while the acute dose risk estimate assumes a one-time, instantaneous exposure. Note that these

estimates are based on observations of individuals exposed to high doses of radiation over short periods of time. The BEIR

committee, in its report, cautions that dose rate reduction factors (DREFs) will need to be applied to low-dose and low-dose-rate

exposures. (see Ref. 16, pp. 171 and 174)
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about 8 cases (approximately 0.01%) compared to the expected 8,000 cases that occur

normally.3

 No significant increase in the number of genetic defects has been observed in the

children of individuals exposed to much higher levels of ionizing radiation at Hiroshima and

Nagasaki, Japan.

4.10 Estimation of Future Occupational Radiation Exposure at Commercial Reactor Sites

Data on occupational exposure from 1973 to 1995 suggest that commercial power reactor

sites show a consistent life-cycle exposure pattern, as illustrated in Figure 4.7. The horizontal

axis shows the average age of reactors at a site in years, while the vertical axis shows the

average yearly collective dose per site in cSv (rem). The general shape of the curve supports

the hypothesis that exposure increases during the startup and “shakedown” phase of

operation, and then gradually decreases as operations become more routine and sources of

exposures are identified and remediated. While BWR and PWR reactors show the same

general pattern, the average exposure levels at PWR reactors are lower until well into the

second decade of operation.

A regression model that captures this life-cycle pattern was developed based on exposures at

U.S. power plants from 1973 to 1995. The model uses information on average site age and

other factors, such as type of reactor, site capacity, and amount of power generated in a year,

that can influence worker exposure. Only reactors completing a full year of commercial power

operation are included. Dose information for reactors that began operation prior to 1973 are

not included, so the initial years of operation for these reactors are not included in the model or

reflected on the graphs. In addition, only those sites where the reactor unit age difference is

<5 years are included. Because the average refueling cycle is 18-24 months, the model uses

a 3-year exposure total to minimize the effect of the year-to-year differences that can occur

within that cycle. The analysis summarizes dose and reactor information by site, because

exposure data per reactor unit are not available. Data that allow separate calculations for

each reactor at a site would increase the model’s accuracy. The model estimates the

collective dose in cSv (rem) at each site based on the parameters shown in Table 4.15.

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11 Assuming that, on the average, each exposed person will have one live-born child in the future, i.e., 76,822 children born to this

worker population. The estimates were calculated from Table 4.1 of reference 17.
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Table 4.15 Parameters Used in Collective Dose

vs. Plant Age Data Model

Parameter Description

Site Age Average age of reactor units at the site in years.

Only includes sites where reactor unit age

differences are < 5 years and only includes data

from 1973 to 1995.

Capacity Total capacity in megawatts

MW Years A measure of amount of power generated during the year

Reactor Type PWR, BWR4

Dose Year 1 Total dose 1 year ago

Dose Year 2 Total dose 2 years ago

Dose Year 3 Total dose 3 years ago

RX Size 1 if average reactor size at site is ≥ 1000 MW;

0 if less than 1000 MW

Site Size 1 if capacity is ≥ 1000 MW or there is more than

1 reactor at the site;

0 if the capacity is less than 1000 MW

Because exposure levels were impacted significantly in the wake of the TMI incident, a single

model will not fit the data before and after this incident. Most of the post-TMI mandated plant

modifications were completed from 1980 to 1985. Collective exposure per site dropped from

860 cSv (rem) for 1973-1985 to 473 cSv (rem) in 1986-1995; exposure per megawatt rated

capacity dropped from 1.1 cSv/MS (rem/MW) to 0.5 cSv/MW (rem/MW) between the two

periods. The model included here uses all the available data, and provides the best fit for the

post-TMI period, to provide the most accurate projections for future years.

The model generates year-by-year estimates of expected dose that can be aggregated to

estimate total U.S. worker exposure for a given year. This allows predictions to be made for

the United States as a whole, while taking into account the varying ages and histories of

reactors at each site. Figure 4.8 compares the actual versus projected aggregate U.S.

exposure levels for 1985-19955

, and shows projections through 2005. The projections (also

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12 Only one site had both PWR and BWR reactors (Millstone Point), and it was classified as a PWR site. The single site does not

provide enough data to test whether its exposures are significantly different from a PWR-only site. The single HTGR reactor did

not provide enough data to be included in the analysis.

13 The projections through 1993 have been adjusted by using actual dose for the first 3 years of a site’s operation, because a site

must have at least 3 years of operating history before the next year’s dosage can be estimated. For 1994 and later years, no

adjustment is needed.
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shown in Table 4.16) take into account all of the factors listed in Table 4.15, the aging of the

reactor population, and the scheduled closing of Big Rock Point in 2000. The actual dates of

future reactor shutdowns are unknown and may be affected by petitions for plant license

extensions. The model does not take into account any exposure associated with

decommissioning after these reactors cease commercial power operations. Because the exact

amount of power generated and actual future dosages are unknown, the most recent 3-year

averages were used as values for megawatt-years and Dose Year 1 through Dose Year 3.

The results are best used to identify expected trends, rather than predicting the actual

exposure in any single year.

Table 4.16 Projected Collective

Dose, 1996 - 2000

 Year Projected Collective

Dose (TEDE)

 person-cSv (-rem)

1996 20,553

1997 19,620

1998 19,531

1999 19,401

2000 19,056

From this analysis, it is anticipated that the total collective dose at reactor sites will continue to

decrease over the next several years. Other factors, such as extended unanticipated outages

or shutdowns, may have a significant impact on future doses. The projections are an

estimation of the general trend over the next 5 years. Any given year may have a collective

dose above or below these estimated values.

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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 contribute only an

additional 0.5% of the transient workforce and an additional 0.4% to the collective dose.

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

Reported they 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 staff’s summation of the exposure reports for 1995. 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 be true because too few workers were reported in the higher dose ranges.

For example, in 1995, Table 5.1 shows that the summation of annual reports for reactor

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licensees indicated that 130 individuals received doses greater than 2 cSv (rem). After

accounting for those individuals who were reported more than once, the corrected distribution

indicated that there were really 713 workers who received doses greater than 2 cSv (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 22% of the workforce that receives measurable dose and

increases the average measurable dose for all licensees by 19% from 0.26 cSv (rem) to 0.32

cSv (rem).

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 cSv (rem) per

year regulatory limit would be identified in Table 5.1 in one of the dose ranges >5 rem. In

1995, no individual exceeded this dose limit, and since 1985, there have been no additional

transient workers identified as having received a dose of >5 cSv (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 cSv (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

lung, liver, and bone surfaces). 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)1

. The annual TEDE limit for adult workers is 5 cSv (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 to 25 cSv (rem) or more; an

eye dose equivalent of 0.75 Sv (75 rem) or more; or a shallow-dose equivalent

to the skin or extremities of 2.5 Gy (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 to 5 cSv (rem) or more; an eye

dose equivalent of 0.15 Sv (15 rem) or more; or a shallow-dose equivalent to

the skin or extremities of 0.5 Sv (50 rem) or more in a 24-hour period. The

Commission must be notified within 24 hours of these events.

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*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

• 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 1995.

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 1995, three workers received doses that exceeded the regulatory limit. There were no

occurrences in which individuals received an exposure of the magnitude described previously

as “Category A.” One “Category B” occurrence was reported.

The incident involved an individual working at a multi-location radiography licensee that

received 5.100 cSv (rem) during 1995. The worker received 2.670 cSv (rem) during the first

half of the year, causing the licensee to begin corrective measures. The licensee counseled

 the worker concerning reducing his exposure, but the individual stated that personal problems

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had distracted him. During the third quarter the licensee limited the individual’s work activities,

but by the end of the year the individual exceeded the 5 cSv (rem) TEDE annual limit. The

NRC regional office was notified via telephone and a written report was submitted as required.

Two exposures to the skin in excess of the annual limit of 50 cSv (rem) were reported in 1995.

Both of these exposures were because of “hot particles,” which are small pieces of radioactive

material that can cause high doses to a localized area of the skin of the exposed worker. Both

of the exposures occurred at the same licensee, which is a manufacturer and distributor of

radionuclides (Type A - Broad, see Section 3.3.2). The exposures were from Iridium-192. One

individual received an estimated absorbed dose to the skin of 230 rads in March 1995, and the

other received 342 rem to the skin in September. After the first incident, the NRC issued a

Notice of Violation. Upon the second event the licensee suspended all operations involving

Ir-192 and the NRC began conducting a review of the licensee’s hot particle procedures.

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.

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7 REFERENCES

1. U.S. Atomic Energy Commission, Nuclear Power Plant Operating Experience During

1973, USAEC Report 00E-ES-004, December 1974.*

2. U.S. Nuclear Regulatory Commission, Nuclear Power Plant Operating Experience

1974-1975, USNRC Report NUREG-0227, April 1977.*

3. U.S. Nuclear Regulatory Commission, Nuclear Power Plant Operating Experience

1976, USNRC Report NUREG-0366, December 1977.*

4. M.R. Beebe, Nuclear Power Plant Operating Experience - 1977, USNRC Report

NUREG-0483, February 1979.*

5. Nuclear Power Plant Operating Experience - 1978, USNRC Report NUREG-0618,

December 1979.*

6. Nuclear Power Plant Operating Experience - 1979, USNRC Report NUREG/CR-1496,

May 1981.*

7. Nuclear Power Plant Operating Experience - 1980, USNRC Report NUREG/CR-2378,

ORNL/NSIC-191, October 1982.*

8. Nuclear Power Plant Operating Experience - 1981, USNRC Report NUREG/CR-3430,

ORNL/NSIC-215, Vol. 1, December 1983.*

9. Nuclear Power Plant Operating Experience - 1982, USNRC Report NUREG/CR-3430,

ORNL/NSIC-215, Vol. 2, January 1985.*

10. United Nations, Report of the Scientific Committee on the Effects of Atomic Radiation,

Annex H, General Assembly of Official Records, United Nations, New York, 1982.

11. A. Brodsky, R. Specht, B. Brooks, et al., Log-Normal Distributions of Occupational

Exposure in Medicine and Industry. Presented at the 9th Midyear Topical Symposium of

the Health Physics Society, 1976.

12. S. Kumazawa and T. Namakunai, A New Theoretical Analysis of Occupational

Dose Distributions Indicating the Effect of Dose Limits, Health Physics, Vol. 41, No. 3,

1981.*

7-1

*Report is available for purchase from the National Technical Information Service, Springifeld, Virginia, 22161, and/or the NRC/GPO

Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
NUREG-0713

13. S. Kumazawa and T. Namakunai, A Method for Implementation of ALARA for

Occupational Exposure Using the Hybrid Lognormal Model. Presented at the 27th

Annual Meeting of the Health Physics Society, July 1, 1982.

14. Licensed Operating Reactors, Status Summary Report, USNRC Report NUREG-0020,

Vol. 20, No. 1. Data for 1995 provided on diskette by D. Hartfield, USNRC Office of

Information Resources Management, Systems Development Branch.

15. L.A. Cross and A.P. Cross, Trends in Nuclear Power Plant Man-Rem Per

Megawatt-Year. Presented to American Nuclear Society European Nuclear Society

International Conference, Washington, DC, November 17-20, 1980.

16. National Research Council, Health Effects of Exposure to Low Levels of Ionizing

Radiation: BEIR V, Committee on the Biological Effects of Ionizing Radiations, 1990.

Available from the National Academy Press, 2101 Constitution Avenue NW.,

Washington, DC 20418.

17. Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Part II:

Scientific Basis for Health Effects Models, USNRC Report NUREG/CR-4214, May 1989.

18. Instructions for Recording and Reporting Occupational Radiation Exposure Data,

USNRC Regulatory Guide 8.7, Rev. 1, June 1992.

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