SOURCES AND EFFECTS OFIONIZING RADIATION
United Nations Scientific Committee on the
Effects of Atomic Radiation
UNSCEAR 2008
Report to the General Assemblywith Scientific Annexes

VOLUME I

UNITED NATIONSNew York, 2010

NOTE
The report of the Committee without its annexes appears as Official Records of the General
Assembly, Sixty-third Session, Supplement No. 46.
The designations employed and the presentation of material in this publication do not imply
the expression of any opinion whatsoever on the part of the Secretariat of the United Nations­
concerning the legal status of any country, territory, city or area, or of its authorities, or ­concerning
the delimitation of its frontiers or boundaries.
The country names used in this document are, in most cases, those that were in use at the time thedata were collected or the text prepared. In other cases, however, the names have been updated,
where this was possible and appropriate, to reflect political changes.

UNITED NATIONS PUBLICATION
Sales No. E.10.XI.3
ISBN 978-92-1-142274-0

Corrigendum to Sales No. E.10.XI.3

18 May 2016

Sources and Effects of Ionizing Radiation: United Nations
Scientific Committee on the Effects of Atomic Radiation
2008 Report
Volume I
Annex B (Exposures of the public and workers from various sources ofradiation)
Corrigendum
1.

Page 328, table 3
In the header, for Collective dose per unit release (man Sv/PBq) readCollective dose per unit release (man Sv/TBq)
In the first row, for 2.1 read 0.0021
In the second row, fo r 270 read 0.27
In the third row, for 0.0074 read 0.0000074
In the four th r ow, fo r 0.044 read 44
In the fift h r ow, fo r 0.0003 read 0.3
In the sixth row, fo r 0.0074 read 7.4
In the seventh row, fo r 0.0014 read 0.0000014
In the eight h r ow, fo r 1000 read 1
In the nint h r ow, fo r 4.7 read 0.0047
In the tenth row, fo r 3.3 read 0.0033
In the eleventh row, fo r 99 read 0.099
In the twelfth row, fo r 98 read 0.098

2.

Page 367, table 72, section headed “1985-1989”, column headed“Monitored workers”
In the row headed “Reprocessing ”, fo r 17 read 12

V.16-02700 (E)*1602700*

Corrigendum to Sales No. E.10.XI.3

3.

Page 378, table 92, last row headed “Total”
For entry for the period 1990-1994, for 0.8 read 1.3
For entry for the period 1995-1999, for 0.9 read 1.8
For entry for the period 2000-2002, for 0.8 read 1.8

2

V.16-02700

Corrigendum to Sales No. E.10.XI.3

May 2011

Sources and Effects of Ionizing Radiation: United Nations
Scientific Committee on the Effects of Atomic Radiation
2008 Report to the General Assembly, with Scientific
Annexes—Volume I
Corrigendum
1. Annex A (“Medical radiation exposures”), page 172, figure D-II
The title should read
Representative isodose distributions: Intensity-modulated radiation therapy
planfor a prostate tumour, showing superior conformation of the 50 Gy isodose line
to the planning target volume
2. Annex B (“Exposures of the public and workers from various sources ofradiation”), paragraph 155
The paragraph should read155. Effluents and solid waste. Mining operations have been carried out in open
pits, in underground mines and by in situ leaching. Uranium mill tailings are
generated at about one tonne per tonne of ore extracted, and they generally retain
5–10% of the uranium and 85% of the total activity [V4]. The estimated amounts oftailings worldwide are shown in figure XVII; they total about 2.35 × 10 9 t. Besides thetailings, waste rock piles may also become a source of public exposure. For openpit mining, the amount of debris produced is from 3 to 30 tonnes per tonne ofextracted ore. For underground mining, about ten times less debris is produced.
On the basis of information provided for 13 mining sites in Argentina [R13],
Canada [M28], Germany [F2] and Spain [S29], the amount of waste rock varies
from 40 to 6,000 times the amount of tailings, with an average value of about
1,600 tonnes of waste rock per tonne of tailings [I38].

V.11-80527 (E)*1180527*

Corrigendum to Sales No. E.10.XI.3

15 December 2020

Sources, Effects and Risks of Ionizing Radiation: United
Nations Scientific Committee on the Effects of Atomic
Radiation 2008 Report to the General Assembly, with
Scientific Annexes – Volume I
Corrigendum
Annex B (Exposures of the public and workers from various sources ofradiation), page 231, paragraph 62, line 12
For the existing equation substitute𝐸̇1 (𝑧) = 𝐸̇1 (0)[0.21𝑒 −1.649 𝑧 + 0.79 𝑒 0.4528 𝑧 ]V.20-07529 (E)*2007529*

ANNEX B
exposures of the public and workers
from various sources of radiation
Contents
INTRODUCTION. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
	I.	 Dose assessment issues . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		A.	Public exposure. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		B.	Occupational exposure . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		C.	Special quantities for radon. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .

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	II.	PUBLIC EXPOSURE . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		A.	Natural sources. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 1.	Cosmic radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 2.	Terrestrial radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 3.	Summary of the exposures to natural sources. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		B.	Enhanced sources of naturally occurring radioactive material . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 1.	Metal mining and smelting. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 2.	Phosphate industry. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 3.	Coal mining and power production from coal . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 4.	Oil and gas drilling. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 5.	Rare earth and titanium oxide industries. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 6.	 Zirconium and ceramics industries. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 7.	Applications of radium and thorium . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 8.	Other exposure situations. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 9.	Summary on exposure to enhanced NORM . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		C.	Use of man-made sources for peaceful purposes . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 1.	Nuclear power production . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 2.	Transport of nuclear and radioactive material. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 3.	Applications other than nuclear power. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 4.	Summary on exposure due to peaceful uses of man-made sources of radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		D.	Use of man-made sources for military purposes . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 1.	Nuclear tests . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 2.	Residues in the environment. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		E.	Historical situations. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		F.	Exposure from accidents. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		 G.	Summary on public exposure. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .

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	III.	 Occupational radiation exposure. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
		A.	Assessment methodology. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 1.	Dose recording. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 2.	Characteristics of dose distributions. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .
			 3.	Estimation of worldwide exposures . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .

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		B.	Natural sources of radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 281
			 1.	Cosmic ray exposures of aircrew and space crew . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 282
			 2.	Exposures in extractive and processing industries . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 284
			 3.	 Gas and oil extraction. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 287
			 4.	Radon exposure in workplaces other than mines . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 288
			 5.	Conclusions on occupational exposure to natural sources of radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 289
		C.	Man-made sources for peaceful purposes. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 290
			 1.	Nuclear power production . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 290
			 2.	Medical uses of radiation. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 300
			 3.	Industrial uses of radiation . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 309
			 4.	Miscellaneous uses . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 315
		D.	Man-made sources for military purposes. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 317
			 1.	Other exposed workers. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 319
		E.	Summary on occupational exposure . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 319
CONCLUSIONS ON PUBLIC AND WORKER EXPOSURE. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 322
Tables. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 325
Figures. .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 379
References . .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  .  . 439

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INTRODUCTION
1. The exposure of human beings to ionizing radiation fromnatural sources is a continuing and inescapable feature of lifeon the earth. For most individuals, this exposure exceeds thatfrom all man-made sources combined. There are two main
contributors to natural radiation exposures: high-energy cosmic ray particles incident on the earth’s atmosphere and radioactive nuclides that originated in the earth’s crust and arepresent everywhere in the environment, including the humanbody itself. The world population is also exposed to radiation
resulting from releases to the environment of radioactive
material from man-made sources, and from the use of fuels ormaterials containing naturally occurring radionuclides. In
addition, there are a wide variety of situations in which people
at work are exposed to ionizing radiation. These situations
range from handling small amounts of radioactive material,for example in tracer studies, to operating radiation-­generatingor gauging equipment, to working in installations of thenuclear fuel cycle. There are also situations where the exposure of workers to natural sources of radiation is sufficiently
high to warrant the management and control of radiation as an
occupational hazard. All these exposures were regularly
assessed in previous reports of the Committee, the most recent
being the UNSCEAR 2000 Report [U3]. The purposes ofthese assessments are to improve the understanding of global
levels and temporal trends of public and worker exposure, toevaluate the components of exposure so as to provide a measure of their relative importance, and to identify emerging
issues that may warrant more attention and scrutiny.

difference in responsibilities for managing the protection ofworkers and of the public that is reflected in the different
interests of users of this annex.
3. This annex supplements and updates previous
UNSCEAR publications on the subject. The estimates ofradiation exposure have been based primarily on the submissions to the UNSCEAR databases for assessment of doses tothe public and workers, supplemented by significant reports
in the open literature. The annex does not cover processes
previously described in detail; whenever pertinent, reference
is made to sources where more detailed information may befound. In particular, because the Committee has separately
evaluated exposures due to radon (annex E of the UNSCEAR
2006 Report [U1]), to medical uses of radiation (annex A ofthe 2008 Report) and to accidents (annex C of the 2008
Report), in particular exposures due to the 1986 Chernobyl
accident (annex D of the 2008 Report), these aspects are notdealt with extensively in this annex. Where appropriate,summaries of other evaluations have been reflected here forcompleteness.
4. The Committee has historically described the exposureof members of the general public to the several different
natural and man-made sources of radiation. The principal
objectives of the analysis of public exposures presented in
section II are:
− To evaluate the radiation levels worldwide to whichhuman beings are usually exposed;

2. This annex comprises three sections. Section I addresses
general issues related to dose assessment for public andoccupational exposure to radiation, and the special quantities for measuring and assessing exposure due to radon. Sections II and III address the exposures to ionizing radiation of
the general public and of workers, respectively. The distinction between public and occupational exposure is kept for
two main reasons: (a) the two groups exhibit significant differences with respect to age, the numbers of people exposed,
the relevant exposure pathways, and the methodologies formonitoring and assessing radiation doses;1 and (b) there is a

− To assess the usual variability of exposure worldwide to different sources;− To identify sources of concern for public exposure;− To allow users to derive benchmarks for comparison purposes, to manage exposures and to derive­relationships for their investigative work;− To analyse temporal trends in the contributions ofdifferent sources to overall public exposure.
5. It is often not straightforward to differentiate between
normal exposures and enhanced exposures to natural sourcesof radiation, and between these and exposures to man-madesources. An illustrative example is the common assessmentof radiation exposure indoors, where the natural backgroundradiation exposure is influenced by the presence of naturalradioactivity in building materials, leading to what are sometimes treated as enhanced exposures. Another example is theimpact of the urbanization process, which is known to alter
natural background radiation exposure (e.g. the laying ofpavement reduces exposure from radionuclides in the soil,

While doses to workers are mostly measured, doses to the public are usually assessed by indirect methods, typically using measurements performedin the environment or of environmental samples, modelling various exposure scenarios and employing data on population habits. The accuracyof assessments made usually differs with the methodology used: doses
assessed for workers are normally more accurate than those for members of
the public. Moreover, doses from occupational exposure relate to a specific
set of people, usually healthy adults. Although assessments of doses to the
public sometimes take account of the properties of different age groups ortheir specific habits, the values of the dose estimates do not usually apply to
any specific individual within the population under consideration, but rather
represent an average dose to groups of people.
1

223

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UNSCEAR 2008 REPORT: VOLUME I

whereas the use of granite and certain ceramic materials in
the construction of buildings may enhance exposure). In
addition, especially for developing countries, the expansionof industries (e.g. a new mining installation in an area withhigh levels of background radiation) may enhance public use
and habitation of an area as new infrastructure becomes available, leading to changes in public exposure. Because of these
difficulties, no attempt will be made here to draw a rigorous
distinction between normal and enhanced exposures to natural sources of radiation. Subsection II.A, on public exposure
to natural sources of radiation, includes consideration of
exposures to cosmic and terrestrial sources of radiation.
6. The exposure of the general public to radiation resulting
from industries deemed non-nuclear—such as the mining,
milling and processing of ores that, apart from the raw material, contain uranium (U) and/or thorium (Th)—is describedin subsection II.B on enhanced sources of radiation. Exposures resulting from nuclear industries (i.e. those related to
the nuclear fuel cycle and to artificial radionuclides) aredescribed in two subsections on public exposure to manmade sources. The first of these, subsection II.C, describes
public exposure to man-made sources arising from peacefuluses of atomic energy (including energy generation and theoperation of the associated fuel cycle facilities, the production of radioisotopes, the transport of nuclear and radioactive
material, waste management and the use of consumer products). The second, subsection II.D, presents the public exposures to man-made sources related to military purposes(including atomic weapons tests and their fallout or radio­
active residues, the military use of depleted uranium in war
situations and sites contaminated by waste from previous
practices, mostly associated with the development of nuclear
weapons technology, but not including the exposures due tothe Hiroshima and Nagasaki bombings). As doses received
by the world population due to nuclear explosions have been
described systematically in previous reports of the Committee and a major overview was presented in the UNSCEAR
2000 Report [U3], only a summary regarding the tests andthe resulting worldwide exposures has been included here
for completeness.
7. In section III the Committee has updated its evaluationsof occupational exposures [U3, U6, U7, U9, U10] for workin six broad categories of practice: practices involving elevated levels of exposure to natural sources of radiation; thenuclear fuel cycle; medical uses of radiation; industrial usesof radiation; military activities; and miscellaneous uses ofradiation (which includes educational and veterinary uses).
8. The Committee has evaluated the distributions of annual
individual effective doses and annual collective effective doses
resulting from occupational radiation exposures in the various
practices or due to various types of source. The principal objectives of the analysis of occupational exposures remain, as in
the previous assessments of the Committee, as follows:− To assess annual external and committed internal
doses and cumulative doses to workers (both theaverage dose and the distribution of doses within

the workforce) for each of the major practices
involving the use of ionizing radiation;− To assess the annual collective doses to workers
for each of the major practices involving the use ofionizing radiation. This provides a measure of thecontribution made by occupational exposures to theoverall impact of that use and the impact per unit
practice;− To analyse temporal trends in occupational exposures in order to evaluate the effects of changes inregulatory standards or requirements (e.g. changes
in dose limits and increased attention to ensuringthat doses are as low as reasonably achievable),
new technological developments and modified
work practices;− To compare exposures of workers in different countries and to estimate the worldwide levels of exposure for each significant use of ionizing radiation.
9. According to the International Labour Organization, theformal definition of occupational exposure to any hazardous
agent includes all exposures incurred at work, regardless ofsource [I62]. However, for radiation protection purposes, inorder to distinguish the exposures that should be subject tocontrol by the operating management from the exposuresarising from the general radiation environment in which all
must live, the term “occupational radiation exposure” is often
taken to mean those exposures received at work which canreasonably be regarded as the responsibility of the operating
management [I7, I16, I47]. Such exposures are normally also
subject to regulatory control [I7]. The exposures are usually
determined by individual monitoring, and the doses assessed
and recorded for radiological protection purposes.
10. The terms “practice” and “intervention” have been
widely used in radiological protection. The term “practice”
has been used for human activities that increase the exposureor the likelihood of exposure of people to radiation or thenumber of people exposed. The International Commission
on Radiological Protection (ICRP) had distinguishedbetween “practices” that increase exposure or likelihood of
exposure and “interventions” that reduce exposure or likelihood of exposure [I7, I47]. However, the latest ICRP
­recommendations [I60] use a situation-based approach tocharacterize the possible situations where radiation exposure
may occur as “planned”, “emergency” and “existing exposure” situations. The ICRP now believes that it is more
appropriate to limit the use of the term “intervention” todescribe protective actions that reduce exposure, while theterms “emergency” or “existing exposure” will be used todescribe radiological situations where such protective actions
to reduce exposure are needed [I60]. In this annex the terms“practice” and “intervention” are applied according to theprevious ICRP definitions [I47].
11. The procedures for the recording and inclusion ofoccupational exposures differ from practice to practiceand country to country, and this may influence theANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

225

respective statistics in different ways. Some countries
may overestimate the size of the exposed workforce, andthereby distort assessment of the individual and population dose distributions. Moreover, some countries report
only the doses of workers in controlled areas, while othercountries report the doses from both exposed and nonexposed workers. Some countries do not adequately trackthe doses to contract workers, who may work and accumulate exposure in different industries, possibly even indifferent countries. These issues are discussed in subsection III.A. These differences in monitoring and reporting
practices mean that caution must be applied in ­interpreting
the reported data.

there are a significant number of workers exposed to ionizing radiation who are not individually monitored. The largest proportion of these workers are those exposed to natural
sources of ionizing radiation. Before the implementation of
the International Basic Safety Standards for Protection
against Ionizing Radiation and for the Safety of Radiation
Sources (the “International Basic Safety Standards”) [I7],few data were recorded in national databases on occupational exposure to natural sources of radiation. Recently,
however, exposures to enhanced levels of natural radiation
have become a focus of attention in the field of radiation
protection. Subsection III.B is devoted to natural sources ofoccupational radiation exposure.

12. Although most workers involved in practices that are
subject to controls established by the national regulatory
authorities are individually monitored on a routine basis,

13. Subsections III.C and III.D address occupational exposure to man-made sources of radiation used for peaceful andfor military purposes, respectively.

I.

Dose assessment issues

14. The basic quantity used here to describe radiation
exposure is the “effective dose”. Although this artificial
quantity was developed strictly for protection purposes, it isused here for the purposes of exposure assessment. The
annual committed effective dose includes the sum of external and internal doses and is usually reported in millisieverts(mSv):   (1)
15. The ICRP [I60] has very recently recommended newvalues for some of the radiation and tissue weighting factors
in the definition of effective dose. However, for the evaluations here, the assessment of effective doses has been madeon the basis of the earlier definition provided in ICRP
­Publication 60 [I47].
16. In particular, the Committee continues to use in its estimations of effective dose a radiation weighting factor (wR) of
1 for all photon and beta emitters, including tritium. A recent
report of an independent Advisory Group on Ionising Radiation to the Health Protection Agency (HPA) in the United
Kingdom recommended that the ICRP consider increasing
this value for tritium from 1 to 2 [A3]. The ICRP has considered this recommendation, taking into account recent reviewsof the scientific basis for this value [L18, L19]. It concluded
that, for assessments covered by their broad approach, i.e.
that are not individual-specific, a value of 1 remains­appropriate [C32].
17. In order to compare the total radiation dose from various sources incurred by different groups, the Committee
uses the quantity “collective dose”, which is defined as thesum of all the individual effective doses received in thegroup under consideration. It is expressed in units ofm­ an-sieverts (man Sv) [I7] and is accompanied by the
number of individuals in the group. While this quantity was
also developed strictly for the purposes of optimization ofprotection, it is used by the Committee to assess the relative importance of various sources of radiation exposure.
The collective dose received by a group divided by the
number of individuals in the group is the “average per caput
dose” in this group.
18. The Committee uses the International System of Units
to report data as values that can be easily used and recalled;
specifically, it uses multiples and submultiples of the­standard units, designated by the following prefixes:
peta

(P)

1015

femto (f)

10-15

tera

(T)

1012

pico

(p)

10-12

giga

(G)

109

nano

(n)

10-9

mega (M)106

micro (µ)

10-6

kilo

103

milli

10-3

(k)

(m)

A. Public exposure
19. It is very rare that doses to members of the public are
directly measured. Usually these doses are assessed on thebasis of environmental or effluent monitoring data, using
models to simulate environmental exposure scenarios. These
scenarios and models have been extensively discussed in the
UNSCEAR 2000 Report [U3], and only a summary of themost relevant aspects will be presented here.
20. The estimation of Eext in Eq. (1) depends on the data
available from environmental measurements. The main
quantity used to characterize external exposure fields due to226

UNSCEAR 2008 REPORT: VOLUME I

natural sources is the absorbed dose rate in air, usuallyreported in nanograys per hour (nGy/h). Some authors report
the air kerma, also expressed in nanograys per hour. Under
the assumption that charged-particle equilibrium exists
within the volume of material, the air kerma and the absorbed
dose in air may be assumed to be equivalent. The factor used
to transform measurements of absorbed dose in air to external effective dose to adults is 0.7 Sv/Gy, as described in the
UNSCEAR 2000 Report [U3]. When describing public
exposure, external exposures are assessed using effective
dose rates expressed in units of either nanosieverts per hour(nSv/h) for instantaneous exposure fields, or millisieverts
per year (mSv/a) for estimating the average annual exposureof individuals. The “occupancy fraction”, related to the fraction of time spent indoors, Iin, and the “shielding factors” ofbuildings, SF, describing the ratio of the absorbed dose rateindoors to the absorbed dose rate outdoors, are also used toestimate average annual effective doses:(2)
21. External doses may also be estimated from environmental concentrations of natural radionuclides in soil, Csoil,
using appropriate dose conversion factors, DCFsoil, as­presented in table 1:(3)
22. Internal doses for adults are calculated using the
50 year committed effective doses (i.e. the integrated internal dose received over the 50 years following intake); forchildren, the committed effective doses are integrated up to
the age of 70 years. Very few assessments include estimatesof doses to children. Internal doses to members of the public
are usually estimated on the basis of the scenarios describedin the UNSCEAR 2000 Report [U3], using data on concentrations of radionuclides in the environment, such as concentration in water or food, Ck, expressed in becquerels per
litre (Bq/L) or becquerels per kilogram (Bq/kg), and concentration in air, Cair, expressed in becquerels per cubic
metre (Bq/m3):

  

(4)

where j refers to radionuclides, k refers to the type of food orwater, I is the intake of radionuclide, IR is the inhalation rateor the ingestion rate of foodstuff k, and e is the coefficient for
conversion from intake to committed effective dose, ej (50),i.e. the committed effective dose integrated for 50 years for
adults, and ej (70), i.e. the committed effective dose integrated up to the age of 70 years for children, separately forinhalation and ingestion. The dose conversion coefficientsused in this annex for adults for doses due to intakes of­natural radionuclides are also presented in table 1.

23. To assess doses due to the operation of nuclear power
plants and other fuel cycle facilities, the dose conversion
coefficients derived in the UNSCEAR 2000 Report [U3]have been used. These coefficients are specified in terms ofthe collective effective dose per unit release of a radionuclide. They are presented in table 2 for nuclear reactors andin table 3 for reprocessing facilities. For other fuel cycle
facilities, collective doses have been estimated on the basisof the electrical energy generated and the same dose coefficients as used in [U3], namely 0.2 man Sv/(GW a) for
operational uranium mining, 0.0075 man Sv/(GW a) for
operational tailings piles, 0.00075 man Sv/(GW a) for
releases from residual tailings piles, 0.003 man Sv/(GW a)
for uranium enrichment and fuel fabrication facilities,
and 0.5 man Sv/(GW a) for the disposal of low- and­intermediate-level waste. The Committee has decided not
to extend its estimates of doses into the far future, as wasdone in previous reports, because of the very large uncertainty inherent in such assessments. Thus only current
doses received by members of the public are described inthis annex.
24. For the assessment of exposures due to military uses ofradiation, the main quantity used is also the effective dose,
although sometimes the equivalent dose to specific organs,such as the thyroid, have also been reported. Both quantities
are expressed in units of millisieverts, but when the term
“dose” refers to a specific organ dose, this is made clear in
the text. In this section, estimates for doses occurring in thepast, present and near future are given. The future doses aremainly related to possible or predicted exposures due to theuse of contaminated sites.

B. Occupational exposure
25. The ICRP, in its Publication 60 [I47], indicated thatthree important factors influence the decision to undertake
individual monitoring: the expected level of dose or intake inrelation to the relevant limits; the likely variations in thedoses and intakes; and the complexity of the measurement
and interpretation procedures that make up the monitoring
programme. Where doses are consistently low or predictable, other methods of monitoring are sometimes used, as in
the case of aircrew for whom doses can be calculated fromflight rosters. The complexity of measurement techniques
results in an approach to monitoring for external irradiationthat is different from that for intakes and the resulting­committed dose.
26. The estimate of the effective dose, E(t), needs to take
into account the contribution from external and internal
exposure, if appropriate. E(t) can be estimated using the­following expression:(5)

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

where HP(d) is the personal dose equivalent during time
period t at a depth d in the body (normally 10 mm for penetrating radiation); ej,inh(50) is the committed effective dose
per unit activity intake by inhalation of radionuclide j, integrated over 50 years; Ij,inh is the intake of radionuclide j byinhalation during the time period t; ej,ing(50) is the committed
effective dose per unit activity intake by ingestion of radionuclide j, integrated over 50 years; and Ij,ing is the intake ofradionuclide j by ingestion during time period t. Uptake
through the skin and wounds can occur in some circumstances. For most forms of intake, the dose coefficients provided by the ICRP are for intakes by inhalation and ingestionand do not take into account uptake through the skin.
27. The United States National Council on Radiation Protection and Measurements (NCRP), in collaboration with the
ICRP, has developed a biokinetic and dosimetric model forradionuclide-contaminated wounds. The multicompartment
model uses first-order linear biokinetics to describe theretention and clearance of a radionuclide deposited on thewound site. Seven default categories have been defined todescribe wound site retention: four relate to contamination
with initially soluble materials (weak, moderate, strong andavid), and three relate to contamination with materials having solid properties (colloid, particle and fragment). The
wound model is coupled to the ICRP systemic models forpredicting urinary and faecal excretion patterns, as well as
for producing wound-specific dose coefficients. However,
the resulting dose coefficients are not yet available, andtherefore the doses estimated in this annex were based on thedose coefficients for ingestion or inhalation [G15].
28. One of the factors regarding the uncertainty of theexternal dose assessment concerns how and where personal
dosimeters should be worn in order to obtain the best estimate of effective dose or equivalent dose, as appropriate. In
general, a dosimeter is placed on the front of the body; this issatisfactory provided that the dosimeters have been designed
to measure HP(10). In medical radiology, where lead aprons
are used, different approaches have been adopted. In somecases, the assessment of effective doses to workers is carriedout by means of a dosimeter worn on the trunk, under theapron. Where doses are likely to be higher, for example ininterventional radiology, two dosimeters are sometimes
used, one worn under the apron and a second worn outside.
The purpose of the second dosimeter is to assess the contribution to the effective dose due to the irradiation of unshieldedparts of the body [N9]. Where doses are low and individual
monitoring is intended only to give an upper estimate of
exposure, single dosimeters might be worn outside theapron.
29. Measurements made on phantoms using X-ray beamsof 76 and 104 kVp have shown that, while estimates of theeffective dose without the lead apron were within 20% of
the expected values, estimates with the dosimeter worn on
the waist underneath the lead apron were lower than the
expected values [M12]. Such results suggest that accurate
estimation of effective dose using personal dosimeters under227

conditions of partial-body exposure remains problematicand that to be fully accurate would probably require thatmultiple monitors be used, which is not often done. Differing monitoring practices in medical radiology may therefore affect the validity of the data for comparison purposes.
Since the position of the dosimeter in relation to the leadapron is not standardized among countries, a large apparent
fluctuation of dose values could result unless algorithms
that yield more precise estimates are used to convert themeasured quantity to effective dose [N9]. Variations in thedesign of the lead apron itself and in its thickness may represent additional sources of uncertainty. These uncertainties
and how they are addressed by dosimetry services could
also have an impact on the comparisons made here. In thisannex it is assumed that all these parameters have been
properly considered in dose estimation.
30. The conversion coefficients for use in radiological protection against external irradiation are given in ICRP Publication 74 [I56]. Except for radon progeny, values of thecommitted effective dose per unit intake for inhalation,
ej,inh(50), and ingestion, ej,ing(50), are found in ICRP Publication 68 [I50], which takes account of the tissue weighting
factors in ICRP Publication 60 [I47] and the new lung modelin ICRP Publication 66 [I51]. It is assumed that the data provided to the Committee have been based on these conversion
coefficients. A number of difficulties may be encountered indetermining occupational exposure. These difficulties may
be addressed in various ways, as is evident in the variety ofmonitoring procedures and dose recording practices adopted
in countries throughout the world. While some countries
have already adopted the recommendations of ICRP Publication 60 [I47], a significant proportion of countries are stillusing the dose limits and the quantities of ICRP Publication
26 [I43], especially for the first period analysed in the current annex (1995–1999). This may be a factor in explaining
the variation in doses for a given practice among different
countries. Quantities for radiation exposure and the methodologies for external and internal dose assessment have been
well described in the UNSCEAR 2000 Report [U3], andbecause the measured quantities and the techniques describedin that report remain unchanged, the issue need not be
addressed further here.
31. Intakes of radioactive material are normally assessedroutinely for workers employed in areas that are designatedas controlled (specifically in relation to the control of contamination) or in which there are grounds for expecting significant intakes [I13, I55]. However, there are difficulties incomparing data on doses due to intakes of radionuclides indifferent countries because of the different approaches used
for monitoring and to interpreting the results. Several international intercomparison exercises for internal dose assessment have been organized, of which the largest so far was theThird European Intercomparison Exercise on Internal Dose
Assessment, organized in the framework of the EULEP/EURADOS Action Group [D11, I15]. The most importantlesson from these intercomparison exercises was that there
was a need to develop agreed guidelines for internal dose

228

UNSCEAR 2008 REPORT: VOLUME I

evaluation procedures in order to promote the harmonizationof assessments between organizations and countries. Significant differences were revealed among laboratories in theirapproaches, methods and assumptions, and consequently intheir results. One major source of divergence at the time ofthe exercise was due to the particular ICRP models used.
Most dosimetry services were using the models from ICRP
Publications 26 [I43] and 30 [I44] for legal reasons. However, most were in the process of moving to the new generation of ICRP models (Publications 56 [I46], 60 [I47], 66[I51], 67 [I49], 68 [I50], 69 [I52], 71 [I53], 72 [I54], 78 [I55]
and 100 [I58]), partly because these are considered to be
more realistic and partly because of the imminent implementation of the International Basic Safety Standards [I7] andthe new Euratom directive, which are based on the new models [C29, D10, D12, H30, I14]. Similar projects aiming toharmonize internal dosimetry procedures have been carriedout in different parts of the world under the auspices of theInternational Atomic Energy Agency (IAEA) [M20].
32. Since its Publication 60 [I47], the ICRP has revised the
biokinetic and dosimetric models used in internal dosimetry,specifically: the model for the respiratory tract [I51]; themodel for the alimentary tract [I56]; systemic models [I46,
I49, I52] and dosimetric models [I54]. The new ICRP biokinetic and dosimetric models have changed the dose coefficients used for internal dosimetry. The ratios of the dosecoefficients for workers based on the models of ICRP Publication 68 [I50] to those based on the models of Publication
30 [I44] have been calculated for about 800 radionuclides.
For inhalation, about 40% of the ratios fall in the range 0.7–
1.5, about 4% of the ratios are greater than 10 and about
1.4% are less than 0.1. For ingestion, about 73% of the ratios
fall in the range 0.7–1.5, about 3.4% are greater than 10 andabout 1.3% are less than 0.1. The analysis addressed both
inhalation and ingestion of radionuclides in the workplaceand included almost all the radionuclides (some 800) considered in ICRP Publication 30. The tissues considered werethe lungs, stomach wall, colon wall, bone surface, red marrow, liver, thyroid, breast, testes and muscle. The solubility
classes were those considered in ICRP Publication 30. Dose
coefficients for the absorption types (Types F, M and S) currently used by the ICRP were compared with coefficients forClass D, W and Y compounds, respectively, as defined in
ICRP Publication 30. The inhalation dose coefficients generated by the models of ICRP Publication 30 were based on
the default particle size of 1 μm (AMAD) recommended inthat publication, and the coefficients generated by models of
ICRP Publication 68 were based on the default particle sizeof 5 μm recommended in that publication. As an example,
the ratio of the dose coefficient from ICRP Publication 68 tothat from ICRP Publication 30 for the inhalation of insoluble239
Pu compound is 0.07 for bone marrow and for the inhalation of insoluble 238U compound is 0.13 for the lung. The
ratios clearly depend on the radionuclide and on factors suchas retention in the body and solubility [L6, P9].
33. The application of different ICRP methodologies forintake and dose calculations obviously affects the resultsof dose assessments. This can be an important source ofvariation between the doses reported by different countries
for the period under consideration, when most of the countries changed from ICRP Publication 26 [I43] to ICRP
­Publication 60 [I47] recommendations.
C. Special quantities for radon
34. The health risk due to exposure to 222Rn (radon) and220
Rn (thoron) comes principally from the inhalation of theshort-lived decay products and the resulting alpha particle
irradiation of the bronchial airways. The radiation dosedelivered to the respiratory system, and the resulting potential health detriment, are a complex function of the radon
decay product aerosol characteristics and the physiologicalparameters of the exposed individual. The radon and thoron
dosimetry described in this annex is a summary of section II
in annex E of the UNSCEAR 2006 Report [U1].
35. Radon and thoron decay product exposure rates are
expressed by the measure of potential alpha energy concentration (PAEC), with units of joules per cubic metre (J/m3)
for the equilibrium equivalent concentration (EEC) or becquerels per cubic metre (Bq/m3) for the working level (WL:
unit of concentration of radon progeny in one cubic metre of
air that has the potential alpha energy of 2.08 × 10–5 J for
222
Rn). The PAEC is derived from a linear combination of theactivities of the short-lived decay products in each radon
decay series (see paragraph 122, annex B of the UNSCEAR
2000 Report [U3]). The constants in the linear combination
are the fractional contributions of each decay product to thetotal potential alpha energy from the decay gas. The EEC (inunits of Bq/m3) can be converted to the PAEC by the
relationships:
1 Bq/m3 = 5.56 × 10‑6 mJ/m3= 0.27 mWL (222Rn)
and
1 Bq/m3= 7.6 × 10‑5 mJ/m3 = 3.64 mWL (220Rn).
36. As discussed in annex E of the UNSCEAR 2006
Report [U1], estimates of radiation dose and the resulting
risk from inhalation of radon decay products can be derived
from either epidemiological studies or dosimetric models.
For occupational exposure to inhaled radon decay products,
the ICRP recommended in Publication 65 [I48] the use of asingle conversion factor based on the results of the uranium
miner epidemiological studies, by equating the radiation
detriment coefficient (risk per sievert) with the miner detriment (risk per PAEC exposure). For worker exposure, thisfactor is 1,430 mSv/(J h m‑3) (rounded to 1,400 mSv/
(J h m‑3)), 5.06 mSv per working level month (WLM)(rounded to 5 mSv/WLM) or 7.95 nSv/(Bq h m‑3) (rounded
to 8 nSv/(Bq h m‑3)) EEC [U1]. The working level monthcorresponds to the exposure resulting from the inhalation of
air containing 1 WL for 170 h. The countries reporting data
often do not specify which dosimetric model was used to
calculate the dose, although it is likely that the ICRP
approach was used [I7].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

37. The results of the dosimetric model agree with theconversion convention within a factor of 2 and depend on
the value for the radiation weighting factor. Until further
clarification of the factor is available, the Committee considers that the established value of 9 nSv/(Bq h m‑3) usedin past UNSCEAR calculations [U3, U6, U7] is still
appropriate for its purpose of evaluating average effective
doses [U1].
38. It is not possible to assess the radiation dose due toinhalation of thoron decay products by epidemiologicalmeans, and the dose conversion factor must therefore be estimated using dosimetric modelling. Annex A of the
UNSCEAR 2000 Report [U3] indicated that a conversion
factor for thoron decay products could be derived on thebasis of the recommendations given in ICRP Publication 50

II.

229

[I45], which in turn were based on the results of an Expert
Group of the Nuclear Energy Agency [N20]. According toreference [U3], this value is intended to include the dose toorgans other than the lungs resulting from the transfer of212
Pb from the lungs to these other organs. The principal
dosimetric assessments of lung dose due to deposited thoron
decay products support the continued use (see annex E of the
UNSCEAR 2006 Report [U1]) of a conversion factor of
40 nSv/(Bq h m‑3) EEC.
39. For the present annex, most countries would probably
have estimated doses on the basis of ICRP dosimetric factors
developed after ICRP Publication 60 [I7, I47]. The ICRP iscurrently reviewing its biokinetic and dosimetric models,
which will certainly influence dose estimation for futureevaluations.

PUBLIC EXPOSURE

40. Public exposure has been evaluated by the Committee
for two broad classes: exposure to natural radiation sourcesand exposure to man-made sources. In previous reports,
these two classes were usually described in separate annexes.
In this annex, exposures to these two types of source are considered together. Exposures to man-made sources frompeaceful and from military uses of nuclear energy aredescribed separately.

most significant part of their total exposure to radiation.
Radon is usually the largest natural source of radiation contributing to the exposure of members of the public, sometimes accounting for half the total exposure from all sources[W6].

41. The data used in this section have been obtained in thesame way as for previous UNSCEAR reports, i.e. from the
UNSCEAR Global Survey on Public Radiation Exposures,conducted by means of questionnaires distributed to member States by the UNSCEAR Secretariat, and from the published scientific literature. There are many uncertainties
associated with the information provided here, owing to thedifferent ways in which countries collect, analyse and manage their own data. These uncertainties reflect differences in
the methodologies for sampling, measuring, treating andreporting the data, as well as differences in assessment
approaches, for example the use of different dose conversion factors. The Committee recognizes that there is a needto establish standard methodologies to be used worldwide
in order to improve the comparison and manipulation ofreported data and ­therefore to be able to draw more reliableconclusions.

43. Cosmic radiation can be divided into different types
according to its origin, energy and type, and the flux densityof the particles. When only the types important for exposureof humans are taken into account, there are three main
sources of such cosmic radiation: galactic cosmic radiation,
solar cosmic radiation and radiation from the earth’s radiation belts (Van Allen belts) [S30].

A.

Natural sources

42. Human exposure to natural radiation sources has
always existed. The earth has always been bombarded byhigh-energy particles originating in outer space that generatesecondary particle showers in the lower atmosphere. Additionally, the earth’s crust contains radionuclides. For mostindividuals, exposure to natural background radiation is the1.

Cosmic radiation

44. Besides the shielding provided by the earth’s magnetic
field, which is discussed in section II.A.1(c) below, life isshielded against this radiation by an air layer of approximately 10,000 kg/m2 (1,000 g/cm2), which is comparable toa 10 m thick water layer. As a result, at sea level the cosmic
radiation contributes about 10% of the total dose rate fromnatural radiation to which human beings have always been
exposed. However, at higher altitudes in the atmosphere or inspace, cosmic rays constitute the dominant radiation fields[H20].
45. These cosmic rays interact with the nuclei of atmospheric constituents to produce a cascade of interactions andsecondary reaction products that contribute to cosmic ray
exposures. These decrease in intensity with increasing depth
inside the atmosphere, from aircraft altitudes to ground level.
The cosmic ray interactions also produce a number of radio­
active nuclei known as cosmogenic radionuclides. The cosmogenic radionuclide most relevant to public exposure is 14C.

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UNSCEAR 2008 REPORT: VOLUME I

Galactic cosmic radiation

46. Galactic cosmic rays (GCRs) arise from sources outside the solar system, from deep space. The GCRs incident
on the upper atmosphere consist of a nucleonic component,
which in aggregate accounts for 98% of the total, and electrons, which account for the remaining 2%. The nucleoniccomponent is primarily protons (85.5% of the flux) and alphaparticles (~12%), with the remainder being heavier nuclei(~1%) up to that of uranium [S30, U3].
47. These primary cosmic particles have an energy spectrum that extends from 108 eV to more than 1020 eV. Below
1015 eV, the shape of the energy spectrum can be represented
by a power function of the form E–2.7, where E is in electronvolts. Above that point, known as the “knee”, the spectrum
steepens to a power of –3. The highest energy measured
thus far is 3.2 × 1020 eV, which was inferred from groundmeasurements of the resulting cascade interactions in theatmosphere [U3].
48. It is thought that all but the highest-energy cosmic rays
reaching the earth originate within our own galaxy. The
sources and acceleration mechanisms that create cosmic rays
are uncertain, but one possibility (substantiated by measurements from a spacecraft) is that the particles are energizedby shock waves expanding from supernovas. The particles
are confined and continually deflected by the galactic magnetic field. Their flux becomes isotropic in direction and isfairly constant in time [U3].
49. Above 1015 eV, protons begin to escape galactic confinement. This leaves relatively higher proportions ofheavier nuclei in the composition of cosmic rays above
this energy level. Protons with energies of greater than
1019 eV would not be significantly deflected by the intergalactic magnetic field. The fact that the flux of protons ofsuch high energy is also isotropic and not aligned with theplane of the galactic disc suggests that the protons areprobably of extragalactic origin. Only astrophysical theories can suggest the origins of these ultra-high-energy
cosmic rays [U3].
50. The GCR fluence rate varies with solar activity,
being lower when solar activity is higher. The spectrum of
GCRs also changes with solar activity; when solar activity is higher, the maximum of the energy spectrum isshifted to higher energies. GCR particles have to penetrate
the earth’s magnetic field; because of this, a geomagnetic
cut-off exists, which is much more important close to theequator than at the geomagnetic poles. The cut-off is characterized by a “rigidity”, Rc. Rigidity is defined as themomentum of the cosmic ray particle divided by its
charge. Owing to this influence, the number of particles
penetrating the atmosphere is higher close to the earth’s
poles and their spectrum there is softer. Because of this,the effect of solar activity is relatively more important
close to the geomagnetic poles [S30].

(b)

Solar cosmic radiation

51. Another component of cosmic rays is generated
near the surface of the sun by magnetic disturbances.
Solar cosmic radiation (SCR) originates from solar flares
when the particles produced are directed towards theearth. These solar particle events are comprised mostly of
protons (~99% of the flux), with energies generally below
100 MeV and only rarely above 10 GeV. These particles
can produce significant dose rates at high altitudes, but
only the most energetic ­contribute to doses at groundlevel.
52. Solar particle events, in addition, can disturb theearth’s magnetic field in such a way as to change the galactic particle intensity. These events are of short duration,
typically a few hours, and are highly variable in theirstrength. They have a negligible impact on long-term doses
to the general population. A long-term forecast of solar
flares in terms of either intensity or energy spectrum is notpossible. Solar flares are more frequent at periods of maximum solar activity, with the largest at the end of such periods. The geomagnetic field also influences the penetrationof SCR to the earth’s surface. Because of the lower energies,
this influence on SCR is much more important than that on
GCRs [S30, U3].
53. The most significant long-term solar effect is the
11-year solar activity cycle, which generates a corresponding cycle in total cosmic radiation intensity. Historical solar
cycles are shown in figure I. The periodic variation in solar
activity produces a similar variation in the solar wind,which is a highly ionized plasma with an associated magnetic field whose varying strength modulates the intensityof galactic cosmic radiation. At times of maximum solar
activity, the field is at its highest and the galactic cosmic
radiation intensity is at its lowest. An example of the effectof solar modulation on dose rate at aircraft altitudes isshown in figure II.

(c)

Van Allen radiation belts

54. The Van Allen radiation belts are formed throughthe capture of protons (mainly) and electrons by theearth’s magnetic field. The proton energy can reach several hundred megaelectronvolts; the electron energy canreach only a few megaelectronvolts and the electrons’
penetration is therefore limited. There are two van Allenradiation belts, an internal one centred at about 3,000 km
and an external one centred at about 22,000 km from theearth’s surface. The daily equivalent dose to the skin in
the internal belt could reach several tens of sieverts for
protons and several thousands of sieverts for electrons.
The internal radiation belt descends rather close to theearth’s surface in the region called the South Atlantic
Anomaly, which is centred at about 800 km east of Porto
Alegre, Brazil [S30].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

(d)

Effects of latitude and altitude

55. Latitude effects. The earth’s magnetic field reduces theintensity of cosmic radiation reaching the upper atmosphere.
The shape of the earth’s magnetic field is such that only particles of higher energies can penetrate at lower geomagnetic
latitudes. This produces the “geomagnetic latitude effect”,
with intensities and dose rates minimal at the equator andmaximal near the geomagnetic poles. The latitude effect at
20 km altitude is shown in figure III.
56. Near the earth, the geomagnetic field acts as a separator of the incident cosmic particles according to their energy(in reality, according to their rigidity). The relationshipbetween particle energy and rigidity, which defines thethreshold below which particles are unable to reach a particular location because of the effective shielding by the­geomagnetic field [B23], is:E = ( RZe / A)2 + m 2 − mwhere E is the energy per nucleon in GeV, R is the rigidity in
GV, Ze is the nuclear charge, A is the atomic weight and m isthe nucleon mass in GeV [O1]. For highly energetic protons,
the particle energy and rigidity are quite similar. Each geomagnetic latitude may be characterized by a cut-off rigidity,
such that particles with less rigidity cannot arrive at this­latitude. The cut-off rigidity (Rc) is given by:

Rc = 14.9 cos4 (λ)
where λ is the geomagnetic latitude. Equatorial latitudes are
the most protected regions. Only particles with rigidities
greater than 15 GV and protons with energies of greater than
14 GeV are able to reach the equatorial regions [B14].
57. Altitude effects. High-energy particles incident on theatmosphere interact with atoms and molecules in the air andgenerate a complex set of secondary charged and uncharged
particles, including protons, neutrons, pions and lower-Z
nuclei. The secondary nucleons in turn generate more nucleons, producing a nucleonic cascade in the atmosphere. Neutrons, because of their longer mean free path, dominate thenucleonic component at lower altitudes. As a result of thevarious interactions, the neutron energy distribution peaks at
between 50 and 500 MeV. A lower energy peak, at around
1 MeV, is produced by nuclear de-excitation (evaporation).
Both components are important for the assessment of cosmic
ray exposures.
58. Pions generated in nuclear interactions are the main
source of other components of the cosmic radiation field in
the atmosphere. Neutrally charged pions decay into highenergy photons; these produce high-energy electrons that inturn produce more photons and so on, resulting in the “electromagnetic” or “photon/electron” cascade. Electrons andpositrons dominate the charged particle fluence rate at middle altitudes. Charged pions decay into muons, whose longmean free path in the atmosphere makes them the dominant

231

component of the charged particle flux at ground level. They
are also accompanied by a small flux of “collision” electrons
that are generated along their path.
59. The changing components of dose caused by secondary cosmic ray constituents in the atmosphere are illustrated
in figure IV. At ground level, the muon component is themost important contributor to dose, while neutrons, electrons, positrons, photons and protons are the most significant
components at aircraft altitudes. At even higher altitudes, theheavy-nuclei component must also be considered.

(e)Exposure to cosmic radiation

60. Exposures at ground level. At ground level, muons(with energies mainly of between 1 and 20 GeV) constitutethe dominant component of the cosmic ray field. They contribute about 80% of the absorbed dose rate in free air arisingfrom the directly ionizing radiation; the remainder comesfrom electrons produced by the muons or present in the electromagnetic cascade. In the early literature, these two components of the charged particle flux were referred to as the“hard” and the “soft” component, respectively, with reference to the difference in their penetrating power, the electrons being much more readily absorbed by any shielding.
As altitude increases, electrons become more important­contributors to the dose rate.
61. The dose rate from the photon and ionizing component
is known to vary with latitude, but the variation is small. The
dose rate is about 10% lower at the geomagnetic equator
than at high latitudes. Considering the population distribution with latitude, an average dose rate in free air at sea levelof 31 nGy/h has been adopted by the Committee [U3]. This
figure also takes into account the variability due to the solar
cycle, estimated to be about 10%. The population distribution of the effective dose rates outdoors at sea level due to theionizing component of cosmic rays is shown in table 4. The
worldwide population considered was 4 × 10 9 persons [U3].
Because the main contributors to human exposure at groundlevel are muons, a radiation weighting factor of 1 is assumed,
leading to a worldwide average annual effective dose at sea
level of about 0.27 mSv.
62. The ionizing component is, however, strongly dependent on altitude. For the same latitude, a variation by a factor
of about 4 in the absorbed dose rate in free air was measured
in China between sea level and 4,000 m altitude in Tibet[W2]. Dose rates in Switzerland were estimated to be in therange 40–191 nSv/h, with an average value of 64 nSv/h.
Combining the results for dose rates with population density,
the average per caput dose rate in Switzerland was estimatedto be 46 nSv/h [R23]. Estimates of cosmic ray dose rates atelevations above sea level are made using a procedure­published by Bouville and Lowder [B45]:.

.

E 1 ( z ) = E 1 (0) 0.21 e−1.649 z + 0.79 e 0.4528 z 

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UNSCEAR 2008 REPORT: VOLUME I

where Ė1(0) is the dose rate at sea level and z is the altitude
in kilometres. Some two thirds of the world population lives
in coastal regions, but because dose rates increase with altitude, the dose rates of populations at high altitudes contribute proportionately more to the weighted average. For the
directly ionizing and photon component, the populationweighted average dose rate is 1.25 times that at sea level.
Using a shielding factor of 0.8 and an indoor occupancy
fraction of 0.8, the worldwide average annual effective dose
due to the ionizing component of cosmic radiation is­estimated to be about 0.28 mSv.
63. For the neutron component, both latitude and altitude
strongly affect exposure rates. A latitude-averaged fluence
rate at sea level of 130 m–2 s‑1 for latitude 50° N has been
derived. The effective dose rate obtained, applying a weighting factor for the neutron fluence energy distribution of
0.02 pSv/m2, is 9 nSv/h. The shape of the neutron energy
spectrum at habitable altitudes is considered to be relatively
invariant, and therefore it is expected to be generally valid to
use a simple coefficient to convert fluence to effective dose
(isotropic). On this basis, the annual effective dose at sea
level and at 50° latitude due to neutrons is estimated to be
0.08 mSv.
64. Neutrons arise from collisions of high-energy protons
within the upper atmosphere. Incoming protons that initiate
the cosmic ray neutron field are strongly affected by theearth’s magnetic field, with the effect that the neutron fluence rate in equatorial regions is less than that in polar
regions. Florek et al. [F11], quoting results of the Los Alamos LAHET code system calculation, suggest that the equatorial neutron fluence rate at sea level is 20% of the polar
fluence rate and that the fluence rate at 50° latitude is 80% of
the polar fluence rate. The world population-weighted average effective dose rate at sea level due to cosmic ray neutrons thus determined is 5.5 nSv/h or 0.048 mSv/a [U3]. The
population distribution for the effective dose rates outdoors
at sea level due to the neutron component of cosmic rays isalso shown in table 4.
65. For the neutron component of cosmic rays, there is also
a substantial altitude effect. Bouville and Lowder [B45] usedboth measurements and calculations to derive expressions of
the altitude dependence at habitable elevations around theworld:.

.

.

E N ( z ) = E N (0) bN e

indoor occupancy fraction of 0.8, the world average annual
effective dose due to the neutron component of cosmic radiation is estimated to be 0.1 mSv. The population-weighted
average annual doses for each hemisphere and for the world
are summarized in table 5. Overall, the range of averageannual effective dose to the world population is 0.3–2 mSv,
with a population-weighted average of 0.38 mSv [U3].
67. Exposures at aircraft altitudes. Exposure to cosmic
radiation increases rapidly with altitude. Persons who fly frequently are exposed to elevated levels of cosmic radiation ofgalactic and solar origin and to secondary radiation producedin the atmosphere, aircraft structure, etc. The cosmic particle
flux depends on solar activity and solar eruptions. The radiation field at aircraft altitudes consists of neutrons, protons,
and neutral and charged pions. Neutrons contribute 40–80%
of the equivalent dose rate, depending on altitude, latitude
and time in the solar cycle.
68. Commercial transport aircraft altitudes are typically
6,100–12,200 m, where the dose rate doubles for every
1,830 m of increased altitude. The aircraft fuselage provides
little shielding against cosmic radiation [B43, W5]. Exposures of aircrew are described in section III.B.1 of this annex.
The dose received during a particular flight depends on altitude, latitude and flight time. For altitudes of between 9 and
12 km and a latitude of 50° (corresponding to a flight from
northern Europe to North America), the dose rate is generally in the range 4–8 µSv/h. Dose rates at lower latitudes are
generally lower; hence a dose rate of 4 µSv/h may be used torepresent the average dose rate for all long-haul (e.g. transAtlantic) flights. For short-haul flights the flight altitude is
generally lower, between 7.5 and 10 km. At this altitude, thedose rate is typically 3 µSv/h. These average dose rates
include an allowance for the dose received during the climb
and descent phases of the flight. A study in the United Kingdom estimated an average per caput dose of about 30 µSv tothe United Kingdom population due to radiation exposure
during air travel. However, this value cannot be extended tothe populations of all countries, because the exposure isstrongly influenced by the frequency of air travel, which inturn depends on the country’s economic and developmentlevel [W6].
(f)Cosmogenic radionuclides

azwhere E N (0) is the effective dose rate at sea level due toneutrons:   bN = 1 and a = 1 km–1 for z < 2 km;   bN = 2 and a = 0.7 km–1 for z > 2 km [U6].
66. Combining these altitude–dose relationships with theiranalysis of the distribution of the world population with­altitude, these investigators derived estimates for the population-weighted average dose rate due to neutrons as 2.5 times
the value at sea level. Using a shielding factor of 0.8 and an

69. The interaction of cosmic radiation with nuclei present
in the atmosphere produces elementary particles and also aseries of radionuclides. A comprehensive list of cosmogenic
radionuclides (with their properties, production rates andaverage tropospheric concentrations) was included in the
UNSCEAR 2000 Report [U3]. Production is greatest in theupper stratosphere, but some energetic cosmic ray neutronsand protons survive into the lower atmosphere, producing
cosmogenic radionuclides there as well. Production isdependent not only on altitude but also on latitude, as well asvarying with the 11-year solar cycle, which modulates­cosmic ray penetration through the earth’s magnetic field.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

70. Except for 3H, 14C, 22Na and 7Be, which are isotopes of
elements with metabolic roles in the human body, the cosmo­
genic radionuclides contribute little to radiation doses andare of relevance mainly as tracers in the atmosphere and,after deposition, in hydrological systems [U3]. Carbon-14(t1/2 = 5,730 a) arises from the interaction of slow cosmic
neutrons with 14N. Transformed into 14CO2, it participates in
the photosynthetic cycle. Today, the specific activity of 14C isapproximately 230 Bq/kg of total carbon, and the content in
the human body is about 2,700 Bq, resulting in an averageannual individual effective dose of about 12 µSv.
71. The production of 14C from cosmic ray neutrons is relatively constant at an annual rate of 1.4 PBq, resulting in a
global atmospheric inventory of 140 PBq [U10]. A best estimate of the specific activity of naturally produced (cosmicray) 14C prior to industrialization is 222 Bq/kg of total carbon [N7]. The nuclear test explosions from the 1950s and1960s introduced an estimated 0.35 EBq. This was absorbed
into the marine environment with a half-life of about 6 a. The
specific activity of 14C from weapons residues is currently
about 0.05 Bq/kg in the atmosphere. Releases from nuclear
power reactors are also very small. It has been suggested thatthe addition of 12CO2 from the burning of fossil fuels would
dilute the naturally produced 14C and that the measurementof the 14C/C ratio could then be used as an indicator of the
carbon addition to the planet on a global scale [S44]. On­goingmeasurements and recent data available are not conclusive inthis respect, as current specific activity levels of 14C are stillslightly higher than those observed in 1950 [R18].
72. Tritium (t1/2 = 12.3 a) results from the interaction ofcosmic rays with nitrogen and oxygen nuclei; the tritiatedwater produced participates in the water cycle. Its concentration level is about 400 Bq/m3 in continental water and
100 Bq/m3 in the oceans. On average a human ingests
500 Bq/a, with a resulting average annual dose of 0.01 µSv.
73. Beryllium-7 (t1/2 = 53.6 d) has a concentration of
3 mBq/m3 in air. It reaches the earth in rainwater, thus contributing to an annual commitment for individuals of approximately 1,000 Bq through the ingestion of fresh vegetables,delivering an annual effective dose of 0.03 µSv.
74. The annual commitment of 22Na (t1/2 = 949.7 d) is approximately 50 Bq, but this contributes an annual effective dose of
approximately 0.15 µSv, significantly more than for tritium.
The radiation exposure of populations due to cosmogenic
radio­nuclides is therefore dominated by the production of 14C
and is slightly greater than 12 µSv/a [M22].
2.

Terrestrial radiation

75. Naturally occurring radionuclides of terrestrial origin,also termed primordial radionuclides, are present in various
degrees in all environmental media, including the humanbody. Only those radionuclides with half-lives comparable
to the age of the earth, and their decay products, exist in

233

sufficient quantity to contribute significantly to population
exposure. Exposures to radon have been described in annex Eof the UNSCEAR 2006 Report [U1].
(a)Sources of external radiation exposure

76. The main contribution to external exposure comes fromgamma-emitting radionuclides present in trace amounts inthe soil, mainly 40K and the 238U and 232Th families. Information on outdoor exposure comes from direct measurementsof dose rate or from evaluations based on measurements ofradionuclide concentrations in soil. The 2004 UNSCEAR
Global Survey on Public Radiation Exposures, which alsosought information on the numbers of people exposed, hasprovided information on the distribution of doses according
to specified ranges and on the average and range of radio­
nuclide concentrations in soil. Data on absorbed dose rates
in air for various countries, including data for high- and­low-background areas, are given in table 6.
77. Additional information on both external dose rates andradionuclide concentrations in soil is available in the recent
literature, as there has been expanded interest in mapping
countrywide exposures. Some data already collected andcomplementary to earlier reports [U3] are presented intable A-1, with average and maximum values for 238U, 232Th
and 40K concentrations in soil shown in figures V–VII. The
new data do not affect significantly the current worldwide
average values of 33 Bq/kg for 238U, 32 Bq/kg for 226Ra and45 Bq/kg for 232Th. The average value for 40K, 412 Bq/kg, isalso close to the previous value (420 Bq/kg). Although theaverage concentrations of natural radionuclides in soils are
low, there is a large variation, with reported levels of up to
1,000 Bq/kg for 238U, 360 Bq/kg for 232Th and 3,200 Bq/kg
for 40K. Therefore, for the purposes of global dose assessment, these data need to be linked with corresponding­population distributions.
78. The data on worldwide average outdoor dose rates presented in table 6 confirm the previous [U3] average value of58 nGy/h. The data available to date on the distribution of
the population with respect to the outdoor absorbed dose
rates in air due to terrestrial gamma radiation are presented
in table 7. The mean value for this distribution is in the range
50–59 nGy/h.
79. Indoor exposures depend on radionuclide concentrations in outdoor soil and in building materials. The relative
contribution from each source is highly dependent on thetype of house and building material. Information on distributions of indoor exposures derived from direct measurements
is not extensive, but these can be assessed on the basis of
information on soil, shielding and building material, andthen linked with the number of people exposed in order toestimate population exposures. Extensive information is
being gathered worldwide regarding activity concentrations
in building materials. New information, complementing thatin reference [U3], is given in table A-2. In general, average

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UNSCEAR 2008 REPORT: VOLUME I

values for natural radionuclides are higher in most building
materials than in soils, with granite and marble presenting
the highest average values for 226Ra (77 Bq/kg) and with
granite also presenting the highest average values for 232Th(84 Bq/kg) and 40K (1,200 Bq/kg).
80. Table 6 also confirms the previous value of 1.4 for theratio of indoor to outdoor exposure rates. Therefore the value
for the worldwide average indoor absorbed dose rate in air of
84 nGy/h given in reference [U3] is considered to be stillvalid. Using 0.7 Sv/Gy as the conversion coefficient fromabsorbed dose rate in air to the effective dose received byadults, and 0.8 for the indoor occupancy fraction, the average annual effective dose due to external exposure to natural
terrestrial sources of radiation is 0.48 mSv, with 0.41 mSv
related to indoor occupancy and 0.07 mSv to outdoor occupancy. The average levels for countries are mostly in therange 0.3–0.6 mSv.
81. Equation (3) is useful for calculating average outdoor
gamma ray exposure rates from global soil concentrations intable A-1. These average and standard error soil concentrations are: 40K: 400 ± 24 Bq/kg; 238U: 37 ± 4 Bq/kg; and232
Th: 33 ± 3 Bq/kg. The table 1 DCFsoil coefficients are
0.0417, 0.462 and 0.604 nGy/h per Bq/kg, for 40K, 238U and232
Th, respectively, and the calculated outdoor terrestrial
gamma ray exposure rate is estimated as 54 nGy/h. Using
0.7 Sv/Gy as the conversion coefficient from absorbed dose
rate in air to the effective dose received by adults, and 0.2 forthe outdoor occupancy fraction, the average annual effective
dose due to external exposure to natural terrestrial sources ofradiation is 0.066 mSv, in close agreement with the estimatedaverage based on absorbed dose rate measurements. For
indoor environments, the estimated dose rate is then
0.43 nGy/h. This can be taken as the contribution from thesoil material, and the difference between this value and theworldwide average value can mainly be attributed to the­contribution from building materials to indoor exposure.
82. Figure VIII shows the distribution of population withrespect to external dose rates outdoors for 38 countries.
From the left-hand figure, it can be seen that the largest
population fraction is in the 50–59 nGy/h range, confirming the previous estimates for external dose rate outdoors.
From the right-hand figure, it can be seen that about 90% of
the world population for which data have been provided forthis annex falls within the range of about 20 to over
100 nGy/h. The Committee has decided to revise the range
previously adopted for external dose rate (0.3–0.6 mSv/a)
to 0.3–1.0 mSv/a.
(b)

Internal exposures due to radionuclides other than radon

83. Internal exposures arise from the intake of terrestrial
radionuclides by inhalation and ingestion. Doses due toinhalation result from the presence in air of dust particles
containing radionuclides of the 238U and 232Th decay chains.
The dominant components of exposure due to inhalation are

the short-lived decay products of radon, which because of
their significance were considered separately in annex E of
the UNSCEAR 2006 Report [U1].
84. The inhalation of natural radionuclides other thanradon and its decay products makes only a minor contribution to internal exposure. These radionuclides are present in
air because of the resuspension of soil particles. The decay
products of radon are present because of radon gas in air.
Assuming a dust loading of 50 μg/m3 and 238U and 232Th concentrations in soil of 25–50 Bq/kg, the concentrations in air
would be expected to be 1–2 μBq/m3, and this is generally
what is observed. There is, however, a large variability associated with this value, as local levels may be affected by several factors, such as climate, soil class and concentrations insoil. Other factors affecting the variability of natural radionuclide concentrations in air are the contribution to the dust
loading of air from burning fuels, because, while organic
content is usually deficient in uranium compared with soil,
fly ash contains much higher concentrations of uranium. In
addition, at coastal locations, concentrations of uranium in
air may be an order of magnitude lower than in continental
or industrialized areas inland.
85. In the UNSCEAR 1993 Report [U6], representativevalues of the concentrations of terrestrial radionuclides in air
were selected. Because the database has changed very little,
most of those values are still considered valid. The highest
concentration, 500 µBq/m3, is for 210Pb. The concentrationsof the other radionuclides are: 50 µBq/m3 for 210Po; 1 µBq/m3
for 238U, 226Ra, 228Ra and 228Th; 0.5 µBq/m3 for 232Th and230
Th; and 0.05 µBq/m3 for 235U. The age-weighted annual
effective dose due to the inhalation of radionuclides from theuranium and thorium series in air was estimated to be about
0.006 mSv [U3].
86. Doses from ingestion are mainly due to 40K and to the
U and 232Th series radionuclides present in foods anddrinking water. The ingestion of natural radionuclides
depends on the consumption rates of food and water and on
the radionuclide concentrations. Reference food consumption profiles were derived in the UNSCEAR 2000 Report[U3] and are summarized in table 8. Although the tabulated
values are in reasonable agreement with other assessments,
substantial uncertainties are implicit in their mode of derivation. Moreover, there are large deviations from this profilefor various parts of the world because of differences in­dietary habits (for example, milk consumption in Asia andleafy vegetable consumption in Africa are lower). The values
in table 8 are to be seen only as reference values; actual­values vary widely.
238

87. The concentrations of naturally occurring radionuclides in foods vary widely because of differences in the
background levels in soil, the climate and the agricultural
conditions that prevail. There are also differences in thetypes of local food included in categories such as vegetables, fruits and fish. It is therefore difficult to select reference values from the wide ranges of concentrations reported.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

The relevance of specific nuclides to the dose depends on
the soil composition, and the ratio of uranium to thorium
varies from place to place, as shown in figure IX, leading tolarge variations in the activity ratios between their daughters, e.g. the 226Ra/228Ra ratio. The soil type also affects theretention/mobility of radionuclides in soil and their availability to plants [F17]. The annual intakes of radionuclides
from the uranium and thorium series in various countries
have an approximately log-normal distribution for eachradionuclide and span an order of magnitude. The highest
concentrations are for 210Pb and 210Po, which have similar
distributions. The lowest concentrations are for 230Th and232
Th, which also have similar distributions, while 226Ra and238
U have ­intermediate concentrations [U3].
88. Because drinking water is important for the intake ofuranium and radium radionuclides, it is necessary to ascertain that this source of ingestion intake has been included indietary intake estimates. The radionuclide contents in natural
water and tap water have been reviewed; spring and mineralwater have also been of particular interest. Some new data
are available and are summarized in table A-3. Worldwide
there is a huge variability in concentrations of natural radionuclides in drinking water. Figure X shows the ranges citedby countries for uranium. There is a variation of about eightorders of magnitude among individual water samples. The
consequence of such variation is a high variability in the values for global per caput doses. Figure XI shows the distribution of average values for 238U given in table A-3, where
there is a variation of three orders of magnitude amongworldwide average values.
89. Several authors have emphasized the disequilibrium
between 234U and 238U in water. A survey of levels in naturalbottled water from northern Italy has shown ratios of234
U/238U concentrations ranging from 0.99 to 1.63 [R21]. A
survey of water from the Euphrates River showed ratios in
the range 0.75–3.11. A survey that included measurementsof tap and well water in the United States showed ratios in
the range 1.16–2.92. At one location, a value of 5.5 wasobserved; at another location, a ratio of 0.37 was observed
for spring water [F9]. Average ratios are of the order of 1.5,
which means that doses due to water ingestion for 234U are
underestimated if they are based on 238U measurements
alone ­assuming radioactive equilibrium.
90. Uranium is retained in the body primarily in theskele­ton. It has been found that the concentrations in various types of bone (vertebrae, rib and femur) are approximately similar but show a large variability among different
countries and different age groups [F9]. An earlier estimate was that 70% of the body content of 238U was in thebone. Assuming the reference concentration of 238U inbone to be 100 mBq/kg, this would correspond to 500 mBq
in the skeleton and 710 mBq in the whole body. The average concentration in soft tissues would then be 3 mBq/kg,
with higher concentrations measured in the lungs and kidneys. Reference values for concentrations in tissues arepresented in table 9. The distributions of measured values

235

in bone for radionuclides of the uranium and thorium
series are ­presented in figure XII [U3].
91. Following intake by ingestion and inhalation, thorium
is deposited mainly on bone surfaces, where it is retained
for long periods. Metabolism models assume that 70% of
the body content of thorium is retained in the skeleton.
From the reference concentrations given in table 9 andassuming the cortical and trabecular bone masses to be
4 kg and 1 kg, respectively, it may be estimated that the
body burdens are 210 mBq of 230Th and 70 mBq of 232Th.
The distributions of uranium and thorium concentrations inbone are typically log-normal within a country. The combined values for various countries have an approximately
log-normal distribution and extend over an order of magnitude, with the variability being caused primarily by differences in intake of the radionuclides in food and water. The
distributions for 238U and 230Th concentrations in bone aresimilar; somewhat lower concentrations are reported for
232
Th. As these data are limited, they remain to be confirmed
as truly representative.
92. Radium is retained primarily in bone, and concentrations have been measured in many countries. Lead also
accumulates in bone. By contrast, polonium is distributed
mainly in soft tissues. Even in the absence of direct intake,
both lead and polonium would still be present in the body
because of the decay of 226Ra, but direct dietary intake is of
the greatest importance in establishing the content in the
body. Early measurements showed the 210Pb/210Po concentration ratio to be 0.8 in bone, 0.5 in the lungs and generally
unity in other soft tissues. Some enhancement of 210Po in theliver and kidneys has also been observed. The presence of210
Pb and 210Po in tobacco greatly increases the intake ofthese radionuclides by smokers; the measured 210Po concentration in the lung parenchyma of smokers is about threetimes that of ­non-smokers.
93. The published measurements of 210Po in human tissue
were summarized and the averages reported by Fisenne[F9]. The total concentration in the organs and the annualorgan equivalent dose are shown in figure XIII. The various
measurements of 210Po in tissue were from Finland, Japan,
the Russian Federation, the United Kingdom and the UnitedStates. The published measurements in bone were reported
from the same countries and additionally from France,­Germany, New Zealand and Poland.
94. The annual effective dose due to radionuclides fromthe uranium and thorium series in tissue at the reference concentrations in the human body was evaluated in the
UNSCEAR 2000 Report [U3] as 0.12 mSv. Evaluation of
the internal doses due to ingestion of radionuclides from theuranium and thorium series was also reviewed in the
UNSCEAR 2000 Report [U3] using the reference values ofconcentrations in foods and worldwide average consumption
rates for infants, children and adults. For adults, the estimated annual dose is 0.120 mSv. These two results are inclose agreement. The main contributor to this dose is 210Po.

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95. Potassium is more or less uniformly distributed in the
body following intake in foods, and its concentration in the
body is under homeostatic control. For adults, the body content of potassium is about 0.18%, and for children, about
0.2%. With a natural abundance of 0.0117% for 40K in natural potassium, a specific activity for 40K of 2.6 × 108 Bq/kg
and a rounded dose conversion coefficient of 0.003 mSv/aper Bq/kg, the annual equivalent doses in tissues from 40K in
the body are 0.165 and 0.185 mSv for adults and children,
respectively. The same values are appropriate for the effective doses, given the more or less uniform distribution ofpotassium within the body.
96. The total annual effective dose due to inhalation andingestion of terrestrial radionuclides is assessed to be
0.29 mSv, of which 0.17 mSv is due to 40K and 0.12 mSv tothe long-lived radionuclides in the uranium and thorium
series.

99. Despite the lack of specific criteria, such areas are ofinterest mainly because they have been used to illustrate highchronic levels of radiation to which human beings are currently exposed and to consider the relevance of such exposures to epidemiological studies on the effects of low-doseand low-dose-rate exposures. Some of these areas are listed
in table 10. The origins of the higher exposures and the characteristic levels that define the area as an ENRA are
included.
100. The results of this preliminary literature review indicate that public exposure to natural radiation may be of special concern in ENRAs. However, most of the currently
available data fail to give the number of persons involved;
the information provided on “dose distributions” typicallyrelates only to the exposure fields and not to population.
Only three countries—the Czech Republic, the Islamic
Republic of Iran and Spain—had responded by April 2006
with information on the population dose distribution; theirdata for high-background areas are presented in table 11.

(c) Inhalation of radon
97. Exposure to radon has been described in annex E of the
UNSCEAR 2006 Report [U1]. The Committee has decided
to keep its previous estimates of 1.15 mSv and 0.1 mSv forthe average annual per caput effective doses due to natural
sources of radon and thoron, respectively [U3]. This represented approximately one half of the estimated dose due toall natural sources of ionizing radiation. Combining the data
presented in table 1 of annex E of the UNSCEAR 2006
Report [U1] with recently updated information available
from the European Commission [D14], the distribution of
average radon concentration indoors among countries isshown in figure XIV. The average values for individual countries ranged from 9 to 184 Bq/m3. The currently available
data fit a log-normal distribution (r = 0.98) with a geometricmean of 45 Bq/m3 (similar to the previous estimated value of
40 Bq/m3) with a geometric standard deviation of 2.1.
(d)Areas with elevated radiation levels due to natural
sources98. Several areas of the world are known to have levels of
exposure due to natural sources of radiation that are inexcess of those considered to be “normal background”.
There is no specific value of dose rate or of activity concentration in the environment that defines what constitutes an“enhanced natural radiation area” (ENRA). Some references cite criteria such as a dose rate of greater than
300 nGy/h or an indoor 222Rn concentration in air of theorder of 150 Bq/m3. However, these are not adequate reference levels, because situations exist in which those levels
are clearly not applicable (for example in areas with highlevels of exposure to cosmic radiation; areas where theexposure is due to high levels of 226Ra and/or 222Rn in water,
often called “dynamic ENRAs”; or areas where the totaldose, including external and internal exposures, is higher
than the usual range).

3.

Summary on exposures to natural radiation sources

101. Although it is recognized that a large effort has been
made to map natural radiation sources (mainly radon, themost relevant radionuclide), the available information cannot be correlated with other exposure pathways for whichdata are not yet presented in such a degree of detail. The
countrywide radon maps already available for most European countries [D14] and for Costa Rica [M25, M30] havebeen provided to UNSCEAR. In addition, distributions ofexternal dose rates are available for some countries, and for
the United States, the distributions of uranium, thorium andpotassium are available on countrywide maps [U28]. Knowing the cumulative exposure to different sources on a geographical basis could change the current exposure assessment
and lead to more precise estimates of the distribution of
exposures worldwide. This aspect will be further discussed
in the conclusion of section II.E of this annex. The new
information available does not currently allow estimates tobe made to characterize worldwide average exposures tonatural radiation that are significantly more accurate thanthose provided in previous reports. It was therefore decided
to maintain the same numerical values but to slightly extend
some ranges (see table 12).
102. The values in table 12 are to be seen as “average”v­ alues, but it should be kept in mind that the worldwide
exposure to each pathway usually follows a log-normal distribution. Therefore they should be seen only as reference
values and not as specific to any particular place. In fact, assome exposure pathways are correlated with each other, theactual distribution may vary significantly among different
places.
103. Besides the large variability in environmental concentrations and in population habits throughout the world, therate at which dose is accumulated may also vary as theANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

individual ages. A study performed in the United Kingdom
found that inhalation doses for infants and children arewithin 20% of those for an adult, while terrestrial gamma
rays give effective doses for infants and children that arelarger than those for adults by about 30% and 15%, respectively. The variation of ingestion doses between individuals
is comparable to that of doses from terrestrial gamma rays
[K8].
104. Regarding public exposure during aircraft flights,
although the estimated doses received by passengers during
individual flights are low, collective doses may be quite highbecause of the huge number of flights worldwide. In addition, doses to specific individuals who fly frequently may
make an appreciable contribution to their overall exposure tonatural sources.
B.

Enhanced sources of naturally
occurring radioactive material

105. Activities related to the extraction and processing ofores can lead to enhanced levels of naturally occurring radioactive material (NORM) in products, by-products and wastes.
An assessment of the situation related to sites with technologically enhanced levels of NORM has been performed incountries of the European Union [V4]. Nine important categories were identified. This annex uses a similar approachand discusses the disposal or use of waste within the category that generates the waste. Eight of the categories are
addressed here: uranium mining and milling; metal mining
and smelting; phosphate industry; coal mines and power
generation from coal; oil and gas drilling; rare earth and titanium oxide industries; zirconium and ceramic industries;
and applications using natural radionuclides (typicallyradium and thorium). The ninth category (disposal of building material, which is recognized to be of little concern) is
not considered here.
106. At least for Europe, the first three categories represent the major contaminating industries with respect to theoverall amount of waste produced, though radionuclide­levels in products and/or waste from the second three cate­gories may be particularly elevated [V4]. Apart from uranium mining and milling, applications using naturalradionuclides and, more recently, zirconium industries,
activities related to the other categories have generally not
been fully evaluated from the perspective of public exposure, though attempts to characterize them according to
the radionuclide content of materials have been made in­previous UNSCEAR reports [U3, U6].
107. For past industries, the main concern is related to thesites where residues were left before present standards ofradiological protection were established. Many of these sites
have already been cleaned up, and residual doses and/or
radionuclide contents are known. For industries currently inoperation, the main focus relates to effluents, releases fromwaste and the relevant exposed groups of the population.

237

Descriptions of environmental liabilities (such as waste rockpiles, waste basins and contaminated areas) can be a valuable
starting point for a database that can be used for futureassessments of exposure and dose. The features of uranium
mining and milling and the related exposures are described
below, together with other fuel cycle exposures, in the section on exposures due to nuclear power production, section
III.C.1 of this annex.
1.

Metal mining and smelting

108. The metals considered include aluminium, copper,
iron (and steel), lead, niobium, tin, zinc, gold and others. The
NORM activity in feed material for metal smelting is generally low, and the same is true for most slags and other waste.
The concentration of radionuclides in intermediary products
and wastes, however, will depend on the content initiallypresent in the ore and on the type of process used to extract
the metal. In the case of thermal processes, a large part of theradionuclide content will be concentrated in metallic slags,as, for example, in those from the tin industry [V6].
109. The activity levels in the niobium industry may behigh, with pyrochlore containing 10,000–80,000 Bq/kg of232
Th [V4]. In one niobium facility in Brazil, activity levels
in waste ranged up to 200,000 Bq/kg of 228Ra (in barium sulphate) and 117,000 Bq/kg of 232Th (in the slag). Exposure of
the public due to feedstock or the metal products is notexpected. The main pathways for public exposure include
contamination of groundwater with radium isotopes andexternal exposure to slag with high thorium content (if this is
not disposed of in an acceptable manner) [I22, P11]. Exposures due to inhalation of resuspended material from tin andniobium slag used as landfill have also been cited [V4].
110. In South Africa, the gold deposits from deep underground mines have low-grade uranium associated with them.
Since 1952, 170,000 t of U3O8 have been recovered as a byproduct of gold mining. Some 6 billion tonnes of mining
tailings, containing about 500,000 t of uranium and 200 kg
of 226Ra, have been deposited. New tailings are being deposited at a rate of 86 million tonnes annually. Elevated 226Ra
concentrations, up to 1.7 Bq/L, have been observed in the
discharges. Annual doses to nearby populations have been
estimated as up to 0.04 mSv due to the ingestion of waterand up to 0.086 mSv due to the ingestion of fish. Annual
doses due to ingestion of land food products are much lower,
ranging up to about 0.002 mSv. Annual doses to the public
due to the inhalation of radon and of dust from the tailings
piles have been estimated to be about 0.04 and 0.02 mSv,
respectively [W18].
2.

Phosphate industry

111. Phosphate rock is used extensively, firstly as a sourceof phosphorous for fertilizers and secondly for making
phosphoric acid and gypsum. Ores typically contain about

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UNSCEAR 2008 REPORT: VOLUME I

1,500 Bq/kg of uranium and radium, although some phosphate rocks contain up to 20,000 Bq/kg of U3O8 [P7]. In
general, phosphate ores of sedimentary origin have higher
concentrations of nuclides of the uranium family. The
magmatitic minerals, such as those from Kola (RussianFederation) and Phalaborwa (South Africa), present lower
concentrations of nuclides of the uranium family and higher
concentrations of nuclides from the thorium family, although
the total activity is lower than that from sedimentary minerals [V6]. In 90% of cases, the ore is treated with sulphuric
acid. The fertilizers become somewhat enriched in uranium(up to 150% relative to the ore), while 80% of the 226Ra,
30% of the 232Th and 5% of the uranium are left in thephosphogypsum.
112. The processing of phosphoric rocks may generate
gaseous and particulate emissions that contain 238U and 226Ra;
when discharged to the environment, these nuclides lead toradiation exposure of the population. Local dump sites forphosphogypsum are usually not protected from rainfall andbecome hydraulically connected to surface waters and shallow aquifers [V4]. The use of phosphate fertilizers in agriculture and of gypsum in building materials is a further
source of possible exposure of the public [P7]. Elevatedradon exposure of the public can further be expected in sites
being developed for housing [V4].
113. For somewhat more than half a century, phosphateores of marine origin containing 226Ra have been processedin Belgium to produce calcium phosphate for use in cattle
fodder. The wastewaters are discharged into two small rivers, one of which is the Laak. Enhanced concentrations of226
Ra are observed along the riverbank, mostly confined to a
10 m strip along both sides of the river, including flooding
zones. As of 1999, no dwellings had been built on top ofthese higher-activity areas and no crops for direct humanconsumption were grown there, so no immediate threat tothe population existed [P6].
114. Prior to 1990, France discharged about 3 million
tonnes of phosphogypsum into the Baie de la Seine. After
1990, waste was stored on land. In the United Kingdom,
the annual discharge of 210Po exceeded 0.5 TBq in the
period 1980–1983. In 1993, about 10 million tonnes ofphosphogypsum waste were generated within the European
Union, with 15% being recycled (for example as building
materials), 25% discharged to sea and 60% stored on land
[E16]. The import of phosphate ore to European Union
countries decreased by about a factor of 2 between 1985
and 1992, reflecting an increasing tendency to import phosphoric acid directly rather than import the ore itself. This
reduced the disposal of uranium to sea, bringing about alarge decrease in environmental concentrations of 210Po, but
in the process transferring the waste disposal problem backto the ore-­producing countries such as Morocco [E13].
115. Phosphate rock can be melted in a furnace at hightemperature with sand, iron oxide and coal for the production of elemental phosphorus. The residual solids in the

furnace contain ferrophosphorus and calcium silicate, also
known as slag [I22]. The slag, which contains 226Ra concentrations ranging from 750 to 1,100 Bq/kg, has been used as
construction material in the United States, specifically incommunities in south-eastern Idaho. Surveys for external
exposure were conducted in 1,472 residences. It was estimated that fewer than 12% of the residences in Soda Springcontained slag, while in Pocatello and Fort Hall no houses
were found containing the slag. The highest individual dose
rate was estimated to be 1.3 mSv/a, and only nine individuals were identified as receiving more than 1 mSv/a above
background. A significant fraction of the public roads, however, contained slag: 27% in Pocatello, 23% in Soda Springand 20% in Fort Hall [A13].
3.

Coal mining and power production from coal116. The average specific activity of both 238U and 232Th incoal is generally around 20 Bq/kg (range 5–300 Bq/kg). Coal
mines in Freital, Germany, which have uranium concentrations of 15,000 Bq/kg coal, are an exception [V4]. During
the burning of coal, the organic compounds are converted
into gases (water vapour and carbon dioxide), while the inorganic elements, which include the significant naturally
occurring radionuclides, are concentrated in the ashes [V6].
In general, the radionuclide enhancement factor in ash isabout 10. Leaching from fly ash is low, and therefore there
are few restrictions on the use of fly ash in landfill and road
construction. The use of fly ash for building construction,
however, results in radiation exposure from both direct irradiation and radon exhalation. Dumping fly ash may increase
the radiation level around the dump site. The most significant exposure pathways are ingestion and inhalation of theisotopes 210Pb and 210Po [V4]. However, recent studies in the
United Kingdom confirm earlier indications that the incorporation of pulverized fuel ash into building materials is
unlikely to contravene either current national legislation or
the European Union directive [H17, H18].
117. The content of natural uranium in coal from Brazil
ranges from 30 to 2,000 parts per million. It is estimated thatthe burning of 2.2 × 106 t of coal per year discharges about
270 t of U3O8 equivalent into the environment [P7].
118. About 50 underground coal mines are located in theUpper Silesian Coal Basin, in the southern part of Poland.
The total water outflow from these mines is about
800,000 m3/d. Waters with high radium content (up to
390,000 Bq/m3) are found mainly in the southern and central parts of the basin where a thick layer of impermeableclay overlies the coal seams. Radium-bearing waters fromcoal mining are discharged into surface settling ponds andlater into rivers. In some cases, radium isotopes are co-­precipitated with barium in these ponds or are absorbed onbottom ­sediments [C7, W19].
119. Slags derived from coal mined in the vicinity of thetown of Tatabánya in Hungary have elevated concentrations

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

of 226Ra (850–2,400 Bq/kg). The slag has been used as fillingand insulating material for building houses, blocks of flats,schools and kindergartens, and to fill playgrounds and roads[N13].
4.

Oil and gas drilling

120. During the extraction of oil and natural gas, the­natural radionuclides from underground formations arebrought to the surface. Elevated activities of 226Ra and228
Ra present in NORM are often released by oil and gas
industries, particularly in production waters. During theextraction process, radium is co-precipitated along withbarium and strontium. A portion of the radium is deposited
during the scale formation process and another portion isdischarged to the sea with effluents. Mean concentrations
in wastewater are 2 and 2.3 Bq/L for 226Ra and 228Ra,
respectively. Although these high activities of radium arepresent in production water for some platforms, water andsediments sampled at a distance of more than 250 m fromthe production site had normal background levels, showing that water mixing sufficed to reduce concentrations in
the environment [J2, V6].
121. The most important radionuclides in scales and otherprecipitates are the isotopes of radium, with specific activities
ranging from 100 to 1,000 Bq/kg. Activity concentrations insludges are typically a factor of 100 lower. Concentrations of210
Pb and 210Po in sludge and scales can vary between 20 and
1,000 Bq/kg [V4]. The sludges on the Bacia de Campos oil
platforms in Brazil have about 105,000 Bq/kg (maximum
340,000) of 226Ra and 78,000 Bq/kg (maximum 286,000) of228
Ra [M14].
122. Activity levels in scales are of the same order as thosein uranium mill tailings and other materials that are regulated because of their potential for 222Rn release. The 222Rn
emanation fraction for pipe scale, however, is generally
lower than that for typical mill tailings [W10]. The disposalof scale from oil extraction industry installations and of
sludge containing NORM can be of environmental significance, with contamination of land being the major concern.
The average radium concentrations in soils sampled at an
oilfield contaminated with NORM in eastern Kentucky,
United States, were 32,560 ± 340 Bq/kg [R2].
123. Tank battery sites, which separate water and sedimentfrom the oil produced, have historically been used for the
initial processing of crude oil. The sediment remaining in thepit is an oily, viscous material often called “sludge”. Thissludge can be radioactive if NORM is associated with thematrix. A radiological survey conducted on six previously
remediated tank battery sites revealed average gamma radiation exposure rates ranging from 27 to 100 µGy/h [H19]. In
older scales, the concentrations of 228Th will have increased
because of ingrowth. Scales and sludges, particularly thosefrom gas fields, may also contain relatively high levels of210
Pb and 210Po [E13].

239

124. Waterborne pathways may make a noticeable contribution to the radiation exposure of persons resident on farmland contaminated with residual NORM arising from crude
oil recovery operations. Persons living in such areas would
incur external gamma exposure and exposure from radon
inhalation [R2]. The exposure from dissolution of 226Ra isincreased in cases where contaminated soil is located nearseawater [A9].
5.

Rare earth and titanium oxide industries

125. Bastnaesite and monazite are the most important minerals containing rare earth metals. Bastnaesite has an activity
concentration of 900–1,200 Bq/kg for radionuclides in the238
U decay series and 700–7,000 Bq/kg for radionuclides in
the 232Th decay series. Monazite, on the other hand, has anactivity concentration of 10,000–50,000 Bq/kg for radio­
nuclides in the 238U series and 5,000–350,000 Bq/kg forradionuclides in the 232Th series. In Europe, minimal amountsof waste are produced by these industries [I22, V4].
126. The Brazilian experience is somewhat different. As aconsequence of monazite processing for the production ofrare earth chlorides, carried out from 1949 to 1992, basically three different kinds of waste were produced: (a) thelight-mineral fraction (activity concentration 170–320 Bq/g)
from the monazite physical purification; (b) “cake II” (average content 20% thorium hydroxide and 1% uraniumhydroxides, approximate activity concentration 1,820 Bq/g)
from the monazite alkaline digestion; and (c) mesothorium
cake (Ba(Ra)SO4) (approximate activity concentration
4,360 Bq/g). It is estimated that about 3 × 104 t of cake II
and 1 × 105 t of mesothorium cake were produced annually.
These wastes and residues were disposed in shallow groundsilos or in rubber drums, or were buried in trenches. Areas
that used the light-mineral fraction as landfill later had to be
decontaminated [L1].
127. Similar situations occurred in the United States, with
waste originating from a Rare Earths Facility that operated
from 1932 until 1973 to produce rare earths and radioactive
elements such as thorium, radium and uranium using anacid leaching process. Production of these elements generated radioactive mill tailings that contained residual levelsof thorium, radium and uranium. Over several decades, themill tailings were available for use as landfill material byresidents and contractors. Winds also may have spread someof the mill tailings to nearby neighbourhoods. Clean-up
actions were performed in the mid-1980s for approximately
120 residential properties in the West Chicago area in Illinois, and later for more than 2,170 properties (covering
approximately 400 hectares (1 ha = 10,000 m2) in andaround West Chicago [E5].
128. For titanium production, activity concentrations in
the ore are about 300–600 Bq/kg for the 238U decay series
and 35–600 Bq/kg for the 232Th series. Specific activities
of radium sulphate precipitates in pigments or scales may

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UNSCEAR 2008 REPORT: VOLUME I

be as high as 400,000 Bq/kg, and 228Th levels may behigher than 1 × 106 Bq/kg. Exposure pathways include
external ­irradiation and migration of radionuclides fromlandfill [V4].
129. Scales formed during titanium dioxide pigment production have 238U series activity concentrations ranging from <102
to 1.65 × 106 Bq/kg and 232Th series activity concentrations
ranging from 4 × 103 to 2 × 106 Bq/kg (the maximum value
could apply equally to 228Ra or 228Th). However, the ­pigments
themselves are essentially free of radioactivity [I22].
130. The use of the ensuing waste can also lead to public
exposure. During the production of titanium dioxide, mostnaturally occurring radionuclides originally present in theore are precipitated as metallic hydroxides, except for radiumisotopes (the radium chlorides remain partially soluble andare discarded with wastewaters) [V6]. The processing ofmonazite in France for rare earth extraction, beginning in
1976, led to the input of significant quantities of 232Th and228
Ra to La Rochelle Bay, within authorized annual limits of
37 GBq and 74 GBq, respectively. Improved waste treatment
beginning in 1990 reduced the annual discharges to about
0.5 GBq of 232Th and 6 GBq of 228Ra [E13].
6. Zirconium and ceramics industries
131. The average activity concentrations in zircon and zirconia, respectively, are 600 and 300 Bq/kg for 232Th, and
3,000 and 7,000 Bq/kg for 238U. Except for refractory bricks,
where 238U activity concentrations of 10 Bq/kg have beenreported, the activity concentrations in the products are comparable to those in the feed material. Long-lived radioactive
dust constitutes the main source of radiation exposure, which
is mainly due to thorium in the dust [V4]. The zirconium
industry and the industrial uses of zirconium may be a sourceof occupational exposure; only the reuse of solid waste could
possibly lead to public exposure [V6]. Doses from gamma
radiation emitted from large stockpiles of zircon sand aremainly an issue for workers, but in principle, individuals outside a zircon milling plant may also receive exposure via thispathway if they are sufficiently close to the facility. The critical group would be individuals working in the industrial areasurrounding the plant, with a maximum conservative annual
effective dose estimated to be about 200 µSv. Individuals
may also receive exposure from material deposited outside
the plant by storm water runoff and from the inhalation of
airborne dust emitted from stockpiles and openings in theplant buildings. In studies from several countries, applying
conservative approaches, the maximum annual effective
dose received by an individual outside the facility was estimated to be less than 1 µSv from discharges to water and56 µSv from emissions to atmosphere. In nearby population
centres, the dose was found to be negligible [I41].
132. The activity concentrations of uranium and thorium
series radionuclides in spent foundry sands or waste are likely
to be of the order of 1,000 Bq/kg or less because of thedilution of zircon with other constituents. It is expected that anannual effective dose of the order of 100 µSv is the maximumthat could be received by a member of the public as a result of
the disposal of these materials in landfill facilities. For the
manufacture of zirconia by fusion of zirconium ­minerals, themain exposure pathways to members of the public are discharges of radionuclides in liquid effluent (floor washings)
and stack emissions, and the migration of radionuclides fromthe landfill disposal of furnace dust. Concentrations of 210Pb of
up to 200,000 Bq/kg and 210Po of up to 600,000 Bq/kg in furnace dusts have been found. The maximum dose received bya nearby resident from the release of radionuclides in liquid
effluents is negligible. The dose received as a result of plume
inhalation and exposure to material deposited from stackemissions was estimated as 37 µSv, of which over 35 µSv wasdue to dust inhalation. The dose received by a future site user
after closure of a landfill facility containing 50 t of furnace
silica dust (excluding the dose from indoor radon, for which
no realistic estimate was made) was 4.5 µSv, of which 3.8 µSv
was due to external gamma exposure.
133. For the manufacture of zirconium compounds bychemical dissolution of zirconium minerals, the main potential exposure pathways to members of the public are those
associated with the landfill disposal of pipe scales and silicacontaining residues. Chemical processing can produce scales
and other residues with radium (226Ra + 228Ra) concentrations of up to a few thousand kilobecquerels per kilogram. A
future resident living after closure on a landfill site into
which 20,000 t of solid residue had been disposed was estimated to receive a dose of 750 µSv/a, mostly from external
gamma radiation. For the chlorination of zircon and the production of zirconium metal, the sludge from the zirconium–
hafnium separation process, owing to its radium content andlarge volume, gives rise to radiological issues similar to
those associated with radium-rich mine tailings. Consequently, sludge stockpiled in ponds and piles represents a
potential source of public exposure through the migration ofradionuclides into the surrounding environment, particularly
if the sludge is stored long-term rather than being used elsewhere, for example as a soil conditioner. Although there are
obvious benefits in using sludge as a soil conditioner rather
than storing it indefinitely in piles, there are radiologicalimplications associated with the use of sludge in this manner. If the 226Ra activity concentration in the sludge is of theorder of about 1,000 Bq/kg, this corresponds to a radon flux
density per unit 226Ra activity concentration similar to that ofnormal rocks and soil. Sludge deposited on agricultural fields
has been found to give rise to a gamma dose rate of
0.1–1 µSv/h at a height of 1 m and to a radon flux density of
0.44 Bq m–2 s–1.
134. Products from the zircon industry, such as ceramictiles and sanitary ware, have activity concentrations far
below 1 Bq/g and would not normally be regarded as giving
rise to exposures of concern. However, since these products
are essentially building materials, some consideration of
their radiological impact on members of the public is­warranted. The potential exposure pathways are throughANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

external gamma radiation and inhalation of radon released
from the product. Several studies in different countries havefound doses attributable to the use of glazed tiles in dwellings to be in the range 19–113 µSv above background. Whiteor near-white porcelain tiles have a higher zircon content
than glazed tiles and would therefore be expected to give rise
to correspondingly higher doses. The use of porcelain tiles
containing about 13% zircon in residences may give rise todoses of up to 120 µSv. The zircon content of glazes applied
to sanitary ware is similar to that of glazes applied to ceramictiles, but since the surface area of sanitary ware glaze in a
typical home is far smaller, the radiological impact of thezircon used in the glazes applied to sanitary ware is verysmall compared with that of ceramic tiles. For refractories,
the only potentially significant source of public exposure is
the burial of spent refractories at a landfill disposal site. Calculations show that the annual effective dose received by amember of the public from the disposal of furnace lining
bricks and refractory nozzles in a landfill, including the dose
received as a result of future, uncontrolled residential use of
the site, is likely to be no more than a few microsieverts.
There are no significant public exposure pathways for theuse of zircon as a source of zirconia in glass. For the use of
fused zirconia in other applications, the disposal of rejectmaterial at a landfill facility is not likely to lead to any significant migration of radionuclides into the surrounding
environment. The production processes of zircon ceramictiles, sanitary ware, ceramic pigments and abrasives do notgive rise to any significant exposure pathways to members of
the public [I41].

137. Also close to the site is an area that had previously
been lower than its surroundings. The difference in level
was removed between 1955 and 1960 by depositing residues of cobalt production, the debris of a building formerly
used for radium production and a limited amount of radium
extraction residues. The area is 9–10 hectares in size andcontains mixed radium and chemical waste to a depth of
3 m. No direct public exposure occurred, because a security
fence surrounded the area. Material in the dump contained
radium with concentrations of up to 34,000,000 Bq/kg.
Some nine or ten stretches of road and several isolated
points were found to contain contaminated pavement to adepth of about 0.3 m. About 5% of the 11,000 dose ratemeasurements performed had values of greater than
0.2 µSv/h. One dwelling (with contaminated material underthe veranda) had an average radon concentration in air of
720 Bq/m3 on the veranda and 370 Bq/m3 in the living room.
Radon measurements were performed in 846 dwellings;
only six showed average radon concentrations in air thatwere greater than 150 Bq/m3 [V5].
138. Thorium is extracted from the same minerals used forrare earth extraction. Specific activities of feed material arein the range 103–104 Bq/kg. Thorium has been used in a large
number of products and processes. Levels in the products(gas mantles, glass and tungsten) are typically higher thanthose in the original ore by a factor of 100 [V4]. Discarding
industrial waste and gas mantles may require particularattention in order to avoid public exposure [V6].
8.

7.

241

Other exposure situations

Applications of radium and thorium

135. Radium has been extracted from uranium-rich ores.
High contamination levels were recorded in soil surrounding
a luminizing facility in London, with 226Ra levels of between
0.4 and 400,000 Bq/kg, and with levels for “hot spots” of up
to 4,000,000 Bq/kg. Similar concentrations were found in
the vicinity of a watch factory at Dieppe, France. Exposures
to the public are mainly due to external exposure and radon
inhalation [V4].
136. An extensive radiological survey identified several
contaminated areas in the vicinity of the former Olen radiumfacility in Belgium. The major contaminated site was the
Bankloop brook, whose bed and banks were contaminatedover a distance of 1,400 m with radium and chemical waste(heavy metals) to a depth of up to 1 m. The contamination
was mainly confined to a narrow strip 5–10 m wide on one
or both sides of the brook. About 64% of the total volume ofcontaminated soil and sediments, which had an associatedexternal dose rate of over 0.15 µSv/h, was in a residential
area. At the mouth of the Bankloop, about 3 hectares offarmland (a former area of flooding) were found to be contaminated with radium up to a depth of 1 m (from deep
ploughing). The area is used for pasture. The average dose
rates are about 0.3 µSv/h, and the maximum value measured
was 5.5 µSv/h.

139. From 1994 to 1999, there were 53 instances where
evidence of radioactivity in ferrous scraps was discovered bysteel companies in Taiwan, China. These involved 15 orphan
radioactive sources, 16 60Co-contaminated rebars, 20 NORMcontaminated scraps and 2 cases whose cause was unknown.
For the NORM, five possible industrial processes may havebeen involved: oil production and treatment; heavy mineralsand processing and rare earth processing; copper mining
and processing; recovery of ammonium chloride by lime
absorption in the ammonium–soda process; and uranium
enrichment processes and tailings [C9, C10].
140. At least eight heavily used streets (approximately
3–5% of all civic road surfaces in the downtown area of­Tayoyuan City, Taiwan, China) were found to exhibit un­usuallevels of radiation. Crushed rock debris and coarse sands
separated from the asphalt pavement were identified as thesource. The activity concentrations of 232Th and 238U were
found to range up to about 4,000 and 1,000 Bq/kg, respectively. The dose rate on the road surface reached about
1.3 µSv/h, compared with the usual background level of
0.08 µSv/h on Taiwan [C8].
141. In the town of Monte Alegre, Brazil, an urban area wasconstructed using stones taken from a nearby uranium anomaly as landfill. The urban area has about 20,000 inhabitants,

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and 222Rn concentrations in air indoors are in the range
9–310 Bq/m3, with an average of about 75 Bq/m3. A small
rural settlement of 3,000 people close to the anomaly showsindoor radon concentrations in air in the range 35–462 Bq/m3,
with a mean value of 116 Bq/m3 [B27, M21].
142. There has been some concern about the exposure due
to waste arising from water treatment. All natural waters
contain certain concentrations of naturally occurring radionuclides. These may be enriched in the waste (mainly filter
sludge), and the handling, transport and disposal of thiswaste may cause radiation exposure of operating personnel
and of the public. A study performed in Europe concluded
that, while the exposure of operating personnel due to direct
gamma radiation and the exposure of the driver and the public during the transport and unloading of waste are of no concern, there are two exposure pathways that do need to beconsidered. The first is the exposure of operating personnel
to radon. The dose due to inhalation will be highly dependent on both the radon content in the water and the ventilationof rooms [H23]. An analysis of raw water samples in Germany indicated a median value of 5.9 Bq/L of 222Rn, with
only about 1% of samples having concentrations of greaterthan 500 Bq/L of 222Rn. The “activity transfer factors”
reported for radon are about 50 Bq/m3 and 0.1 Bq/m3 for
1 Bq/L in water for unventilated and ventilated rooms,
respectively. The annual doses for a worker working
2,000 hours in a year in such areas, assuming a 222Rn concentration of 500 Bq/L in water, would be 155 mSv and 0.3 mSv
for unventilated and ventilated rooms, respectively. The
annual doses corresponding to the geometric mean concentration of 5.9 Bq/L of 222Rn in water would be 2 mSv and0.004 mSv for unventilated and ventilated rooms, respectively. The second pathway that may deserve attention relates
to the use of waste sludges as a fertilizer on arable land.
Using very conservative approaches, the estimated annual
doses range from 0.02 mSv to 2 mSv for adults, dependingon the origin of the water generating the sludge, with thedose for infants being about one order of magnitude higher
than that for adults [H23].
143. A similar analysis was performed in the United Kingdom. Exposure scenarios relating to the treatment of tap andmineral waters include the transport, the unloading and theuse of the sludges on arable land as a fertilizer. For transporting sludges resulting from the treatment of mineral waterwith high measured radon content, the annual dose to amember of the public was conservatively estimated as
8 × 10–3 µSv. The corresponding value for sludges resulting
from tap water treatment is also negligible. The dose resulting from a single unloading event was found not to exceed
10 µSv for any type of sludge and for any exposure group,
even using very conservative approaches [H23].
144. Sludges from tap water treatment can be directly usedin agriculture, as fertilizers, while sludges from the treatment of mineral water are fed into a sewage plant, where
they are diluted with sludges of other origin, reducing thefinal radionuclide concentration. Land contamination due tothe spread of sludge will depend on the radionuclide concentration in the sludge and on the thickness of the sludge
layer on the land. The use of the land for agricultural production can give rise to public exposure via the ingestion
pathway. Annual doses in the United Kingdom due to theuse of sludges from water treatment as fertilizers were estimated to be in the range 0.01–0.3 mSv for sludges frommineral water and 0.02–33 mSv for sludges from tap water
treatment [H23].
145. There are also several sites with residues from former
installations around the world. Most of these sites are contaminated with radium from former luminizing industries.
Some European countries, such as the United Kingdom andBelgium, as well as the United States and Canada, have suchcontaminated sites. However, these sites have already been
identified and most of them have already been remediated,so that the current levels of public exposure are very low.
9.

Summary on exposure to enhanced NORM

146. Several types of facility worldwide that are not related
to the use of nuclear energy may give rise to exposures ofmembers of the public from enhanced concentrations of naturally occurring radionuclides in industrial products, byproducts and wastes. A large effort is under way at both thenational and the international level to assess exposure to
NORM and to develop strategies to address existing situations that give rise to exposure [E16, I22]. Table 13 presents
a summary of the dose estimates for members of the publicin the United Kingdom due to the release of NORM fromsome typical industries [W6]. Besides these, NORM canalso expose people as a result of several common practices,such as the agricultural use of sludges from water treatment,or the use of residues as landfill or building material.
Although doses to the public are usually low, of the order ofa few microsieverts or less, some critical groups could
receive doses in the millisievert range, which may deserve
attention. The Committee encourages the further development of inventories and methodologies for dose assessment
in order to have a more comprehensive view of the issue inthe context of public exposure.
C. Use of man-made sources for peaceful purposes
1.

Nuclear power production

147. The Committee has routinely collected data on
releases of radionuclides due to the operation of nuclear
fuel cycle installations. The UNSCEAR 1993 Report [U6]provided an overview of annual releases of radionuclides
for each of the basic types of reactor and other fuel cycle
installations since the practice of commercial nuclear
power generation began. Data for individual mines, mills,
reactors and reprocessing plants were provided for theyears 1985–1989. In the UNSCEAR 2000 Report [U3], thedata for an additional period, 1990–1997, were assessed.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

The present annex provides additional operational data forthe period 1998–2002 for nuclear power reactors and forthe period 1998–2003 for uranium mining.
148. The generation of electrical energy by nuclear means
has grown steadily ever since it started in 1956. The relatively rapid expansion that occurred from 1970 to 1985, anaverage increase in energy generation of over 20% per year,
slowed to a pace averaging just over 2% per year from 1990
to 1995. Although there has been an increase in the decommissioning and the shutdown of nuclear reactors, nuclear
energy production is still growing, although with lower rates
of increase in generated energy: about 0.2% from 1996 to2000 and about 0.1% from 2000 to 2005. In addition, thenumber of countries using nuclear power has increased [I27,
I28, I31].
149. The nuclear fuel cycle includes: mining and milling of
uranium ore and its conversion to nuclear fuel material; fabrication of fuel elements; production of energy in the nuclear
reactor; disposal of irradiated fuel or its reprocessing, withrecycling of the fissile and useful materials recovered; andstorage, release, treatment and disposal of radioactive waste.
For some types of reactor, enrichment of the isotopic content
of 235U in the fuel material is an additional step in the fuelcycle. The nuclear fuel cycle also includes the transport ofradioactive material between the various installations.
150. Radiation exposures of members of the public resulting from discharges of radioactive material from installations of the nuclear fuel cycle were assessed in previous
UNSCEAR reports [U3, U6, U7]. In this annex, the trends innormalized releases and the resultant doses due to nuclear
power reactor operation are presented for the years 1998–
2002. Doses are estimated using the environmental and dosimetric models described in annex A, “Dose assessment
methodologies”, of the UNSCEAR 2000 Report [U3].
151. The doses to exposed individuals vary widely from
one installation to another, between different locations, withdifferent population habits and with time. Generally the individual doses decrease markedly with distance from a specific source. To evaluate the total impact of radionuclides
released at each stage of the nuclear fuel cycle, the results
are evaluated in terms of collective effective dose per unit
electrical energy generated, expressed as man Sv/(GW a).
Only exposures to members of the public are considered inthis section. Occupational exposures associated with nuclear
power production are addressed in section III of this annex,“Occupational radiation exposure”.
(a)Uranium mining and milling152. In the period 1998–2003, a total of about 35,000 t ofuranium was produced annually in 24 countries (table 14).
The major producer in this period was Canada, with about
30% of world production, followed by Australia, with 21%
of total production. Since the beginning of the nuclear era,243

37 countries have been involved in uranium production. The
cumulative production up to 2003 is presented in figure XV.
Canada produced about 21%, the United States 20% andGermany 12% of the total amount of uranium produced globally up to 2003, except for the amount produced in theformer Soviet Union (about 20% of total production) and the
production in China before 1990 [O16, O17, O21]. Annual
production has decreased since 1990 but since 2000 has beenquite stable (figure XVI).
153. There are a large number of mining areas being
decommissioned. The countries that have declared mining
areas decommissioned or under decommissioning throughtheir National Reports to the Joint Convention on Spent Fueland Radioactive Waste Management [I38] are Argentina[R13], Australia [C26], Bulgaria [R9], Canada [M28], theCzech Republic [C31], Denmark [N5], France [F14], Germany [F2], Slovenia [R12], Spain [S29] and the UnitedStates [U24]. Other countries with environmental liabilities
resulting from uranium mining are Brazil [F5], Estonia [R3],
Kazakhstan [K12], Romania [B18] and Ukraine [R19].
154. Milling operations involve the processing of the ore to
extract the uranium in a partially refined form, known as
yellowcake. In 2003, there were 294 uranium milling installations in operation and eight under construction worldwide;
149 installations had already been decommissioned and 231
were shut down or being decommissioned [I28].
155. Effluents and solid waste. Mining operations havebeen carried out in open pits, in underground mines and byin situ leaching. Uranium mill tailings are generated at about
one tonne per tonne of ore extracted, and they generally
retain 5–10% of the uranium and 85% of the total activity[V4]. The estimated amounts of tailings worldwide areshown in figure XVII; they total about 2.35 × 109 t. Besides
the tailings, waste rock piles may also become a source of
public exposure. For open-pit mining, the amount of debris
produced is from 3 to 30 tonnes per tonne of extracted ore.
For underground mining, about ten times less debris is produced. On the basis of information provided for 13 mining
sites in Argentina [R13], Canada [M28], Germany [F2] andSpain [S29], the amount of waste rock varies from 40 to
6,000 times the amount of tailings, with an average value ofabout 1,600 tonnes of waste rock per tonne of tailings [I38].
156. Tailings are often confined because of the associatedrisk. At some locations, exposure to radon may be of considerable concern, but it is sometimes not addressed. For example, at some tailings locations, exposure to radon may
become important where the site is subsequently used for
housing, as has happened in eastern Germany, the Czech
Republic and other eastern European countries [V4]. Problems may also arise from exposure via aquatic pathways,
since acid drainage can leach uranium from waste piles[A14, F5]. The erosion of covers, structural failure ofembankments, seepage to ground or surface water and244

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emanation of radon are some of the more important­mechanisms for release of pollutants to the environment.
157. Critical exposure pathways tend to be site-dependent.
The radionuclides of greatest concern for atmospheric pathways are 222Rn, its decay progeny, and airborne particulates
containing thorium, radium and lead. The main concern foraquatic pathways is 226Ra, although 238U, 230Th and 210Pb may
be equally important [V4]. Many abandoned sites exist, andonly a few have been remediated. Problems associated withpublic exposure resulting from past practices include radon
release, water contamination, the proximity of contamination
to human settlements, the removal of wastes for construction,
large inventories and appreciable aerial dispersion [V4].
158. Some remediated sites related to former uranium
mines have follow-up monitoring and assessment programmes on contamination in the environment. Althoughsome limited descriptions of these are available in the literature, little or no useful information exists on exposures toactual population groups, because most assessments were
performed conservatively to demonstrate compliance withregulations limiting doses to hypothetical critical groups.
159. There are few new data on releases of radionuclides
due to mining and milling operations. Previous UNSCEAR
reports have estimated the average release of radon forunderground mines as approximately 75 TBq/(GW a). There
were no estimates of releases due to open-pit operations. In
the UNSCEAR 1993 Report [U6], the average normalizedradon release from mills in Australia and Canada was estimated from the limited data available to be 3 TBq/(GW a)[U6]. These values are not expected to change with current
mining and milling practices. The long-lived precursors of
222
Rn, namely 226Ra (half-life 1,600 a) and 230Th (half-life
80,000 a) are present in mill tailings and constitute a longterm source of radon release to the atmosphere. On the basisof the UNSCEAR 2000 Report [U3], the normalized radon
releases are 3 and 1 TBq/(GW a) for operational and abandoned tailings, respectively, and these values are used here.
The in situ leach facilities have no surface tailings and little
radon emission after closure.
160. Dose estimates. The methodology used by the Committee to estimate the collective dose due to mining and milling is described in the UNSCEAR 1977 and 1982 Reports[U9, U10]. Dose estimates are based on representative
release rates from a “model” mine and mill site having thetypical features of existing sites. The results are therefore notapplicable to any particular site without due consideration ofsite-specific data, and rather are meant to reflect the overall
impact of mining and milling facilities. The collective effective dose per unit electrical energy generated is estimated tobe 0.2 man Sv/(GW a) during operation of the mine and mill,and 0.0075 man Sv/(GW a) per year of release from the pilesof residual tailings of operational mining and milling sites.
161. With the current production of about 35,000 t/a andwith the assumption that 12 countries produce more than

500 t/a, the average annual individual effective dose of
25 µSv (which assumes that the collective dose is received
by the population within 100 km of the mine and mill sites)
is still valid for the major producing countries. Considerable
deviations from the representative values of parameters
selected are possible for the more general conditions ofpresent practice. There are locations in Brazil, for example,
where acid leaching may be responsible for high concentrations in drainage waters from the mining area [A14, F5].
Also, very high population densities are reported in areas
surrounding the mills in China. In some cases, previously
abandoned tailings may not have been so carefully secured
as they might have been. Although careful management oftailings areas would be expected in the future, the extremes
in management approaches (from leaving the tailings uncovered to providing secure and covered impoundment) could
increase or decrease the estimated exposure by at least anorder of magnitude.
(b)

Uranium enrichment and fuel fabrication

162. For light-water-moderated and -cooled reactors(LWRs) and for advanced gas-cooled, graphite-moderated
reactors (AGRs), the uranium processed at the mills needs tobe enriched in the fissile isotope 235U. Enrichments of 2–5%
are required. Before enrichment, the uranium oxide (U3O8)
must be converted to uranium tetrafluoride (UF4) and then touranium hexafluoride (UF6). Enrichment is not needed for
gas-cooled, graphite-moderated reactors (GCRs) or heavywater-cooled and -moderated reactors (HWRs).
163. There were 29 uranium conversion/recovery facilities
in operation and 1 under construction in the world in 2003;
2 had already been decommissioned, and 14 had been shut
down or were being decommissioned. For uranium enrichment, there were 21 operating facilities, 2 under construction, 5 decommissioned and 7 shut down or being
decommissioned. For fuel fabrication or heavy-water production, there were 66 operating facilities, 5 under construction, 23 decommissioned and 27 shut down or being
decommissioned [I28]. Nominal capacities for uranium
enrichment, hexafluoride conversion and fuel fabrication bycountry are presented in table 15, while countries with
nuclear fuel production facilities are shown in figure XVIII[I35].
164. The releases of radioactive material from conversion,enrichment and fuel fabrication plants are generally small
and consist mainly of uranium series isotopes. For the“model” installations, the normalized collective effective
dose due to these operations was estimated to be
0.003 man Sv/(GW a). Inhalation is the most important
exposure pathway. The collective doses to local and regionalgroups resulting from liquid discharges comprise less than
10% of the total exposure. The average annual collective
dose for the period 1998–2002 is estimated to be 0.8 man Sv.
Considering that 18 countries have nuclear fuel enrichment
and/or fabrication facilities, the estimated annual individual

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

effective doses would be about 0.2 µSv for local populationgroups and about 0.1 nSv for regional groups.
(c)

Nuclear power reactors

165. Reactors used for electrical energy generation are forthe most part classified according to their coolant systems
and moderators: light-water-moderated and -cooled pressurized- or boiling-water reactors (PWRs, WWERs, andBWRs,); heavy-water-cooled and -moderated reactors(HWRs); gas-cooled, graphite-moderated reactors (GCRs);
and light-water-cooled, graphite-moderated reactors(LWGRs). These are all “thermal” reactors, in which themoderator material is used to slow down the fast fission neutrons to thermal energies. In fast-breeder reactors (FBRs),
there is no moderator, and fission is induced by fast neutrons; the coolant is a liquid metal. FBRs make only a minor
contribution to energy production. A list of reactors that
operated in the period 1998–2002 and their installed capacities is presented in table A-4, and the worldwide distributionof operational reactors for the same period is shown in figure XIX. The electrical energy generated by these various
types of reactor up to 1997 has been presented in previous
UNSCEAR reports, and values for individual reactor sites
for the period 1998–2002 are given in table A-5 [I31]. A
summary for each reactor type is presented in table 16.
166. The average energy generated by nuclear power from1998 to 2002 was 278 GW(e)/a (net gigawatts of electrical
power per year), ranging from 264 GW(e) in 1998 to
288 GW(e) in 2001. The tendency for increasing amounts ofenergy to be generated by nuclear power continues. The net
installed electrical energy capacity of nuclear power plants,
the number of operating reactors and the average net installedcapacity  per unit power reactor are still increasing worldwide (figure XX). In the period 1998–2002 covered by thisannex, there were 452 operational reactors. Of these, 23 hadstarted operating in the period, 14 were shut down and 8 had
not generated energy in the period. Between 2003 and 2005,
10 new reactors started operation and 8 were shut down. In
the same period, there were also 22 nuclear power reactors
being built in 10 countries. By 2007, the number being built
increased to 30 reactors in 13 countries [I31]. The time trend
for total energy generated by reactor type is shown infigure XXI.
167. PWRs contribute the largest fraction of the total
nuclear energy generated worldwide, about 67% for theperiod 1998–2002, followed by BWRs, with a contributionof about 24%. The contributions of other reactor type are:about 5% for HWRs, 3% for LWGRs and 2% for GCRs.
FBRs contribute very little, only about 0.1% of the totalenergy generated. The average contribution for each reactor
type can be seen in figure XXII for the period 1998–2002
covered by this annex and for the period 1970–1997. The
current smaller contribution from GCRs reflects the interruption in nuclear power production (later resumed) by somereactors in the United Kingdom.

245168. The Committee derived average releases of radionuclides from reactors on the basis of reported data; these averages have been used to estimate the resulting exposures for areference reactor. The geographical location of the reactor,
the release points, the distribution of the population, foodproduction and consumption habits, and the environmental
pathways of radionuclides are factors that influence the calculated dose. The same release of activity and radionuclide
composition from different reactors can give rise to different
radiation doses to the public. Thus the calculated exposures
for a reference reactor provide only a generalized measure ofreactor operating experience but nevertheless serve as standardized measures for analysing longer-term trends from thepractice.
169. Effluents. Information on effluents released fromoperating nuclear power plants have been provided by United
Nations Member States for the UNSCEAR Global Survey
on Public Radiation Exposures, and by the International
Atomic Energy Agency (from its DIRATA database [I30]).
Data have been published by the European Commission[E15, V2] and the United States [N16, N17, N18, N19]. For
the Republic of Korea, data were obtained from the national
report to the Joint Convention on the Safety of Spent FuelManagement and on the Safety of Radioactive Waste Management [I38, R11]. Most of the available data are related to
PWRs (including WWERs), BWRs and HWRs, with only
very limited information for AGRs GCRs, LWGRs and
FBRs. The radioactive material released in airborne and liquid effluents from reactors during routine operation for theperiod 1998–2002 are reported in tables A-6 to A-12. For
airborne effluents, the releases of noble gases (table A-6),
tritium (table A-7), 131I (table A-8), 14C (table A-9) and particulates (table A-10) are given. For liquid effluents, thereleases of tritium are given in table A-11 and of other­radionuclides in table A-12.
170. The normalized releases have traditionally been compiled separately for each reactor type. This is justified
because of the different composition of the releases, mainly
for noble gases, and different “dose factors” are required toestimate the doses for different reactor types. With relatively
complete data, little extrapolation is needed for estimating
the collective doses resulting from the total releases, and thenormalized values are retained by reactor type mainly forconvenience. The results are presented in table 17. These
values are intended only for use in estimating the contribution of operating nuclear power plants to the overall public
exposure and should not be used for comparison between
different reactor types for other purposes. The choice of a
specific type of reactor for generating purposes must take
into account several aspects, such as the safety of the reactors; the impact of the complete fuel cycle, including waste
generation; and the industrial infrastructure available in eachcountry—factors that are not covered in this annex. In addition, effluent releases are dependent on the reactors’ age (theperformance of older reactors is usually different from thatof more modern reactors) and also on improvements in waste
management systems. Also, reactors that have had long

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shutdown periods for maintenance operations may present
higher than usual values because, while effluent discharges
may be enhanced, the power generated is zero. The information used in this annex includes the total energy generated
and the total effluents released in each year, and these figures
do not explicitly take account of the difference due to maintenance periods. Only those reactors that have not generated
energy during a whole year have been excluded from thisanalysis. On the whole, the values for the average release perunit energy generated are consistent with results from previous UNSCEAR reports. In general, normalized releases are
decreasing with time.
171. The largest contributions to the activity of the effluents
released are associated with tritium and noble gases. Fromthe information available, the release of noble gases per unit
energy generated is higher for LWGRs than for other reactor
types. The amount of tritium release in both atmospheric andliquid releases is higher for heavy-water reactors. Normalized values for noble gases released from nuclear power
plants over different time periods are shown in figure XXIII.
Except for LWGRs, all other reactor types show a decrease in
the noble gas activity released per unit energy generated; thismay reflect improvements in waste management procedures
and in the design characteristics of modern reactors.
172. Local and regional dose estimates. The concentrationsin the environment of released radionuclides are generally
too low to be measurable except close to the nuclear facility
and then only for a limited number of radionuclides. Therefore dose estimates are based on effluent data. Environmental
transfer and dosimetric models were reviewed in annex A,“Dose assessment methodologies”, of the UNSCEAR 2000
Report [U3]. Again, because of the variability in annual
releases, normalized releases, in TBq/(GW a), are averaged
over a five-year period to estimate collective doses. The doseconversion factors used in estimating doses were the same as
those used in the UNSCEAR 2000 Report [U3] and weresummarized in table 2.
173. The collective doses estimated for local and regional
population groups combined are presented in table 18. The
collective dose for 1998–2002 is lower than that for theperiod 1990–1994 given in the UNSCEAR 2000 Report[U3]. The main reasons for this are the lower values for noble
gas releases from BWRs and the absence of a contribution
from GCRs in the United Kingdom that were not in operation in the period 1998–2002. The average annual collective
dose to local and regional groups due to effluents released
from nuclear power plants in the period 1998–2002 was estimated as about 75 man Sv. (If the estimates were to be madeusing a simpler approach, i.e. considering the total effluent
releases from all reactors of a specific type divided by the
total power generated by those reactors, the averages would
be less sensitive to the performance of individual reactors
and would probably be a more representative estimate of theworldwide average dose. The results of such a calculation
would provide a value for the annual collective dose of about
42.6 man Sv.)

174. To estimate values for the per caput local and regionaldoses, it is assumed that the total collective dose relates tomodel population groups around all nuclear power plants:
the local population is assumed to lie within a 50 km radiussurrounding a nuclear power plant and its population density
is taken to be 400 km-2; the regional population is assumed tolie within a 2,000 km radius from the nuclear power plantand its population density is taken to be 20 km-2. Using themodel site described in reference [U3], with 444 operational
reactor units and an average of two reactors per site, theCommittee has estimated that the per caput effective dose
due to each site would be about 0.1 μSv annually for localgroups (50 km radius) and only a fraction of a nanosievert
for the regional groups within a 2,000 km radius surrounding
a site.
175. The annual doses estimated for critical groups used
for licensing and effluent control of nuclear power plants are
considered to apply to the area within a 3 km radius of thereactors and in most countries are constrained by an annual
dose limit in the range 200–300 µSv, but actual doses areusually much lower than this. Considering that more than
80% of the collective dose is due to airborne effluents, andtaking the difference between the values for dilution factors
for the representative source and long-term average conditions as defined in annex A of the UNSCEAR 2000 Report[U3] for the distance of 1 km for the critical group, it can beassumed that, for the period 1998–2002, the expected maximum annual effective doses to critical groups within 1 km ofreactor sites due to effluent releases from nuclear power
plant operation are of the order of 0.02 mSv.
176. Some information was also available for releases fromsome shut down reactors. These releases cannot be treated as“operational” releases, because they are not associated withthe generation of nuclear energy. The values of the total
releases from some shut down reactors are presented intable 19. A comparison of total releases from these reactors
with those from operational reactors of a similar type andpower shows that releases from shut down reactors are significantly smaller than those from the equivalent operational
reactors, although some exceptions may be found, mainly
related to old and low-powered shut down reactors.
(d)Fuel reprocessing

177. The reprocessing of spent fuel is performed to separate out and recover reusable uranium and plutonium fromwaste. Most spent fuel from reactors is retained on site ininterim storage, pending decisions on ultimate disposal or
retrievable storage. It is estimated that about one third of the
spent fuel already produced has been submitted to the repro­
cessing stage of the nuclear fuel cycle [I34]. France, Japanand the United Kingdom are the main countries operating
commercial reprocessing plants.
178. Effluents. Relatively large quantities of radioactive
material are involved at the fuel reprocessing stage, and theANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

potential for its release in waste discharges is greater than forother stages of the fuel cycle. Routine releases have been
largely in releases of liquid effluents to the sea. Operating
standards have been considerably improved at reprocessing
plants over the years, with substantial reductions in theamounts released.
179. In 2003, there were 13 fuel reprocessing plants inoperation, 3 under construction, 13 decommissioned and 18
shut down or being decommissioned [I28]. Information on
releases from some of these installations for the period
1998–2002 is presented in table A-13 for airborne effluentsand in table A-14 for liquid effluents. The origins of thesedata were countries’ responses to the UNSCEAR Global
Survey on Public Radiation Exposures, the IAEA DIRATA
database [I30] and the open literature [E15, V2]. Included
are data for the reprocessing facilities at La Hague (France),
Karlsruhe (Germany), Krasnoyarsk and Tomsk-7 (RussianFederation), Dounreay and Sellafield (United Kingdom) andTokai (Japan).
180. Collective doses from nuclear fuel reprocessing can be
estimated from the normalized releases per unit energy generated, the electrical energy equivalent of the fuel repro­cessed
and the collective dose per unit release of radionuclides [U6].
Previous UNSCEAR reports used dose factors based on theelectrical energy equivalent of the fuel repro­cessed. The same
methodology cannot be used here, because information on theamount of fuel reprocessed is not available. Doses were thus
estimated on the basis of the activity released in the effluents,using the dose conversion factors presented in table 3. The
data collected are currently not complete, and this necessarily
introduces large uncertainties into the resulting estimates.
Using only the available reported data, the average annual collective dose is estimated as 30 man Sv, with about 8 man Sv
due to airborne effluents and about 22 man Sv due to liquid
effluents, as shown in table 20. The estimate for the total collective dose since the beginning of reprocessing is
4,828 man Sv. The largest contribution to the total dose estimate is still associated with the release of 14C. The actual values for the total doses, however, are probably a little larger
than these estimates, because some data are missing thatwould be needed to estimate doses accurately.
181. The estimate of the annual collective dose is still in
the range 20–30 man Sv; if this were exposing a single local
population group (say, 3.1 × 106 persons within a 50 km
radius), the per caput effective dose would be about 10 µSv
per year of operation. The corresponding value for regionalgroups would be about 0.12 µSv/a. Considering that five
installations have contributed to this collective dose, theaverage effective doses would be of the order of 2 µSv for
local population groups and 0.024 µSv for regional groups.
(e) Globally dispersed radionuclides
182. Radionuclides that are long-lived and easily dispersedin the environment can give rise to doses to people across the247whole planet. The radionuclides of specific interest are 3H,
14C, 85Kr and 129I, with half-lives of 12.26, 5,730, 10.7 and1.6 × 107 years, respectively. The large uncertainties involvedin estimating doses over prolonged time periods are due toproblems in predicting environmental pathways, population
distributions, dietary habits, climate change, etc. The uncertainties in dose calculations increase when the integration iscarried out for very long periods of time—hundreds or thousands of years or even longer. Considering the 100-year truncated dose coefficients of 0.004 man Sv/(GW a) for 3H,
6.3 man Sv/(GW a) for 14C, 0.12 man Sv/(GW a) for 85Kr and
0.0008 man Sv/(GW a) for 129I releases [U7], and a continuing practice of about 300 GW a energy production per year,
the worldwide maximum per caput effective dose rate would
be about 0.18 µSv/a.
(f)

Solid waste disposal

183. Solid wastes arise at various stages in the nuclear fuelcycle. They include low- and intermediate-level wastes,
mainly from reactor operation; high-level waste from fuelreprocessing; and spent fuel for direct disposal. Low- andintermediate-level wastes are generally disposed of by shallow burial in trenches or concrete-lined structures, but moreadvanced disposal sites also exist. High-level waste andspent fuel are currently retained in interim storage tanks
pending the development of adequate methods for disposal
and the selection of disposal sites.
184. The radiological impact assessment of a high-level
waste repository has to rely on modelling of the long-term
behaviour of the waste and the migration of released radionuclides both near the site and at greater distances over along period of time. To carry out such performance assessments, a number of site-specific data are needed, including
those called for by waste characterization and transportmodels. Such assessments have been performed, mainly foruse in formulating design criteria for the hypotheticalrepositories.
185. Information on spent fuel has been obtained fromthe National Reports of countries that are parties to the
Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management[I38]. The direct comparison among different countries isdifficult because inventories are specified in different
ways: some declare the total mass stored, while others
declare the uranium mass or the heavy-metal (HM) mass.
Other countries report the volume, and a few present theactivity (but sometimes it is not clear if the value givenrefers to the uranium component or if it includes activity
from fission products). From the material provided by afew countries that have declared their total inventory ofspent fuel by nuclear power plant and also declare thatthey do not reprocess their spent fuel, average values forthe annual spent fuel generation per unit installed electrical capacity have been estimated and are presented intable 21.

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UNSCEAR 2008 REPORT: VOLUME I

186. Considering the number of operating years, the typeof reactor and the net electrical capacity, a total amount of
about 210,000 t of HM in spent fuel is estimated to havebeen generated worldwide up to the end of 2002 from nuclear
power plants that were operational in the period 1998–2002.
This amount includes the material that has already been
reprocessed, which amounts to about 90,000 t worldwide[I34]. Because this figure also includes material reprocessedfrom reactors already shut down, there are at least some
120,000 t of HM in spent fuel from nuclear power plants
being stored, most of it currently in temporary storage
conditions.
187. Before final disposal, all such material will have to be
manipulated and transported, which will give rise to both
occupational and public exposures. Public exposure due tothe transport of spent fuel is discussed in section II.C.2 ofthis annex. The transport of radioactive material of various
types between nuclear fuel cycle installations may cause
members of the public who happen to be near the transport
vehicles to be exposed. The transport of radioactive andnuclear material is addressed as a separate item in this annex.
For the nuclear fuel cycle, doses may be estimated using thefactor of 0.1 man Sv/(GW a), as in previous UNSCEAR
reports [U3, U9, U10].
188. The routine operation of nuclear power plants generates large amounts of long-lived and high-activity radioactive waste. Although there is information on waste inventories
for several countries, only a few of these inventories aredescribed in detail with respect to the specific origin of thewaste. Some countries report the volume after treatment andconditioning while others report the weight produced; care isneeded in interpretation. Nevertheless, on the basis of information provided by Argentina [R13], Canada [M28], Hungary [R10], the Republic of Korea [R11], Spain [S29] andSwitzerland [D6] in their National Reports to the Joint Convention on the Safety of Spent Fuel Management and on the
Safety of Radioactive Waste Management [I38], the annual
amount of radioactive waste generated by different types ofreactor per unit installed capacity was estimated, and theseresults are also presented in table 21.
189. Doses due to solid waste disposal have been estimatedon the basis of the projected eventual migration of radio­
nuclides through the burial site into groundwater. These estimates depend critically on the assumptions about thecontainment of the solid waste and the site characteristics,
and accordingly are generally highly uncertain. The approximate normalized collective effective dose due to low- andintermediate-level waste disposal is, however, relatively low,of the order of 0.5 man Sv/(GW a), and is due almost entirely
to 14C [U6, U9]. The worldwide average annual per caput
effective dose rate would be about 1 nSv per year ofpractice.
190. The decommissioning of nuclear facilities gives rise
to radioactive waste, and decommissioning experience isbeing accumulated. Worldwide a considerable number of

installations have been shut down or are being decommissioned. The 2003 list includes 231 uranium milling facilities,
14 uranium conversion/recovery plants, 7 enrichment facilities, 27 fuel fabrication/heavy-water production facilities
and 18 fuel reprocessing plants. Also, 107 commercial
nuclear power plants had been shut down or were undergoing decommissioning. There were also 21 research reactors
and several research units undergoing decommissioning.
191. According to the information provided by countries in2005 under the arrangements for the Joint Convention on the
Safety of Spent Fuel Management and on the Safety ofRadio­active Waste Management, the largest decommissioning programme in 2005 was that of the United States, which
was decommissioning 16 nuclear power reactors, 20 research
reactors, 66 radioactive installations and about 1,186 sites
formerly used for activities related to defence [U24]. France
was decommissioning 9 nuclear power reactors, 15 research
reactors, 3 small reactors used for defence activities and 16
other installations [F14]. Germany had 17 nuclear powerreactors and 14 research reactors undergoing decommissioning [F2]. All these processes will generate large amounts ofradioactive waste, including spent fuel from both researchand power reactors, which will have to be handled, transported and disposed of. With the information available it is
not possible to estimate the total amount and activity of thewaste to be disposed of. The amount of waste will be highly
dependent on the facility type, size and operational history.
As an example, the Republic of Korea has estimated thatsome 620 m3 of waste with an activity of about 1.24 × 1012 Bq
will be generated by the decommissioning of two research
reactors, and some 380 m3 of waste with an activity of about
6.5 × 105 Bq by the decommissioning of one conversion
facility. The decommissioning of two nuclear power reactors
in Canada, at Douglas Point and at Gentilly-1, has left a totalof 300 and 80 t, respectively, of uranium in spent fuel. The
expected total volumes of conditioned waste from existing
United Kingdom facilities to the end of their lives are
1.5 × 106 m3 of low-level waste, 2.4 × 105 m3 of intermediatelevel waste and 1.5 × 103 m3 of high-level waste [I38].
192. There are also currently five reprocessing facilities
with a capacity of greater than 1 t/a undergoing decommissioning and two more for which decommissioning isplanned [I34]. Information related to exposures from thedecommissioning of such facilities is scarce and relates
more to the level of compliance with regulatory constraints
than to actual public exposure. While worker exposure mayarise from dismantling, demolition and waste management
operations, public exposure will depend on the criteria
adopted for residual radioactivity at the site and on thetransport of waste to disposal sites. The information available indicates that the exposure of the public due to decommissioning will be very low and will be constrained in thelong term by national regulations regarding acceptable­levels for residual radioactivity in recycled materials and in
the environment. Estimates of doses due to waste rock andtailings were, however, included in the doses estimated for
uranium mining and milling.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

249

193. During the decommissioning of facilities, many materials may be recycled. Different criteria are being applied bydifferent countries, but the information currently available isnot sufficient to estimate the contribution from recycledmaterials to public exposure [B13, I19, I24, I40, L13].

often considered as a subset of the “public” who are generally exposed to the highest dose rates [V23]). However, in
this annex, exposure of this group of workers will, wheneverpossible, be considered separately from the exposure ofmembers of the general public, such as pedestrians, passengers and bystanders.

(g) S ummary of estimates of doses due to nuclear power
production

197. The data on numbers of packages are generally wellknown for nuclear fuel cycle operations, but for other transport operations the number can in most cases only be estimated. There is also a large variation between countries in
the number of packages, because some countries havenuclear fuel cycle operations and some have major suppliersof radionuclides [H29]. The IAEA estimates that 10 million
shipments of radioactive material are transported annually.
Each shipment is made up of either a single package or anumber of packages [I5]. The vast majority, some 95%, of
these shipments are unrelated to the nuclear fuel cycle, only
5% being related to fuel cycle transport [W16].

194. The estimated doses to members of the public due tothe generation of electrical energy by nuclear power are summarized in table 22. For local and regional population groups,
a normalized factor of 0.27 man Sv/(GW a) has been determined. This is slightly lower than the value derived in the
UNSCEAR 2000 Report [U3], 0.44 man Sv/(GW a). For all
activities related to the production of energy, a normalized
collective effective dose of 0.72 man Sv/(GW a) was determined. Using this coefficient, with an average of 278 GW(e)/aper year produced in the 1998–2002 period, an annual collective dose of about 200 man Sv is estimated for all operations
related to energy production. The annual per caput doses torepresentative local and regional populations surrounding
nuclear power installations are less than 10 μSv. The collective doses from globally dispersed radionuclides are delivered over very long periods and to an assumed “maximum”
population of the world. If the practice of nuclear power production were limited to the next 100 years at the present
capacity, the maximum annual per caput effective dose to theglobal population would be less than 0.2 μSv.
2.

Transport of nuclear and radioactive material

195. This section describes exposures related to the “normal transport” of radioactive material. Normal transportrefers to operations that occur without loss or damage to the
package and without an accident involving conveyance.
Events in which the shipment is not timely, the package orconveyance is damaged or the contents are lost or destroyed,
are considered to be transport accidents or incidents [I20].
Accidents and incidents do occur during transport, but theirconsequences are normally limited by built-in safety features of the package together with the controls required fortransport, including emergency response procedures [H28].
The consideration of accidents is outside the scope of thisannex and is discussed in annex C, “Radiation exposures in­accidents”, of the UNSCEAR 2008 Report.
196. Radioactive materials of natural or artificial origin areused widely around the world and are transported within andbetween countries. A wide range of different materials aretransported, from small quantities of radiopharmaceuticals
for medical purposes to highly radioactive spent nuclear fueland vitrified waste arising from the nuclear fuel cycle. The
handling and transport of these radioactive materials cangive rise to the radiation exposure of workers and of members of the public. In a lay sense, the term “public” is often
taken to include transport workers (i.e. transport workers are

198. Road, rail, air and sea transport are all commonly used
for the transport of nuclear fuel cycle material, of radioactive
material to be used in medicine, industry and research, andof waste. Air transport of nuclear fuel cycle material is carried out only to a limited extent [W14]. The available data
indicate that exposures under normal conditions of transport
are low. At least for the United Kingdom and the UnitedStates, the transport of fuel cycle material contributes significantly less to the exposure of transport workers than does
the transport of non-fuel-cycle material [I5].
199. Mobile radiography sources are relatively numerous
throughout the world. For example, there are about 850 inFrance, and about one half of these are transported daily byusers from storage to their place of use. The transport dose
for radiography operators due to these sources has not been
included in this annex because of the difficulty in distinguishing between doses arising from the transport of radioactive material and doses resulting from the radiography
operations themselves [C3, H4].
200. The number of fuel transports in Germany is shownin figure XXIV for the period 1994–2002. The number
includes transports of irradiated and non-irradiated fueland of waste by rail, road, sea and air. The overall numberof transports decreased from 1994, reaching a minimum inabout 1999, and started a relatively slow growth up to
2002. In 2002, the major contributor to the number oftransports was the international transport of non-irradiatedfuel by road, which accounted for nearly 50% of all nuclear
fuel transports, with the international sea transport of­non-irradiated fuel ­accounting for about 20% [B48].
(a)Transport by land routes

201. Most transport operations include the initial roadtransport from the production site to the railway station, airport, harbour or collection centre. During this part of the250

UNSCEAR 2008 REPORT: VOLUME I

journey, vehicles may pass through residential or crowded
areas and along busy roads and highways. In certain places
the pedestrian density may be unusually high, while in ­others
there may be only people waiting at bus stops to represent
the potentially exposed population. Passengers in vehicles
near the delivery vehicle will also be exposed. Overall, theexposure of such members of the public is expected to bemuch lower than of the exposure of workers involved in
actual transport and cargo operations. Collective doses will
depend mainly on the population density along the transportroute [V23]. Although it would appear that both collective
and individual doses to the public are low [I5], data arescarce.
202. A survey was performed in Mumbai, India, regarding
the carriage of radioactive material in “Type A” packages[I33] for use in medicine, industry and research; such material is supplied by the Bhabha Atomic Research Centre,
Mumbai, to a large number of users all over India. Packages
delivered to the air cargo terminal are ultimately sent to various parts of the country. Radiation doses to the urban public
due to shipments of radioactive material are likely to bemuch greater in Mumbai than in any other city in India.
Essentially it is the pedestrians and the passengers travellingin nearby vehicles who are exposed to radiation during thetransport of radioactive material. The estimated radiation
exposures were found to be low, but collective doses could
theoretically be large because of the relatively high pedestrian and passenger density in Mumbai. Measured dose rates,
used to estimate public exposure, ranged from 1 to 55 µGy/h
for passengers in vehicles beside the delivery van and forpedestrians on sidewalks. The annual collective dose resulting from the transport of radioactive consignments in­Mumbai was estimated to be about 0.1 man Sv [V23].
203. The collective doses accruing to the general public
due to incident-free road transport of sealed spent sources
generated by nuclear application institutes in India were estimated for the year 2001 [U43]. The main contribution to collective dose was due to the transport of decayed 192Ir sourcesthat had been used for industrial radiography, brachytherapy
and nucleonic gauges, with a collective dose to both ­transport
workers and to the public of about 46 man Sv.
204. Analysis of the shipments by road for the period
1987–2000 by authorized carriers in Italy concluded that
doses mainly arise from transport operations associated withradioisotope supply and distribution, and with the transport
of non-nuclear radioactive waste. Negligible doses arise
from transport operations associated with the nuclear fuelcycle, because of the very small number of shipments ofnuclear material. The greater part of the exposures to people
due to the shipment of radioactive material for industrial andmedical uses arises from transport for medical purposes,with an estimated maximum annual individual effective doseof 0.0012 mSv [C3].
205. In France, spent fuel is carried mainly by rail. Roads
are used only between certain power plants and the nearest

railway station, and then between the rail terminal and the
La Hague reprocessing facility. Also, transport of waste
between the various producers and the storage centre at
La Hague is accomplished principally by rail and only partly
by road. The contribution of waste transport to the irradiation of workers and the public remains very low. Membersof the public residing along the (road and rail) transportroutes or near the sites of storage in transit may receive
doses due to the transport of radioactive material. The
annual collective dose received by the public was estimatedto be at most about 0.10–0.15 man Sv, about half of the dose
received by ­workers [H4].
206. Collective doses were assessed for population groups
in the vicinity of the transport routes within Germany andwere related to the incident-free transport of radioactive
material and spent fuel. The population groups outside
nuclear facilities considered in this assessment included
railway personnel in shunting yards, populations in thevicinity of shunting yards and rail routes, and railway passengers. Collective dose estimates for three types of spent
fuel management are presented in table 23. Comparison
with estimates for low-level waste showed that the main
contributor to the collective dose for all population groups
would be the transport of low-level waste from PWRs to thedepository at Gorleben, which exceeded the exposure due tothe transport of spent fuel from PWRs by two orders ofmagnitude. The transport of spent fuel accounts for lessthan 1% of the total dose. The annual dose to a hypothetical“critical group” passenger who passes every shipment ofradioactive material from the reprocessing facility to thedepository was estimated to be 0.08 mSv. Inhabitants who
spend the entire year within a distance of 1 km from therailway track would receive a conservatively estimatedannual dose of 0.03 mSv [B7].
207. A study was performed in the former German Democratic Republic related to an impact assessment of the transport of waste to the Endlager für radioaktive Abfälle
Morsleben (ERAM), a former salt mine located in SaxonyAnhalt. After a temporary shutdown, ERAM restarted disposal operations in 1994. From then to the end of 1996, some
11,000 m3 of waste, primarily low-level solid waste fromoperating nuclear power plants and from decommissioning,were delivered and placed in deep geological formations.
The preferred mode of transport for waste shipments wasrail, except for a small fraction of the journey within 40 kmof the repository site. Estimates of annual doses to membersof the public were generally less than 0.1 mSv [S12].
208. Some 500,000 packages of radioactive material areshipped annually within the United Kingdom by road andaround 4,000 movements annually by rail. About 52,000 ofthese packages are shipped to and from hospitals; about 15%of these contain technetium generators. Doses to members of
the public due to the transport of radioactive material tend tobe very low. The estimated maximum individual doses wereless than 20 µSv/a. The estimated collective dose to the public due to the movement of radioisotopes to and from

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

hospitals was 0.013 man Sv, with a further 0.005 man Sv due
to exports. The estimated collective dose to the public due tothe movement of spent nuclear fuel flasks within the United
Kingdom was no more than about 0.001 man Sv [W7]. A
study on the transport of NORM in the United Kingdom
found that the annual dose to any member of the public fromthe shipment of any type of NORM would be much less thana microsievert [H30].
(b)

Transport by sea

209. Spent fuel from Japan is transported by sea in dedicated vessels for reprocessing in Europe, arriving at sea terminals close to the reprocessing plants and then undergoingshort road/rail journeys. Spent fuel flasks are handled by
cranes at the sea terminals, with limited access by workers.
Some spent fuel is likewise transported by sea from continental Europe to the United Kingdom. The limited transportof high-level waste, for example from La Hague in France tostorage facilities elsewhere in Europe and in Japan, follows
procedures similar to those for spent fuel transport [W14].
210. Non-irradiated nuclear fuel material (such as ores,
concentrates and chemical derivatives) are shipped aroundthe world in various types of container, while irradiatedmaterial is generally transported by special ships dedicatedfor this purpose. Packages containing radioactive material
are usually carried in containers on board ships or in vansand lorries. The containers are usually loaded with material
for the nuclear fuel cycle, while the vehicles usually carrypackages of radionuclides for medical or general industrialuse [B11].
211. In 1994 radioactive material was carried on some
1,100 voyages to, from or in transit through the United Kingdom. Of these voyages, about 55% were of nuclear fuel cycle
material (50% non-irradiated and 5% irradiated), and theremaining 45% involved radionuclide consignments for
medical and general industrial use. The carriage of nuclear
fuel cycle material on freight vessels and dedicated ships
leads to the possible exposure of the crew, a small number ofpassengers and dockworkers. The transport of packages containing radionuclides in vehicles on ferries may result in the
exposure of both crew and passengers. Exposures to passengers are low, with annual individual doses unlikely to exceed
0.032 mSv [B11].

Harbour workers, with an annual collective dose of about
3 × 10‑4 man Sv, were the population group that received thelargest individual doses [S1].
(c)Transport by air

213. A major producer of radionuclides for worldwide
medical use is located in the United Kingdom. The radionuclides are packaged and then sent by road either for domestic
delivery or for export via a number of airports. Packages
containing radioactive material arrive at the airport in lighttrucks and are then unloaded and checked in the carrier’s
warehouses. The packages are sorted and grouped according
to destination. The majority of packages transported by air
are either excepted or Type A. Excepted packages have surface dose rates of less than 5 µSv/h. However, some Type A
packages containing technetium generators have surface
dose rates approaching 1 mSv/h. Measurements have indicated typical dose rates close to packages containing technetium generators of around 40 µSv/h, with surface dose rates
of up to 800 µSv/h [W3].
214. An extensive survey on the routes, kinds of airplane,cargo operations, crew flight schedules and numbers ofpassengers was carried out in the United Kingdom in 2001.
Aircrew and passengers may be exposed to packages stored
in holds during flight. The number of packages transported
by air in the United Kingdom in 2001 is shown in table 25.
A number of the carriers stated that they did not consider“excepted” packages to be “radioactive material” and therefore exclude this category from package totals. For passengers, measured dose rates ranged from 0.5 to 9 µSv/h in themain cabin area, and from 4 to 15 µSv/h in the front seats.
Half the passengers were exposed to dose rates of less than
1 µSv/h, and the average passenger dose rate was 3 µSv/h.
No information was provided regarding frequent flyers andcouriers, but it was considered that frequent flyers using
short-haul flights (which have a Radioactive Traffic Factorof 1 in 475) are unlikely to receive a significant dose due toradioactive cargo [W3]. (The Radioactive Traffic Factor(RTF) is the ratio of flights carrying radioactive cargo to
the total number of flights.) Estimates of collective doses
due to air transport in the United Kingdom are presented intable 26.
(d)

212. A study performed in Egypt assessed the exposure of
populations living alongside the Suez Canal due to the intensive transport of radioactive material by ships passing throughthe canal. The quantities of radioactive material that potentially exposed coastal populations are presented in table 24.
The estimated average annual collective doses to the publicin the period 1986–1992 in the towns of Port Said, Ismailia
and Suez are 4.11 × 10‑8, 3.01 × 10‑8 and 5.04 × 10‑8 man Sv,
respectively. The transport of low-­activity material, such asuranium (as U3O8), represented the largest contribution to thecollective dose to the public within the Suez Canal area.

251Summary on the exposure to radioactive material
during transport

215. In general, doses to members of the public due to thenormal transport of radioactive material are verifiably very
low. Some results of initial surveys on this topic are presented in table 27. More recent surveys produced similar
results (table 28). In Germany, the highest conservatively
estimated annual dose to members of the public due tonuclear fuel shipments was typically less than 0.1 mSv. In
France, these shipments are estimated to give rise to a maximum annual dose of 0.2 mSv, while shipments of waste at a

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storage facility are estimated to give rise to a maximum
annual dose of 0.12 mSv, and shipments by road could leadto an annual dose of up to 0.07 mSv owing to the vehicles
waiting at traffic lights. In the Netherlands, the estimatedmaximum annual dose due to both nuclear and non-nuclear
shipments was 0.02 mSv [I5]. More recent estimates predicted annual doses of less than 0.002 mSv for critical groups[E6]. In the United Kingdom, 0.02 mSv was the maximum
annual dose estimated for sea and air passengers, whileannual exposures due to road and rail transport were lessthan 0.01 mSv [I5].
3.

Applications other than nuclear power(a) Production of radioisotopes
216. Radioisotopes are widely used in industry, medicine
and research. Radiation exposures may occur owing to trace
amounts being released in production or at subsequent stagesof the use or disposal of the radionuclide-containing products. For very-long-lived radionuclides, such as 14C, all of theamount utilized may ultimately reach the environment. For
short-lived radionuclides, such as most radiopharmaceuticals,
radioactive decay prior to release is an essential consideration. The isotopes used most widely in medical examinations
and nuclear medicine procedures are 131I and 99mTc.
217. Estimates of doses resulting from radioisotope production and use are uncertain, owing to the limited availability ofdata on the commercial production of the radioisotopes andon the release fractions during production and use. The mainradionuclides of interest are 3H, 14C, 125I, 131I and 133Xe. The
estimated annual collective effective dose due to radioisotope
production and use is of the order of 100 man Sv [U6].
218. An important use of radionuclides is in medical diagnostic examinations and therapeutic treatments. Medical
radioisotopes or their parent radionuclides can be produced
in a reactor (by fission of uranium, e.g. 99Mo, 131I; or by activation, e.g. 59Fe) or in a cyclotron (by nuclear reactions, e.g.
123 201
I, Tl). The most important radioisotope, used in 80% of
all diagnostic examinations, is 99mTc (from 99Mo). In many
countries the production, isolation and incorporation of theradioisotopes into generators, diagnostic kits or pharmaceuticals are often carried out in different facilities, which hampers quantification of the releases resulting from the overall
production.
219. Limited data on 131I releases from hospitals were citedin the UNSCEAR 1993 Report [U6]. There is high excretionof 131I from patients following oral administration, but waste
treatment systems with hold-up tanks are effective in reducing the amounts in liquid effluents to a small fraction (e.g.
5 × 10‑4) of the amounts administered to patients. This seems
to be confirmed by the very low concentrations of 131I measured in the surface waters and sewage systems of several
countries [U6], although such information seems not to besystematically collected or reported.

220. With the global annual usage of 131I in therapeutic
treatments estimated at 600 TBq, a release fraction of
5 × 10‑4 and a dose coefficient of 0.03 man Sv/TBq for 131I
released in liquid effluents (taken from annex A, “Doseassessment methodologies”, of the UNSCEAR 2000 Report[U3]), the annual collective dose is estimated to be only
0.009 man Sv. The use of hold-up tanks should reduce therelease of 99mTc, the other major radionuclide, to negligible
levels as well.
221. In the United Kingdom, radioactive material, including radiolabelled materials for use in medicine, researchand industry, is manufactured at two sites: Amersham andCardiff. At Amersham, the total annual dose to criticalgroups in 2003 due to liquid discharges was assessed to beless than 5 µSv. Summing freshwater fish consumption andexternal exposure, doses to critical groups were estimated to
be of the order of 5 µSv in 2003. The doses estimated for thecritical group for terrestrial food were also less than 5 µSv
in 2003. At Cardiff, the laboratory produces radiolabelled
products containing 3H and 14C to be used in research andmedical diagnostic kits. The dose to the most exposed groupof seafood consumers was 24 µSv in 2003, including a contribution from external exposure. The hypothetical criticalgroup for terrestrial foodstuffs comprised infants who
ingested food produced on land conditioned by pelleted
sludge from the wastewater treatment works. It was assessedthat in 2003 the highest dose would have been less than
16 µSv, with doses from non-foodstuff pathways being lessthan 1 µSv [W6].
222. According to the results of a 2006 survey conductedby the IAEA, there were 246 cyclotrons operating in
39 IAEA Member States. The IAEA has estimated thatworldwide there are about 300 cyclotrons currently operating that are involved in some aspect of radionuclide production. The number of cyclotron institutions that distributeradiopharmaceuticals, and in particular 18F-labelled fluorodeoxyglucose (18F FDG), is significant and growing [I37].
No information on public exposure due to the operation of
cyclotrons has been found.
(b)Research reactors

223. Research reactors, given their wide variety of designs
and modes of operation, as well as their wide range of uses,
differ from reactors producing electrical energy. Research
reactors are used for testing nuclear fuels and various materials, for investigations in nuclear and neutron physics, bio­
logy and medicine, and for the production of radioisotopes.
The use of research reactors is globally much more widespread than the use of reactors for energy production. In
2003 there were 70 countries listed as having operated
research reactors; among the 57 countries that still operate
research reactors, there were 274 in operation; and 8 countries had a total of 10 research reactors under construction.
The number of reactors is presented in figure XXV ­according
to operational status and nominal power [I29].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

224. Three sites in the United Kingdom—Dounreay,­Harwell and Winfrith—house research reactors that havebeen or are in the process of being decommissioned. At
Dounreay, the critical group of people who consumed foodfrom the terrestrial environment was estimated to havereceived 6 µSv in 2003, which also includes a contribution
from weapons test fallout. At Harwell, although there was noevidence that fish from the river were consumed, an assumed
annual consumption rate of 1 kg was used in the dose assessment, leading to a dose estimate of 11 µSv for 2003. The
dose to the critical group of local consumers from gaseousdischarges was estimated to be less than 5 µSv. Doses­estimated for Winfrith are of similar magnitude [W6].
(c)Consumer products

225. A number of products bought for everyday use contain low levels of radionuclides. Some of these items contain
low levels of NORM, but the majority of consumer products
containing radioactive substances have had the radioactive
material deliberately added in order to make use of its chemical and radioactive properties. Historically the most significant radionuclide for use in radioluminous consumer
products was 226Ra. However, production of items luminizedwith radium ceased a few decades ago, with radium being
replaced by 147Pm and 3H because these radionuclides areless radiotoxic. For timepieces containing tritium compounds, some leakage of the radioactive source may occur,
because tritium is very mobile. Tritium emits only very weakbeta radiation that cannot penetrate the skin, so that it contributes to the effective dose only when the tritium hasentered the body [W6].
226. Ionization chamber smoke detectors are used to givean early warning of fire. Modern smoke detectors contain asmall foil of 241Am with an activity of not greater than
40,000 Bq. The dose rate at a distance of 2 m from a detector
is about 2.4 × 10‑5 µSv/h, assuming that the detector contains
the maximum amount of activity. In the United Kingdom,
about 80% of homes have a smoke detector fitted. Assuming
an exposure of 8 h/d at a distance of 2 m from the detector
results in an estimated annual dose of 0.07 µSv [W6].
227. Glass to which uranium is added to produce a yellow
or green colour is called Vaseline glass. It was very popularin the 1800s and is still produced in the United States and theCzech Republic. The gamma dose rate close to the surface ofthe glass item is very low and was measured as less than
0.1 µSv/h. A typical surface dose rate due to beta radiation
was 15 µSv/h, while the beta doses measured a few centimetres from the surface were negligible. Individual doses forsome collections of uranium glass could be up to 0.5 mSv
annually. However, for a large collection with a range ofitems, a typical maximum dose would be an order of magnitude lower. Uranium salts have also been used in the glaze on
ceramic products such as tableware and tiles. They were also
used as a colourant in ceramic tableware produced in the
1930s and 1940s in the United States. These items may now

253

be found on collectors’ markets. It was found that handlingsuch items may give rise to very low levels of contaminationon the skin, and the use of this tableware for eating could
lead to very low ingestion doses [W6].
228. Some members of the public have collections of fossils,
rocks or minerals. In some parts of the United Kingdom thenative rocks contain significant concentrations of uranium andits decay products. The overall dose from such specimens
under normal conditions of handling and display are only asmall fraction of the overall dose from natural radiation. Photo­graphic lenses used to have 232Th added to them in order toincrease the refractive index. Photographers carrying a camera
with such lenses around the neck for several hours a day on
many days of the year could receive an annual effective dose ofa few hundred microsieverts. Currently these lenses are out ofuse in United Kingdom. A summary of doses associated with
exposure to consumer products is presented in table 29 [W6].
229. The United States Nuclear Regulatory Commission(NRC) has assessed the potential individual and collective(population) radiation doses associated with selected products containing “by-product” material2 [U35]. The dose
assessments were in general based on reasonable assumptions, although in some cases the NRC noted that there wasan absence of reliable data on the actual use of the products
by individuals either in the workplace or elsewhere. The estimates reported are for effective dose equivalent to the average member of the critical group. The individual andcollective dose estimates discussed here are restricted todoses estimated for the normal life cycle of a particular product or material, covering distribution and transport, intended
or expected routine use, and disposal occurring over a 1 year
time period. Actual or expected quantities of radioactive
material in products and materials, when known, were used
for estimating doses; otherwise, a value was used equal tothe maximum allowed under the United States legislation on
exempted quantities.
230. The estimates of individual doses incurred annuallyduring the normal life cycle of a product or material associated with the current exemptions for by-product material in
the United States ranged from less than 1 × 10‑5 mSv to
0.2 mSv. A summary of individual effective doses from byproducts in the United States is presented in table 30. The
estimated individual doses were equal to or greater than
0.1 mSv annually for two products: (a) instruments used formeasuring ionizing radiation that contain by-product material, with an estimated annual dose of 0.2 mSv received by alaboratory technician working with a bench-top instrument;and (b) spark gap irradiators containing 60Co, with an estimated annual dose of 0.1 mSv received by a maintenanceworker installing and maintaining spark gap irradiators.
By-product material here includes any radioactive material associated withthe operation of nuclear reactors, except for the source material for nuclear
fuel and the special nuclear material which constitutes the fuel in a reactor.
Source material is the raw material from which nuclear fuel is made; it
includes uranium or thorium in their natural isotopic abundances.

2

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231. The estimates of collective dose incurred during thenormal life cycle of these products ranged from 0.1 to40 man Sv for 1 year’s distribution of products. For two cate­gories of products, the estimated collective doses were equal
to or greater than 10 man Sv: (a) the collective dose arisingfrom the use of timepieces with hands or dials containing 3Hor 147Pm was estimated to be 40 man Sv, which was incurred
mainly because of the large number of individuals who wearsuch timepieces (wristwatches); (b) the collective dose arising from the use of electron tubes containing by-productmaterial was estimated to be 10 man Sv over the tubes’ useful lifetime of 10 years. In this case, most of the collective
dose would result because of the large number of people
exposed to radiation from electron tubes in the home andworkplace. However, individual doses are ­normally very
low, usually less than 0.1 mSv/a.
232. The estimates of individual doses incurred annuallyduring the normal life cycle of a product or material associated with the current use for source material ranged fromless than 1 × 10‑5 mSv to 40 mSv. The estimated annual individual doses exceed 10 mSv for the following two cases(table 30): (a) chemical mixtures, compounds, solutions or
alloys containing less than 0.05% by weight source material;and (b) rare earth metals and compounds, mixtures and products containing not more than 0.25% by weight source material. The high estimates in these cases result from the large
volumes of exempted material present in workplaces and thehigh concentrations of uranium and thorium in this material.
These estimated doses would be reduced substantially forthe case of the workers using respiratory protection.
233. The estimated annual individual doses were equal toor greater than 1 mSv but less than 10 mSv for three materials: (a) for unrefined and unprocessed ore containing source
material, the estimated dose of 3 mSv/a to a truck driver
results from the large volume of exempted material that ishandled and the relatively high concentration of uranium inthe material; (b) for incandescent gas mantles, the estimatedannual dose to a person using only gas lanterns for lightwould be 2 mSv and that to an individual who uses portablecamping lanterns would be 0.1 mSv; (c) for welding rods
containing thorium, the estimated annual dose of 8 mSv to a
dedicated grinder of welding rods probably represents anunusual situation that would occur only at construction sites
where many welders are employed.
234. The estimates of collective dose incurred during thenormal life cycle of a product or material associated with thecurrent exemptions in the United States for source material
ranged from 0.001 man Sv to 700 man Sv for 1 year’s distribution. There are five situations for which collective dose
estimates are equal to or greater than 100 man Sv: (a) for
chemical mixtures, compounds, solutions or alloys containing less than 0.05% by weight source material, the collective
dose is a combination of estimated doses due to the use ofophthalmic glass, doses due the use of phosphate slag forbuilding construction, and doses to future on-site residents
from the disposal of coal ash, phosphate slag and water

treatment sludge; (b) for incandescent gas mantles, the usersof portable camping lanterns contribute most to the collective dose. The current trend towards the use of gas mantles
not containing thorium and the use of other lighting devices
should significantly reduce this collective dose estimate;(c) for welding rods containing thorium, the collective dose
estimate is 300 man Sv, although this is predominantly
received by professional welders over a 1 year time period,and only a fraction of it can be related to public exposure;(d) for glassware, the dose due to the display of large numbers of items (in homes and museums) contributes to the collective dose; (e) for thorium in finished optical lenses, theestimated doses to users of 35 mm photographic cameras
contribute most of the collective dose.
235. There are also two situations where the collective
doses were equal to or greater than 10 man Sv but less than
100 man Sv: (a) for rare earth metals and compounds, mixtures and products, the contributors to collective dose arebastnaesite and cerium concentrates (industrial workers),
television faceplates and waste disposal (future on-site residents at landfills); (b) for glazed ceramic tableware, the estimated doses are due to the display of large numbers of items(in homes and museums).
(d)Other sources of public exposure236. Estimated potential annual doses from exposures at
hospitals, institutions of higher education and other researchlaboratories where radioactive material is used in the United
Kingdom ranged from 0.02 to 13 µSv. The highest annual
dose estimated for an industrial site was 170 µSv, but thecalculation assumed authorized discharge levels as opposedto actual discharge levels, which are generally much lower.
Landfill sites may also give rise to exposure of members of
the public. Doses in 2003 to the critical group of people
who live close to the Drigg disposal facility in the United
Kingdom were 46 µSv (including components due to deposits from the Chernobyl accident and to weapons tests fallout). Low levels of radioactive material may be disposed ofat some landfill sites. It is estimated that the annual dose
incurred by ingesting water containing a leachate arisingfrom a landfill that accepts 125I in waste would be 5 µSv.
Tritium has also been detected near some landfill sites. A
person drinking water from a nearby borehole with about
1,000 Bq/L would receive an annual dose of less than
12 µSv [W6].
237. The use of radioactive substances in an unsealed form
is widespread in medicine. These substances are employedin nuclear medicine and radiotherapy departments for medical diagnosis and for treating cancers and other diseases withinternal irradiation, and also in clinical biology and medical
research laboratories. These uses result in significant volumes of radioactive waste, only a small part of which istransferred to specialist radioactive waste processing centres, while the major part is stored on the site until the activity has decreased to a level allowing the waste to be treatedANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

as normal hospital waste. In view of the large number of
establishments and departments involved and the multiple
ways of managing the waste, regulatory systems have been
put in place. However, chance incidents, such as the discovery of radium needles or radioactive waste in areas normallyaccessible to the public, or of quantities of radioactive iodine
in river waters, although without consequences for publichealth, have nevertheless alarmed the general public [E12].
238. “Orphan radioactive source” is a term utilized by
nuclear regulators to denote radioactive sources that are outside official regulatory control. Orphan sources include:
sources that were never subject to regulatory control; sourcesthat were subject to regulatory control but have since beenabandoned, lost or misplaced; and sources that were stolenor removed without proper authorization. Exactly how manyorphan sources there are in the world is not known, but the
numbers are thought to be in the thousands. The NRC reportsthat United States companies have lost track of nearly
1,500 radioactive sources within the country since 1996, andmore than half have never been recovered. A European
Union study estimated that every year up to about 70 sources
are lost from regulatory control within the Union. Although
the majority of these sources would not pose a significant
radiological risk, the risk of accidents is the major concern
arising from orphan sources. Sealed sources or their containers can be attractive to scavengers for the scrap metal trade
because they appear to be made of valuable metals and may
not display a radiation warning label. Cases where unsuspecting people or even members of the public have tamperedwith sources have led to serious injury and in some cases
death. Some of the more notable such accidents are describedin annex C of the present report.
239. Orphan sources are a widespread phenomenon in the
Newly Independent States (NIS) of the former Soviet Union.
For example, a legacy of Georgia’s sharp economic decline
after the break-up of the Soviet Union was a loss of control
over radioactive sources used in industry. The collection andsale of scrap metal from abandoned factories has provided ameans of livelihood for some persons, and some orphan
sources have been found in shipments of scrap. Not all theincidents reflect deliberate attempts to steal radioactive
sources. The great majority of the trafficking incidents
detected appear to involve opportunists or unsophisticatedcriminals motivated by the hope of profit. In some cases, thetheft of sources was incidental to the theft of vehicles. As
many as 300 radioactive sources have been recovered in
Georgia since the mid-1990s, and these sources have caused
at least one death and many injuries to the public. In 2006,
two abandoned and potentially dangerous radioactive devices
were successfully secured, one in the village of Iri, where
background radiation levels were elevated to 12 times above
normal in the village centre, and the other in the village of
Likhaura. The radioisotope in both sources was 137Cs. In
Moldova, several large devices containing about 130 TBq(3,500 Ci) of powdered 137Cs chloride used for agricultural
purposes in the former Soviet Union were found abandoned
or stored in precarious conditions [G13, I36, W8].

255

4. Summary on exposures due to ­peacefuluses of man-made sources of radiation
240. A summary of dose estimates related to public exposures due to peaceful uses of man-made sources of ionizing
radiation is presented in table 31. Currently available information does not allow estimates of global doses to be made,although individual doses are very low for sources unrelated
to nuclear power production. Although individual doses may
be up to a few millisieverts per year for specific populationgroups, in connection with some specific practices and exposure scenarios, the worldwide average annual per caput dose
is of the order of microsieverts.
D. Use of man-made sources for military purposes
1. Nuclear tests(a)Global fallout

241. Nuclear test explosions in the atmosphere were carried out at a number of sites, mostly located in the northernhemisphere, between 1945 and 1980. The periods of mostactive testing were 1952–1958 and 1961–1962. In all,502 atmospheric tests, with a total fission and fusion yield of
440 Mt, were conducted. The number and yields of worldwide atmospheric nuclear explosions as estimated by
UNSCEAR [U3] are summarized in table 32 and figure XXVI. After the Treaty Banning Nuclear Weapon Tests
in the Atmosphere, in Outer Space and Under Water was
signed in Moscow on 5 August 1963, nuclear test explosions
were mostly conducted underground [I9]. A summary of all
atmospheric and underground nuclear weapons tests bycountry is presented in table 33. Besides these, there were
39 safety tests that took place above ground, in which moreor less fully developed nuclear devices were subjected tosimulated accident conditions (i.e. the nuclear weapon cores
were destroyed by means of conventional explosives, withno or very small releases of fission energy) [I12].
242. The earlier atmospheric tests remain the principal
source of current radiation exposure worldwide due tonuclear weapons testing. Table 34 provides estimates of theactivity of radionuclides released and globally dispersed in
all atmospheric nuclear tests [U3]. Radioactive debris froman atmospheric nuclear test is partitioned between the local
ground or water surface and the tropospheric and strato­
spheric regions, depending on the type of test, the locationand the yield. The subsequent precipitation of the debris
and its deposit on to the earth is termed “local fallout”
when deposited locally, and “tropospheric fallout” and­“stratospheric fallout” when deposited globally [I9].
243. Local fallout can contain as much as 50% of the total
fallout produced in the case of above-ground tests, andincludes large radioactive aerosol particles that are deposited
within about 100 km of the test site. Tropospheric fallout
consists of smaller aerosols that are not carried across the

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UNSCEAR 2008 REPORT: VOLUME I

tropopause after the explosion and that deposit with a mean
residence time in the atmosphere of up to 30 days. During
this period the debris becomes dispersed, although not well
mixed, in the latitude band of the initial injection and following trajectories governed by wind patterns. From the viewpoint of human exposure, tropospheric fallout is importantfor nuclides with half-lives of a few days to two months,
such as 131I, 140Ba and 89Sr.

248. The short-lived radionuclide 95Zr (with its decay
product 95Nb) was the main contributor to external exposure
during active testing. Of the radionuclides contributing toexternal exposure, only 137Cs has a half-life of greater than a
few years, thus it became the most important contributor toannual doses after approximately 1966. At present it is the
only radionuclide contributing to continuing external­exposure from deposited radionuclides.

244. Stratospheric fallout, which makes up a large part of
the total fallout, consists of those particles that are carried
up into the stratosphere, disperse and later give rise toworldwide fallout, the major part of which occurs in thehemisphere of the initial injection. Stratospheric fallout
accounts for most of the worldwide residues of long-lived
fission products. The exposure of humans to fallout comprises internal irradiation (inhalation of radioactive material
in surface air and ingestion of contaminated foodstuffs) andexternal irradiation from radioactive material present in surface air or deposited on the ground [I9]. Atmospheric pro­
cesses related to dispersion and deposition of nuclear test
fallout were comprehensively reviewed in the UNSCEAR
2000 Report [U3].

249. Several radionuclides contribute to exposure via theingestion pathway. For the short-lived radionuclides (131I,
140
Ba, 89Sr), the exposures occur within weeks or months following deposition. Further exposure via ingestion of longerlived radionuclides comes from 55Fe and the transuranicelements. Committed doses due to the transuranic radionuclides are very low and the contributions to annual doses
negligible. During active testing, 137Cs was the most significant component, owing to its more immediate transfer to dietand subsequent delivery of dose. Because of the continuingtransfer of the long-lived 90Sr to diet, as well as the longer
retention of 90Sr in the body, this radionuclide became themost important contributor to dose beginning in about 1967.
The short-lived radionuclides have been relatively insignificant contributors to ingestion exposure. Important contributors to inhalation exposure were 144Ce, the transuranic
radionuclides, 106Ru, 91Y, 95Zr and 89Sr. Deposition (and thus
concentrations of these radionuclides in air) decreased rapidly after atmospheric testing ceased in 1980. Even for the
long-lived transuranic radionuclides, inhalation exposure
became insignificant after 1985.

(i)

Doses from global fallout

245. The basic input for calculations of doses due to fallout
radionuclides has been the measured deposition density of90
Sr. The measured annual hemispheric deposition for representative middle-latitude sites is given in table 35. Generalprocedures for deriving dose estimates from the measured or
calculated deposition densities of radionuclides were
described in detail in reference [U3], and only a summary of
the main conclusions from previous reports will be presented
here for completeness.
246. Estimates of the total annual effective doses due toradionuclides produced in atmospheric nuclear testing aresummarized in table 36, and the variation with time of theaverage per caput effective doses from nuclear weapons
fallout is presented in figure XXVII. These results are forthe average deposition of fallout radionuclides weighted
according to hemisphere and the world population. Dosesfor specific regions of the world can be obtained by adjusting these results for the latitudinal distribution of deposition­determined from 90Sr measurements.
247. The estimated global average annual per caput
effective dose due to atmospheric nuclear weapons testing
was highest in 1963 (0.11 mSv) and subsequently declined
to less than 0.005 mSv in the 2000s. External exposure
generally made the largest contribution to annual doses;
initially it was due to short-lived radionuclides and subsequently to 137Cs. The annual doses at present are due
almost equally to external exposure (53%) and internal
exposure due to ingestion (47%). The dose from 14C (30%of the total) now exceeds that from ingestion of otherradionuclides [U3].

250. One further contribution to the annual exposure comesfrom the globally dispersed radionuclides 3H and 14C. For
both radionuclides, there is no external exposure component
and only negligible exposure from inhalation; exposure
arises almost entirely from ingestion. The long-lived radio­isotope 14C is the dominant contributor, accounting for 70%of the total effective dose commitment to the world population. However, if only 10% of the 14C dose commitment isincluded in the comparison, i.e. if dose commitments aretruncated approximately to the year 2200 (by which time allother radionuclides will have delivered effectively all of theirdoses), 14C contributes only 19% to the truncated effective
dose commitment to the world population. About one quarter of the collective dose will have been delivered by the year
2200. The global estimates include a contribution from thedoses to people close to the sites used for atmospheric tests.
Although this contribution is small in global terms, somelocal doses were substantial [I9].
(ii)Local and regional exposures251. Local fallout can constitute as much as 50% of thetotal produced by surface tests and includes large radioactive
aerosol particles that are deposited within about 100 km of
the test site [I9]. A summary of the estimated yields in different atmospheric layers was shown in figure XXVI. Since
atmospheric nuclear weapons tests were conducted inANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

relatively remote areas, the exposures of local populations
did not contribute significantly to the global collective dosefrom this practice. Nevertheless, individuals living downwindof the test sites received higher doses than average.
252. Areas within a few hundred kilometres of the test site
are generally designated as “local” and those within a few
thousand kilometres as “regional”. A detailed description of
the main characteristics of all tests can be found in reference[U3]. The locations of the main test sites are shown infigure XXVIII.
253. Nevada test site (United States test site). The Nevada
Test Site (NTS) in the United States was the location for
86 atmospheric nuclear tests, carried out from 1951 to 1962.
In addition, 38 of the approximately 800 underground tests
involved releases of radioactive material. Although small incomparison with releases from the atmospheric tests, they
were sufficient to be detected off-site [S22]. Additional­cratering tests also injected debris into the atmosphere. Relatively few underground tests led to releases that affected
local areas [U3].
254. Estimates of external exposures due to atmospherictests at the NTS were derived from survey meter and film
badge measurements for 300 communities in the local areas(at distances of less than 300 km) around the test site in
Nevada and in south-western Utah. The effective dose
exceeded 3 mSv in 20% of the population of 180,000. The
highest effective doses were in the range 60–90 mSv; thepopulation-weighted average was 2.8 mSv. Exposuresresulted primarily from short-lived gamma emitters (withhalf-lives of less than 100 days). The collective external
whole-body dose within the 300 km closest to the NTS was
about 500 man Gy, and 12,000 man Gy for the area within
about 800 km of the test area [S22], arising primarily fromthe exposure of areas with large populations.
255. Internal exposures resulting from atmospheric testingat the NTS were estimated from deposition measurements
using an environmental transfer model. Absorbed doses toorgans and tissues from internal exposure were substantiallyless than those from external exposure, with the exception of
the thyroid, to which 131I from the ingestion of milk contributed relatively higher doses. Estimates of absorbed doses tothe thyroid in 3,545 locally exposed individuals ranged from0 to 4.6 Gy, with an average of 0.098 Gy. Mean thyroid doses
for residents of Utah, Nevada and Arizona were estimated tobe 0.17, 0.05 and 0.012 Gy, respectively [U3].
256. Bikini and Enewetak Atolls, Marshall Islands (UnitedStates test sites). In 1946, Bikini Atoll was the first site in the
Marshall Islands to be used for nuclear weapons testing by
the United States. In 1948, Eniwetok, a neighbouring atoll,
replaced Bikini as the test site. In 1954, Bikini was reactivated as a test site and was used until nuclear weapons testing in the Marshall Islands was ended in 1958. Bikini Atoll
was the site of 23 of the 66 tests, which were conductedunder water, at ground level and above ground. The yields of257

the tests at Bikini Atoll amounted to about 72% of the total
yield for the two test sites in the Marshall Islands.
257. Bikini Atoll, located 850 km north-west of the capitalof the Marshall Islands, Majuro, comprises more than
23 islands and islets. Bikini, Eneu, Nam and Enidrik Islands
account for over 70% of the land area. Bikini and Eneu are
the only islands of the atoll that have had a permanent population. Before nuclear weapons testing started, the population of Bikini Atoll (at that time 167 people) was evacuatedand resettled.
258. The test resulting in the most significant local exposures was the thermonuclear test Castle Bravo on 1 March
1954 at Bikini Atoll. Unexpectedly heavy local fallout
occurred east of the atoll owing to a sudden and unusualchange in wind direction, predominantly from the west
rather than the east, on the day of the test, and an unexpected
increase in fission yield. High radiation doses were received
by the inhabitants of Rongelap Island (67 persons, including
three in utero), about 210 km from Bikini Atoll, and by someRongelap islanders temporarily residing on Ailinginae Atoll,
about 150 km away (19 persons, including one in utero).
Further east, exposures occurred at Rongerik Atoll (28 UnitedStates servicemen) and Utirik Atoll (167 persons, including
eight in utero). These individuals were evacuated within a
few days of the initial exposures [I9, U3].
259. Effective doses as a result of external exposures,
mainly from short-lived radionuclides, ranged from 1.9 Sv
on Rongelap Island and 1.1 Sv on nearby Ailinginae Atoll to
0.1 Sv on Utirik Atoll. The collective effective dose wasabout 160 man Sv [I9]. Equivalent doses to the thyroid,caused by several isotopes of iodine and tellurium and by
external gamma radiation, were estimated to be 12, 22 and52 Sv on average, and 42, 82 and 200 Sv maximum, toadults, nine-year-old and one-year-old children, respectively, on Rongelap Island. Exposures due to residual radiation on Utirik and Rongelap Atolls of residents who returnedto these islands in 1954 and 1957, respectively, were of theorder of 20–30 mSv from external irradiation and
20–140 mSv from internal exposure over the subsequent
20-year period.
260. External exposure of the servicemen on Rongerik
Atoll due to the Castle Bravo test was 0.8 Sv. The Japanesefishing vessel Lucky Dragon was also in this area at the timeof the test, and 23 fishermen were exposed. Their external
exposures from fallout deposition on deck ranged from 1.7
to 6 Sv, mostly received on the first day of the fallout but
continuing for 14 days until the ship returned to its port.
Thyroid doses to these fishermen were estimated at 0.2–
1.2 Gy due to 131I on the basis of external counting; however,since other short-lived iodine isotopes were also present,
total doses to the thyroid due to inhalation over a period offive hours were estimated to have been 0.8–4.5 Gy [U3].
261. No other tests seem to have resulted in significant
exposures to the population in the Pacific region, even

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UNSCEAR 2008 REPORT: VOLUME I

though press and other official spectators did observe thetwo Crossroads explosions, in 1946, from relatively shortdistances. Military and test personnel probably received
some exposure from handling radioactive debris during
clean-up operations [S22].
262. In 1968, following radiological surveys that had been
carried out since 1958, resettlement of the Bikinian people
on the atoll was approved, and in 1969 the atoll was clearedof debris. Fruit trees, including coconut, breadfruit, pandanus, papaya and banana, were replanted. Eventually,
139 Bikinians resettled there. Further radiation survey andsampling programmes showed, in 1978, a tenfold increase in
the body content of 137Cs for the inhabitants of Bikini Atoll;
this was mainly due to increased consumption of coconut
fluid for lack of adequate supplies of freshwater. In responseto the high uptake of caesium in the population, the residents
were again relocated [I9]. During the temporary resettlementof Bikini Atoll from 1971 to 1978, total whole-body­exposures were estimated at 2–3 mSv/a [U3].
263. Johnston Island (United States test site). The UnitedStates used Johnston Atoll, located about 1,330 km southwest of Honolulu, Hawaii, as a launch site for 12 high-­
altitude nuclear tests beginning in 1958. All tests wereintended as airbursts, but three resulted in unintended nonnuclear destruction that led to contamination of the atollwith radioactive debris. The contamination was primarily
in the form of particulate debris, much of it being metalfrom the rockets accompanied by considerable amounts of­fissionable ­plutonium and/or uranium.
264. The atoll had been a United States military installation for several decades and currently is a wildlife sanctuary. There is no evidence of native populations ever having
lived on the atoll, and certainly none were present during
the years of nuclear testing. Hence there is no evidence thatmembers of the public within the immediate region were
exposed to the radioactive debris from the aborted tests. The
nine successful tests, because of their large yields and highaltitude of detonation, contributed mostly to global fallout,
as the closest populated islands would have been those of
Hawaii [S22].
265. Amchitka Island (United States test site). The threetests on Amchitka Island, Alaska, represent 15–16% of thetotal effective energy released during the United States
underground nuclear testing programme from 1951 to 1992.
Long Shot was detonated at a depth of 716 m in 1965, Milrow was detonated at a depth of 1,220 m in 1969, and­Cannikin, the largest United States underground nuclear test,
was detonated at a depth of 1,790 m in 1971 [D2].
266. Christmas Island and Malden Islands, Kiribati(United States and United Kingdom test sites). Christmas
Island and the Malden Islands in Oceania were used by theUnited States and the United Kingdom for testing nuclear
devices. Both islands are now part of the Republic of­Kiribati. The land area of Christmas Island is about 390 km2

and its 1990 population was about 2,500. There were six
British nuclear tests on Christmas Island in the period
1957–1958 and 24 United States tests in 1962 in the vicinityof the island [H7].
267. Nearby Malden Island is an uninhabited atoll today,and has been so since the British occupation in 1956. There
were three British nuclear tests near Malden Island. The
tests in the Pacific at Malden Island and the Christmas
Islands were airbursts over the ocean or explosions ofdevices suspended from balloons at 300–450 m over land[U3]. Local fallout would have been minimal following
these tests. Little or no information is available on exposureof the public or of civilian test personnel at either site,
although Fijian troops that participated in the tests and afterwards were involved in clean-up operations made claims
related to these events [S22].
268. Monte Bello, Emu and Maralinga, Australia (United
Kingdom test sites). The United Kingdom nuclear weapons
testing programme included 21 atmospheric tests at sites inAustralia and the Pacific. Twelve tests were conducted
between 1952 and 1957 at three sites in Australia: the Monte
Bello Islands, Emu and Maralinga. The Maralinga tests
included seven nuclear explosions and hundreds of minor trials involving chemically generated explosions of radio­active
material. Tests conducted at the Emu site, about 200 km northof Maralinga included two nuclear explosions and fivesmaller-scale experiments in 1953. These tests in continental
Australia led to residual radioactive contamination of the twoareas, covering some hundreds of square kilometres in total[H7]. These were mainly surface tests, with yields of 60 kt orless. Trajectories of the radioactive cloud were determinedfor each of these tests, and local and ­countrywide monitoringof air and deposition were performed.
269. Estimates of local external exposures were not madefor the earlier tests; for the tests in 1956 and 1957, theexternal effective doses were less than 1 mSv. The numbers
for local populations were not indicated [U3]. Estimates ofinternal exposures were also made for the overall Australian population. The average effective dose was 70 µSv,
83% of which was due to internal exposures, and the collective effective dose was 700 man Sv for the overall population of Australia [S22]. A number of safety tests conductedat the Maralinga and Emu sites in South Australia resulted
in the dispersion of 239Pu over some hundreds of square
kilometres [U3].
270. Semipalatinsk, Kazakhstan (Soviet test site). The
Semipalatinsk test site is located in the north-east corner of
Kazakhstan, 800 km north of the former capital Almaty,
400 km east of the present capital Astana and about 200 km
south-west of the border with the Russian region of Altai.
At this site, 456 nuclear tests were conducted, including
86 atmospheric and 30 surface tests. Five of the surface
tests were not successful and resulted in dispersion of plutonium in the environment. The site covers about
19,000 km2. The local populations most affected lived

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

mainly in the ­Semipala­tinsk region of Kazakhstan (nowpart of the Ust-­Kamenogorsk region of Kazakhstan) andthe Altai region of the Russian Federation, east and northeast of the test site. Traces of radioactive contamination
were also found in southerly and south-easterly directions
after some tests [S22, U3].
271. The earliest tests were above ground (atmospheric andsurface) and were carried out in the northern technical area Š.
The centre of the first (surface) explosion historically isreferred to as “Ground Zero”. The 340 underground tests wereconducted in widely separated technical areas in the south(between 1961 and 1989) and east (from 1968 to 1989). This
total includes four cratering nuclear explosions where theexplosive charge was placed at a shallow depth below ground.
Chagan was the first and largest of these tests. It resulted in alake about 0.5 km in diameter and 100 m deep, with cliffs up
to 100 m high, called Lake Balapan, or the “Atomic Lake”. Amuch smaller lake was formed by the Tel’kem-2 test. Of thetests carried out deep underground, 13 resulted in the releaseof radioactive gases to the atmosphere.
272. The only settlements within the nuclear test site during the 40-year test period were the town of Kurchatov, northof technical area Š (built for servicing the test site), and thesmall settlements of Akzhar and Moldari along its northernedge. Two tests led to the most significant exposures of the
population of Kazakhstan: the first test, on 29 August 1949,
and the first thermonuclear test, on 12 August 1953. Theseand two additional tests (24 September 1951 and 24 August1956) are stated to have contributed 85% of the total collective effective dose from all tests combined. The accumulated
effective doses for several districts were in the range 0.04–
2.4 Sv. The collective effective dose for ten districts was estimated to be 3,000–4,000 man Sv. Representative average
doses for seven villages close to the site (in Kazakhstan)were estimated to be 0.2–900 mGy for whole-body exposureand 0.3–3.8 Gy for thyroid exposure. Absorbed dose to thethyroid from ingestion of radioiodines is quite uncertain, but
may have been as high as 8 Gy for children in the Akbulak
settlement [I10, S22, U3].
273. Novaya Zemlya, Russian Federation (Soviet test sites).
Novaya Zemlya is an island located at the most northerly edgeof Europe. Soviet testing on Novaya Zemlya began in 1955.
Novaya Zemlya was the site of the world’s largest nuclear
weapons test, a 50 Mt detonation at an altitude of about
3.5 km. In all, 91 atmospheric nuclear tests took place on
Novaya Zemlya, and tests performed on the island account for
about one half of the total energy yield of all nuclear tests carried out in the entire world. Only one test, in 1957, was conducted directly on the ground surface. In addition, there were
two tests on the water surface and three tests under water at
the site. There were also 17 underground tests that vented, inmost cases resulting in on-site contamination only.
274. The nearest village, Amderma, is 280 km away, andthe much larger population centre of Arkhangelsk is approximately 1,000 km away. Three villages lie at intermediate

259

distances [S22]. Very little information is publicly available
concerning the local doses resulting from those tests. It is
likely, however, that doses to local residents were relatively
low, as most of the atmospheric devices were exploded athigh altitude so that the expanding fireballs did not touch theground surface. Preliminary information has been presented
in the open literature concerning external radiation doses at
the regional scale. The average external dose for the population of the eastern part of the Russian Federation (35 million) due to regional fallout in the years 1955–2000 is about
1 mSv [L24]. Concerning ingestion exposure, it is known
that 137Cs is abundant in lichen, reindeer and other environmental media. The 137Cs concentrations in reindeer meat aremuch greater than those in milk, fish, geese or ducks, andreindeer herders are likely to receive much higher internal
doses than the urban residents in the area, who consume
reindeer meat only occasionally. The estimated internal dose
due to 137Cs (and to a lesser extent to 90Sr) for reindeer herders has averaged about 1 mSv annually since the early 1960s;
average annual doses to urban residents are estimated to belower by a factor of 100.
275. Kapustin Yar, Russian Federation–Kazakhstan (Soviet
test site). Kapustin Yar is located 250 km north-west of the
Caspian Sea. Soviet testing at Kapustin Yar began in 1957.
In all, 10 atmospheric nuclear tests took place at Kapustin
Yar over six years. Very little information is publicly available about exposures resulting from the nuclear tests
launched from Kapustin Yar. All the Kapustin Yar tests werehigh-­altitude explosions (10.4–300 km), which in general­contribute more to global fallout than to local fallout [L25,S22].
276. Reganne and In Ecker, Algeria (French test sites).
Between 1960 and 1966, France conducted a series of fouratmospheric and 13 underground nuclear tests at Reganneand In-Ecker, remote sites located in the south of Algeria.
The French nuclear testing programme began with four lowyield surface tests in 1960 and 1961 at a site near Reganne in
the Algerian Sahara, about 50 km south-east of Reganne (avillage/oasis of a few thousand inhabitants) and about 150 km
south of Adrar, a city with approximately 50,000 inhabitants.
No information was found regarding local exposures following these tests. It is claimed that 15 people were probably
contaminated when radioactive vapour and aerosol escaped
through a fissure in the rock during a test in May 1962 thatwas performed under adverse wind conditions. Nine soldiers
received about 600 mSv, mainly due to external irradiation(>90%) [S22, U3]. No early radiological or clinical effects
were observed [B12]. Some residual contamination remains
at both this site and a nearby site, In-Ecker, where 13 underground tests were conducted. Small quantities of plutonium
were dispersed at these sites from safety experiments, which
involved conventional explosives only. No information has
been located on estimates of doses to the public from the tests
conducted in Algeria by France [I32].
277. Mururoa and Fangataufa (French test sites). The
Mururoa and Fangataufa Atolls in French Polynesia, situated260

UNSCEAR 2008 REPORT: VOLUME I

in the South Pacific Ocean, have evolved from extinct submarine volcanoes, and each rests upon a massive igneous
volcanic basalt substratum capped by a sedimentary carbonate coral reef platform hundreds of metres thick and surrounded by ocean water thousands of metres deep. France
conducted 193 nuclear experiments above and beneath theatolls of Mururoa and Fangataufa between July 1966 andJanuary 1996. Of these, 178 were nuclear tests, in which
nuclear devices were exploded with large releases of fission
energy, and 15 were safety trials. Forty-one were atmospheric tests (37 at Mururoa Atoll and 4 at Fangataufa Atoll,
between July 1966 and September 1974), and 137 were
underground nuclear tests (127 at Mururoa Atoll and 10 at
Fangataufa Atoll, between June 1975 and January 1996). Of
the 15 safety trials, all of which were carried out at Mururoa
Atoll, 5 were atmospheric and 10 were underground safety
trials [I12].
278. The atmospheric nuclear tests were mostly carried out
at a detonation altitude that was sufficient for the fireball not
to reach sea level, thereby minimizing the production of
local fallout. There were, however, four atmospheric nuclear
tests (three at Mururoa Atoll and one at Fangataufa Atoll) in
which the devices were mounted on barges floating in the
lagoon. Most of the residual radioactive material presently inthe accessible environment of the atolls was produced bythese nuclear tests. Five atmospheric safety trials were­conducted on the northern part of Mururoa Atoll.
279. The underground nuclear tests were conducted in thebasalt basement at depths of between about 500 and 1,100 m
in shafts drilled vertically beneath the rims of the lagoons.
Much of the residual radioactive material associated with theunderground nuclear tests was trapped in molten basalt rockthat solidified as glass-like lava, but some radionuclides were
deposited on fractured basalt rock that collapsed into thecavity-chimney and remained available for exchange withwater in the cavity-chimney. The ten underground safety trials were carried out in shafts drilled vertically beneath therim on the north-eastern part of Mururoa Atoll. The three
underground safety trials that involved some fission energy
release took place in carbonate formations at depths in excess
of 280 m [I12].
280. The closest inhabited atoll was Tureia (population
140) at a distance of 120 km to the north; only 5,000 persons lived within 1,000 km of the test site. A larger population (184,000 in 1974) was located 1,200 km to thenorth-east, at Tahiti. Under the conditions that normally prevail at the test site, radioactive debris of the local and tropospheric fallout was carried to the east over uninhabitedregions of the Pacific. On one occasion, however, material
was transferred to the central South Pacific by westerly
moving eddies within a few days of the tests. French scientists have identified five tests where regional populationgroups were more directly exposed. A single rainout event
caused exposures in Tahiti after the test of 17 July 1974.
Exposures resulted mainly from external irradiation from
deposited radionuclides. Milk production on Tahiti issufficient for only ~20% of local needs, and consumption is
low in any case, which limited ingestion exposures. Estimated effective doses to maximally exposed individuals
from the five events combined were in the range 1–5 mSv inthe year following the test. A collective effective dose of
70 man Sv was estimated for all local ­exposures at this testsite [U3].
281. Lop Nor test site (Chinese test site). The Chinese
nuclear weapons testing programme was carried out at the
Lop Nor test site in western China; 22 atmospheric tests and
12 underground tests were conducted between 1964 and1988 [S22, U3]. Limited information is available in the literature on local deposition following the tests. External
exposures in cities or towns within 400–800 km downwindof the test site are estimated to average about 0.044 mSv,
assuming 80% indoor occupancy and a building shielding
factor of 0.8 [S22].
282. The adult thyroid dose estimates range from0.06 mGy in Taiyuan to 2.5 mGy in Lanzhou. Thyroid
doses of infants would have been about 10 times higher.
The average thyroid dose received by the Chinese population as a result of the tests conducted at Lop Nor was estimated to be about 0.14 mGy. Even though the averagedeposition density of 90Sr seems to have been lower in
China than in the rest of the northern hemisphere, internal
doses from 90Sr are estimated to be higher in China as aconsequence of the diet of the Chinese population. The
average effective dose resulting from intake of 90Sr was
estimated to be 0.27 mSv, most of this due to tests notc­ onducted on Chinese soil.
(b)

Underground tests

283. There have been 1,877 underground nuclear tests.
Some gaseous radionuclides were unintentionally vented
during a few underground tests, but available data are insufficient to allow an accurate assessment of the radiologicalimpact. The total explosive yield of the underground tests is
estimated to be 90 Mt, much smaller than for the earlier
atmospheric tests. The yields for the tests performed byIndia, Pakistan and the Democratic People’s Republic ofKorea (DPRK) are not included in this total. Although mostof the debris remains underground, it is a potential longterm source of human exposure. The total number of tests­performed by each country is shown in figure XXIX.
284. The most recent test prior to the Committee’s report
was performed by the DPRK, on 9 October 2006. Between
21 and 25 October 2006, elevated levels of atmospheric 133Xe
were observed in Yellowknife, Canada. The measurements
could not be traced back to known nuclear facilities, andapplying atmospheric modelling to backtrack the dispersion
shows that the amount measured is consistent (to within anorder of magnitude) with simple leak scenarios assumed fora low-yield underground nuclear explosion on the Korean
peninsula [S3].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

(c)

Nuclear weapons production

285. In addition to actual weapons tests, the installations
where nuclear material was produced and weapons fabricated were another source of radionuclide releases to which
local and regional populations were exposed. Some information on this practice was presented in the UNSCEAR 1993
Report [U6]. Especially in the earliest years of weapons production, pressures to meet production schedules and the lackof stringent waste discharge controls resulted in higher local
exposures than in later years. Also, at some sites, weapons
are now being dismantled.
(i)

United States

286. Nuclear weapons plants in the United States included:
Fernald, Ohio (materials processing); Portsmouth, Ohio,and Paducah, Kentucky (enrichment); Oak Ridge, ­Tennessee(enrichment, separation, manufacture of weapon parts, laboratories); Los Alamos, New Mexico (plutonium processing, weapons assembly); Rocky Flats, Colorado (manufactureof weapons parts); Hanford, Washington (plutonium production); and Savannah River, South Carolina (plutonium
production). There are many more sites at which such operations were conducted and where wastes were stored or disposed of. Estimates of historical releases of radioactive
material during different periods of operation of the nuclear
installations have been reviewed in reference [U3].
(ii)

Former Soviet Union

287. There are three main sites where weapons materials
were produced in the former Soviet Union: Chelyabinsk,
Krasnoyarsk and Tomsk. Relatively large routine releases
occurred during the early years of operation of these facilities. In addition, accidents contributed to background levelsof contamination and to the radiation exposure of individuals
living in the local and regional areas.
288. Chelyabinsk. The Mayak nuclear material production
complex is located in the Chelyabinsk region between thetowns of Kyshtym and Kasli near the eastern shore of Lake
Irtyash. Uranium–graphite reactors for plutonium production and a reprocessing plant began operating in 1948. Relatively large discharges of radioactive material into the Techa
River occurred between 1949 and 1956. The available information on exposures to the local population was summarized
in the UNSCEAR 1993 Report [U6]. The individuals mosthighly exposed as a result of the releases into the Techa River
were residents of villages along the river, who used the riverfor drinking water, fishing, waterfowl breeding, watering
livestock, irrigation of gardens, bathing and washing. In
April–May 1951, a heavy flood resulted in contamination of
the flood plain used for livestock grazing and hay making.
The collective dose to the most exposed population from1949 to 1956 was 6,200 man Sv, with an average individual
effective dose of about 300 mSv, ranging from 36 to

261

1,400 mSv [A7]. Doses due to external irradiation decreased
in 1956, when residents of the upper reaches of the river
were moved to new locations and the most highly contaminated parts of the flood plain were enclosed. For some inhabitants, however, the Techa River contamination remains asignificant source of exposure to the present day.
289. Krasnoyarsk. The Krasnoyarsk nuclear material production complex is located about 40 km from the city ofKrasnoyarsk. The radiochemical plant for irradiated fuelreprocessing began operation in 1964. In 1985, a storage
facility was put into service for spent fuel assemblies fromreactors in the Soviet republics of Russia and Ukraine. There
are plans to reprocess fuel from the civilian nuclear fuelcycle at the Krasnoyarsk site in the future.
290. Radioactive waste discharges from the Krasnoyarsk
complex enter the Yenisei River. Trace contamination canbe found along the river from the complex to the estuary,
about 2,000 km away. An estimate for the collective dose
resulting from radioactive discharges from the Krasnoyarsk
complex during 1958–1991 was about 1,200 man Sv [U3].
The most important contributor (70%) to this dose was fishconsumption. External exposure due to the contaminatedflood plain accounted for 17% of the collective dose. The
main radionuclides contributing to the internal dose due tofish consumption were 32P, 24Na, 54Mn and 65Zn. The main
contributors to the external dose (over 90%) were gammaemitting radionuclides, primarily 137Cs, 60Co and 152Eu. Individual doses varied over a wide range, from 0.05 to
2.3 mSv/a. The major portion of the collective dose (about84%) was received by populations living within 350 km of
the site of the ­radioactive discharges.
291. In 1992, the direct-flow reactors of the Krasnoyarsk
complex were shut down. This reduced considerably theamount of radioactive discharges to the Yenisei River, andthe annual collective dose to the population was decreased
by a factor of more than 4. Estimates of average annual
doses for the period 1993–1996 were 30 μSv for external
doses and 20 μSv for internal doses. With a local populationof 200,000, the annual collective effective dose is estimatedto be 10 man Sv.
292. Tomsk. The Siberian nuclear material production complex is located in the town of Tomsk-7, on the right bank of
the Tom River 15 km north of the city of Tomsk. The Siberian complex was commissioned in 1953. Radionuclides inliquid waste are discharged into the Tom River, which flows
into the Ob River. An estimate for the collective dose due toradioactive discharges from the Siberian complex between
1958 and 1996 is 1,200 man Sv [U3]. During the period
1990–1992, three of the five reactors of the Siberian Complex were shut down, reducing considerably the amount ofradioactive discharges to the Tom River and the annual collective dose to the population. The collective effective dose
was estimated to be 200 man Sv. The largest contributor(73%) to this dose was from fish consumption. The mainradionuclides contributing to the internal dose due to fish

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UNSCEAR 2008 REPORT: VOLUME I

consumption were 32P and 24Na. About 80% of the collective
dose was received by the populations living within 30 km of
the site of the radioactive discharges [U3].
(iii)

United Kingdom

293. The production of nuclear material and the fabricationof weapons began in the 1950s in the United Kingdom. The
work was continued for several years at sites such as Springfields (uranium processing and fuel fabrication), Capenhurst(enrichment), Sellafield (plutonium production reactors andreprocessing), Aldermaston (weapons research) and Harwell(research). Subsequently, work related to the commercial
nuclear power programme was incorporated at some of thesesites. In the earliest years of operation of these installations,radionuclide discharges were associated almost wholly withthe military fuel cycle.
294. Plutonium production reactors were operated in the
United Kingdom at Sellafield (two graphite-moderated, gascooled reactors known as the Windscale Piles) and later at
Calder Hall on the Sellafield site and at Chapelcross in
Scotland.
(iv)

France295. A nuclear programme in France began in 1945 withthe creation of the Commissariat à l’énergie atomique. The
nuclear research laboratory at Fontenay-aux-Roses began
activities the following year. The first experimental reactor
went critical in 1948 and a pilot reprocessing plant began
operation in 1954. A second experimental reactor was constructed at the Saclay centre. From 1956 to 1959, threelarger production reactors began operation at the Marcoulecomplex on the Rhône River. These gas-cooled, graphite-­moderated reactors operated until 1968, 1980 and 1984,
respectively. A full-scale reprocessing plant was built andoperated from 1958, also at the Marcoule site. Two more
plants to reprocess fuel from commercial reactors wereconstructed at La Hague in the north of France, being completed in 1966 and 1990. The systematic reporting of radionuclide discharge data may also reflect the reprocessing of­commercial reactor fuel.
(v)

China

296. The Institute of Atomic Energy was created in 1950.
The first experimental reactor was constructed in Beijing,
and a uranium enrichment plant was built at Lanzhou in
Ganzu Province in western China. A nuclear weapons development programme was initiated in China that led to the firstnuclear explosion by that country in 1964. The first nuclear
test was of an enriched uranium device. Plutonium production and reprocessing were conducted at the Jiuquan complex, also located in Ganzu Province. The production reactor
began operation in 1967 and the reprocessing plant in 1968.
Production and reprocessing also occurred in Guangyuan in

Sichuan Province, where larger installations were constructed. Weapons were assembled at the Jiuquan complex.
Assessments of exposures due to nuclear weapons production in China have been reported and doses to populations
surrounding specific installations have been estimated [U3].
This experience relates to the military fuel cycle, since­China’s commercial nuclear power programme started onlyin the 1990s.

2.
(a)Residues in the environment

Nuclear test sites

297. As described earlier, radioactive debris from an atmospheric nuclear weapons test is partitioned between the local
ground or water surface and the tropospheric and stratospheric regions, depending on the type of test, the locationand the yield. The subsequent precipitation or depositing of
the debris is termed “local fallout” when it is locally dispersed, and “tropospheric fallout” and “stratospheric fallout”
when globally dispersed.
298. Exposures due to global fallout were described earlier
in this annex. Local fallout can constitute as much as 50% of
the production for surface tests, and includes large radioactive aerosol particles deposited within about 100 km of thetest site. In some tests, the contributions to total fallout exposure of doses to people close to the sites have been substantial, and these sites must be considered actual or potential
sources of public exposure. This subsection focuses on
recent efforts towards estimating potential exposures associated with present and future occupation of former nuclear
test sites.
(i)

Maralinga and Emu

299. As a result of the nuclear weapons tests, residual
radioactive contamination in the Maralinga and Emu areas
covers some hundreds of square kilometres. The possible
exposures associated with present and future occupation ofthese areas would be mainly of local aboriginal populations,
who are likely to constitute the majority of future inhabitants of the areas. The migratory lifestyle of the aboriginal
people in the areas makes an assessment of population doses
uncertain, and only best estimates for doses to individuals
will be discussed here. The assessment has been limited toconsideration of the consequences of existing surface contamination. The consequences of the removal of activity
from the burial pits known to exist in the areas have notbeen ­considered [H7].
300. The possible exposure pathways foreseen are:− Inhalation of material resuspended from the ground,including both natural wind-driven resuspension andresuspension arising from mechanical ­disturbanceof both soil and fire ash;

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

− Ingestion of foodstuffs and associated soil (contamination of foodstuffs with soil and fire ash) andwater ingestion; special consideration of deliberate soil ingestion (a practice called “pica”) is also
discussed;− Contamination of sores and wounds;− External gamma irradiation due to radioactive
material on the ground;− Beta irradiation due to radioactive material on the
ground and on skin and clothing.
301. A further potential exposure pathway, the handling ofcontaminated objects and fragments, has not been includedin this assessment. Measurements have been made of thesecontaminated items, and doses resulting from prolonged
proximity to or handling of such items may be considerable.
There is, however, no information on the likelihood andduration of such exposures, and for this reason an ­assessmentof dose has not been attempted.
302. Doses are calculated to the aboriginal population having a semi-traditional lifestyle. It may be assumed that doses
to other groups will be lower, with the exception of persons
carrying out particular activities such as souvenir hunting forcontaminated fragments. There is also considerable difficulty in estimating individual doses realistically because of
the great variability in the radionuclide levels in different
areas. In areas contaminated by the atomic explosions (the“major trials”), the significant radionuclides currently areneutron activation products, principally 60Co and 152Eu, andfallout radionuclides, principally 90Sr, and 155Eu. More significant radionuclide levels remain as a result of the various
chemically triggered explosions (the “minor trials”).
303. The dose assessment for different contaminated zones,
identifying the critical groups and the most relevant radionuclides, is shown in table 37. The calculated doses assume
100% residence in the area over the period of a year and thatcaught food is obtained and cooked locally (for kangaroo, a
representative and site-independent average concentration
for the meat was used). There is therefore a degree of conservatism incorporated into the calculations, which is substantial for the smaller zones. A considerable range of annual
effective dose estimates exists, from 0.5 mSv in the area of
Emu–Totem I (at the limit of aerial detection of 137Cs) to
500 mSv at Inner Taranaki. As expected, the highest doses
would be incurred from occupancy in the regions immediately surrounding the test sites. Continuous occupancy insuch areas is very unlikely because of their small size. Considerably lower but still significant doses would be incurredat the outermost contour lines defined by aerial survey.
(ii)

Mururoa and Fangataufa

304. The aim of recent assessments of the situation at the
Mururoa and Fangataufa Atolls was to estimate the radiation
doses that people anywhere in the South Pacific would

263

receive due to the residual radioactive material alreadypresent in the accessible environment of Mururoa and Fangataufa and their surrounding waters. The main scenario
addressed was the release of residual radioactive material
currently underground at the atolls into the lagoons ordirectly into the surrounding ocean as a result of the normal
migration of the residual radioactive material through thegeosphere, modified by the hydrogeological effects of thenuclear testing. Particular attention was paid to three radionuclides of potential radiological significance—239Pu, 137Cs
and 90Sr—and additionally to 3H, which was a useful tracer
for validating models.
305. There are no records of previous permanent indigenous habitation of the Mururoa and Fangataufa Atolls,
although some intermittent habitation of Mururoa Atoll has
occurred. The study postulated hypothetical dwellers on theatolls eating largely local seafood and locally grown produce, and estimated the upper bound of doses that might beincurred if the atolls were actually to be inhabited. It also
provided a conservative estimate of the doses being received
by the present population of Tureia Atoll, the nearest inhabited land (about 130 km from the Mururoa and Fangataufa
Atolls).
306. The most important contributors to the overall radionuclide release rates were the 12 nuclear tests carried out atMururoa Atoll early in the nuclear test programme. In termsof activity, tritium dominated the early releases, but withactivity concentrations that were of no radiological significance. Since the tests, other radionuclides, including 137Cs
and 90Sr, have been effectively retained underground within
the basalt basement, most of their activity decaying and only
small amounts being released. Plutonium continued to bereleased over long periods of time but at very low rates. The
modelling predicted that concentrations of 137Cs and 239+240Pu
in the lagoon water would be unlikely to exceed present levels at any time in the future. Concentrations of 90Sr and 3H
could rise marginally above current levels, but only during
the next few decades. The dispersion of residual radioactive
material throughout the ocean will lead to long-term concentrations of some radionuclides, which will decrease to background oceanic levels beyond about 100 km from the atolls.
Thus at Tureia Atoll the predicted concentrations will bearound background levels [I12].
(iii)

Bikini307. In 1997, the official journal of the Health Physics Society, Health Physics, devoted a complete issue [H16] to theconsequences of nuclear weapons testing in the MarshallIslands. The information presented in this section is mainly
related to the prevailing radiological circumstances and theirimplications for the future habitability of Bikini Atoll. Currently the significant residual radionuclides from nuclear tests
that remain in the soil and the surroundings of the atoll are137
Cs, 90Sr, 239+240Pu and 241Am. These are found to varying
degrees in both terrestrial and marine environments. The

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UNSCEAR 2008 REPORT: VOLUME I

unique composition of coral soil, which is primarily calcium
carbonate with no clay, produces a pattern of availability toplants of 137Cs and 90Sr very different from that for whichmost data (which relate to aluminium silicate clay soils of theAmericas and Europe) are reported in the literature [R17].
308. Bikini Island, the primary island for habitation at
Bikini Atoll, has the highest concentrations of 137Cs per unit
mass of soil and vegetation in the atoll. The average 137Csconcentration varies over a considerable range among theatoll’s islands. The average 137Cs concentration in soil andvegetation on Eneu Island, the other main island of residence, is about 10–13% of that on Bikini Island. The 137Csconcentrations in soil on Nam Island and Enidrik Island (thetwo other islands large enough for possible residence) areabout 70% and 15%, respectively, of that on Bikini Island.
309. Concentrations of transuranic radionuclides (239+240Pu
and 241Am), and their ratios to concentrations of 137Cs and90
Sr, vary around the atoll, reflecting differences in the designof the nuclear devices detonated near the various islands. In
general, radionuclide concentrations decrease rapidly withdepth in the soil column, although there are exceptions inparts of some islands. The activities of radionuclides per unit
dry weight of soil on Bikini Island are shown in table 38. The
concentration of 137Cs in coconut reaches values up to
6,000 Bq/kg. Some other fruits, such as pandanus and breadfruit, have average 137Cs concentrations of about 4 and
400 Bq/kg, respectively. The 90Sr activities are less than 10%of the respective 137Cs activities in the relevant foodstuffs.
The activities of 239+240Pu and 241Am are even lower than the90
Sr activities [R17]. The results from resuspension studies
show that the average resuspension of surface soil is very
low, with resuspension factors ranging from 10‑10 to 10‑11 m‑1.
On the basis of the measured activity concentrations in soil,
the concentrations of 239+240Pu and 241Am in air are expected
to be very low, and consequently the expected contribution
to doses due to radiation exposure via inhalation pathways is
judged to be insignificant.
310. The residual radionuclides, 137Cs, 90Sr, 239+240Pu andAm, are present in the atoll’s lagoon, mainly in sediments
but also in water and biota. Caesium-137 is found in very
low concentrations in lagoon sediment, water and fish. Caesium compounds are generally highly soluble, and the majorpart of the original inventory of 137Cs in the lagoon has longsince dissolved and become mixed into the world’s oceans.
Strontium-90, which is chemically similar to calcium (amajor component of the coral soils as calcium carbonate),
competes with the very large quantities of calcium available
for uptake by and distribution in marine species. It is alsochemically bound in the growing coral and coral sediment,
and remains in the lagoon environment primarily in the carbonate matrix. Consequently, 90Sr is relatively unavailable tomarine life.

dose rate in air measured at 1 m above the ground varied
from about 0.01 to 5 mGy/a in studies conducted in August
1978. The values decay-corrected to 1999 would be about
60% of the 1978 values, i.e. from 0.006 to 3 mGy/a. Other
potential routes by which exposure could occur (such as
swimming or diving in the lagoon) have been analysed. The
contributions to dose via these pathways were found to be sosmall that they could be neglected in the general dose
assessment.
312. Assessments performed to evaluate the potentialcommitted doses to the population that might in future live
on Bikini Island have estimated the average annual effective
dose due to external gamma radiation, based on typical local
occupancy habits and decay-corrected to 1999, as 0.4 mSv.
The overall annual individual dose was predicted to be about
8.0 mSv for a low-calorie diet. For a high-calorie dietassumed to consist of both imported and locally derived
foods, a value of 4.0 mSv was estimated, and for a diet consisting of only locally derived foodstuffs, the overall annual
dose was estimated as 15 mSv. In practice, doses resulting
from a diet of locally derived foodstuffs are unlikely to be
incurred under the current conditions, as the present Marshallese diet contains (and would in the near future presumably continue to contain) a substantial proportion ofimported food, which is assumed to be free of residual radionuclides. The uptake of 137Cs into terrestrial foodstuffs
accounted for the largest fraction of the total estimated dose(table 39) [B34].
313. Transuranic radionuclides in the lagoon remain an
important potential source of radiation. There is evidence
that plutonium is indeed transferred from sediments into theaquatic ecosystem in small but measurable concentrations
through the action of biogeochemical processes. However,
the observed transfer of these radionuclides through themarine food chain to human foodstuffs is very low. The
available information further indicates that actions of severe
storms and hurricanes in the area over the past 40 years do
not appear to have mobilized or transported the transuranic
radionuclides to any significant extent [I9].

241311. The best estimate for the total inventory of 239+240Pu
and 241Am in Bikini Atoll sediments is 103 ± 25 TBq and
93 ± 10 TBq, respectively. On Bikini Island the absorbed(iv)

Semipalatinsk, Kazakhstan

314. Emphasis in this assessment is given to residual radioactivity from nuclear testing. As such, the main tests of interest are those that resulted in local fallout. These include thesurface tests, excavation experiments and three undergroundtests in which an unplanned venting of radioactive material
to the atmosphere occurred. In most areas outside the nuclear
test site, external radiation dose rates and activity concentrations in soil are similar to typical levels in other regions andcountries where no nuclear weapons testing has been carriedout. The estimated annual effective dose to persons outside
the nuclear test site due to residual radionuclides is 0.1 mSv
at most. Actual exposures are more likely to be of the orderof a few microsieverts per year, a dose rate very close to theglobal average due to fallout [I10].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

315. Over most of the test site there is little or no residual
radioactivity. However, the Ground Zero and the Lake
Balapan areas are exceptions and are heavily contaminated. The only on-site inhabitants during the testing programme were in the town of Kurchatov and in the small
settlements of Akzhar and Moldari along the northern edgeof the site. Recently there has been limited resettlement
within the area, mostly by semi-nomadic farmers and herders. There is some evidence that they have grazed animals
in both the Ground Zero and the Lake Balapan areas. It is
not known if there are any settlements close to the othercratering test sites.
316. Activity concentrations in soil are available for themost radiologically important radionuclides at most occupied locations off-site, but for few locations on site. Outside
the nuclear test site, the results of 137Cs measurements fromIAEA missions in 1993 and 1994 all fell within the range
5–100 Bq/kg. Most results were at the lower end of thisrange, which is typical of global average fallout levels.
Results for plutonium in soil fell within the range 0.2–7 Bq/
kg, measured in 1991 and 1992. (For perspective, concentrations of 239Pu in surface soil in south-central England as aresult of weapons fallout are in the range 0.5–1.7 Bq/kg.) An
exception to this is in the village of Dolon, where muchhigher plutonium levels (by a factor of up to 100) have beenrecorded.
317. The absorbed dose rates due to terrestrial sources outside the nuclear test site have been extensively measured andare shown in table 40. Taken together, the values represent
the results of a survey conducted between 1991 and 1994 ofapproximately 600 locations around the entire nuclear testsite perimeter. All nearby centres of population are believed
to have been included. The values measured outside the testsite are almost entirely within the range of dose rates due tonatural sources measured in different countries and reported
by UNSCEAR (0.024–0.160 µGy/h).
318. Measurements of activity from inside the nuclear
test site are scarce in comparison with the data available
for outside. The gamma spectrometry aerial survey undertaken in 1990 indicated that the absorbed dose rate overthe entire test site was within the range 0.07–1 µGy/h.
Measurements made at Ground Zero with survey metersindicated that the dose rate changed rapidly with increasing distance from the epicentre, such that values close tonormal background levels were indicated at distances of afew hundred metres. Similar variations were observed inand around the Lake Balapan crater. High levels of actinides and fission products are present close to Ground
Zero and Lake Balapan.
319. Low concentrations of artificial radionuclides in soil
from the vicinity of the main settlements suggest, however,that the local food chain is unlikely to be a significant pathway of exposure. A limited food-sampling programme supports this [I10]. Drinking water samples taken from local
wells outside the test site and one inside the test site indicated265that 137Cs and 90Sr concentrations were not significant. The
possible future contamination of ground­water owing to theleaching of radionuclides from underground tests must be
considered, however. Air sampling carried out during 1991–
1992 inside and around the test site by the former Soviet
Union indicated negligible airborne levels of 137Cs and239+240
Pu in Dolon and other villages.
320. External radiation exposure has been assessed frommeasurements of absorbed dose rates. Internal radiation
exposure from inhalation has been assessed on the basis ofactivity concentrations in soil and assumptions regarding
the levels of resuspended dust. The ingestion pathway has
been modelled using environmental transfer factors (representing transfer from soil to the food chain) and a typical
local diet. The ingestion of soil has also been assessed. The
estimated doses to adults, assuming continuous habitationof the area, are given in table 41. The exposure of children
has also been estimated, and in all cases the total annual
doses are lower than those for adults. The annual dose estimated to persons living in settlements outside the test site is
0.06 mSv, with a higher value of 0.14 mSv for Dolon.
Because of the conservative assumptions made in the assessment, these values are likely to be overestimates; a more
realistic estimate of the dose to an average person living inthe settlements is likely to be about one tenth of theseestimates.
321. Two exposure scenarios were considered for the
nuclear test site. The first assumes a group of visitors thatstay at the highly contaminated areas for one hour per day
and keep animals that take 10% of their feed from theseareas. The values in table 41 indicate the level of dose that asmall number of frequent visitors might receive. The external exposure pathway dominated the doses to visitors tothese areas. The second scenario considered potential future
settlement. The most pessimistic future scenario is one inwhich persons permanently inhabited the Ground Zero or
Lake Balapan areas and derived all their crops and animal
products from within these areas. The estimated potential
future doses to permanent inhabitants are also given intable 41. External exposure would be the main exposure
pathway for persons who might in the future permanently
inhabit these two areas, but ingestion would also make a significant contribution, owing to the production of food in thecontaminated areas. The estimated annual doses to permanent residents due to residual radioactivity on the site areabout 140 mSv [I10].
322. Recent surveys at the Semipalatinsk test site highlighted the high degree of variability in the radiostrontium
contamination. The highest values measured were associated with leakage from tunnels in the Degelen area, where
239 underground tests were performed, including one as partof the programme on peaceful nuclear explosions. It wasalso suggested that some 90Sr may be in a highly mobile form
and that 90Sr ingestion is a comparatively important pathway
of exposure compared with other radionuclide exposures at
the test site and in the surrounding areas [H25].

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(v)

UNSCEAR 2008 REPORT: VOLUME I

Novaya Zemlya, Russian Federation

323. Current dose rates in the Novaya Zemlya islands generally vary from 0.08 to 0.12 µGy/h, which is similar to therange observed in adjacent areas not used for testing andwhich essentially corresponds to natural background levels,
although in small areas much higher dose rates can bedetected. The internal dose rate due to 137Cs (and to a lesser
extent due to 90Sr) for reindeer herders is estimated to havebeen about 1 mSv/a since the early 1960s; dose rates tourban residents were estimated to have been about 100 times
lower [S22].
(vi)

Nevada, United States

324. Four areas in Nevada have been used under the UnitedStates nuclear test programme: the NTS, the Tonopah Test
Range, Project Shoal and the Central Nevada Test Area. The
NTS encompasses 3,496 km2 of land under the jurisdictionof the United States Department of Energy (USDOE). The
Tonopah Test Range was withdrawn from public use formilitary use in the 1940s. Since 1956, the Tonopah Test
Range has been managed by the USDOE and encompasses
1,606 km2 of land used for defence and related research,
design and testing activities. The Project Shoal Area waswithdrawn from public use for purposes of undergroundnuclear testing. The Project Shoal underground nuclear test
took place on 1963. The area is currently used by the UnitedStates Navy for testing and training for tactical manoeuvring
and air support. Subsequent to an underground test in 1968,
the withdrawal of public lands for the Central Nevada Test
Area has remained unchanged and the area remains under
the control of the USDOE. Cattle grazing and recreation are
the main uses of the area around this site.
325. Radioactive waste management and disposal operations began at the NTS in the early 1960s, and low-level,transuranic mixed and classified low-level wastes havebeen disposed of in selected pits, trenches, landfills andboreholes on the NTS. The NTS currently serves as a disposal site for low-level waste generated by USDOEapproved operators and also as a storage site for a limited
amount of transuranic mixed waste. The topography of the
NTS has been altered by historic USDOE actions, particularly underground nuclear testing. The principal effect of
testing has been the creation of numerous craters in Yucca
Flat and on Pahute and Rainier Mesas. Underground nuclear
testing has resulted in impacts on the physical environment
in terms of ground motion, disruption of the geologicalmedia, surface subsidence, and contamination of the subsurface geological media and superficial soils. Waste disposal operations have also resulted in surface disturbanceand the placement of material having long-term impacts on
the environment. Table 42 summarizes the baseline information on the residual radionuclide ­inventory at the NTS.
326. Most of the areas considered in the NTS are locatedwithin the Great Basin, an area from which no surface water

leaves except by evaporation. Streams in the area are ephemeral. Although precipitation is very low in the region, during
extreme precipitation events there is some risk of flooding
along arroyos and around playa lakes. Throughout the region,
springs are the only natural sources of perennial surface
water, but they are not used for human consumption. A considerable volume of groundwater, estimated at 2.7 × 109 m3,
is held in recoverable storage beneath the NTS and thes­ urrounding region.
327. Radioactive contamination of surface areas at the NTS
resulted primarily from the atmospheric testing of nuclear
weapons between 1951 and 1962. Additionally, safety tests
conducted at the surface between 1954 and 1963 resulted inradioactive contamination of the soil. More than 200 areasthat are controlled because of radioactive contamination
have been identified and mapped on the NTS.
328. More than 800 underground nuclear tests have been
conducted at the NTS. Underground testing has resulted
in unavoidable adverse impacts to portions of the land
and the geological and groundwater resources, making
them un­usable for most purposes. Pockets of radioactive
contamination surround each underground test location.
From data on the number and dates of the undergroundtests at the NTS, the total activity of radionuclides
remaining underground is estimated to be 1.1 × 10 19 Bq.
Much of this radioactive material remains captured in theoriginal cavity and thus is not available to leach into thegroundwater. The impacts of conducting subcriticalexperiments underground would be much less than thoseof nuclear testing, since no self-­sustaining fission chain
reactions occur and much less radioactive material isdeposited in the geological environment. As in the caseof nuclear testing, the radioactive ­material is captured
underground.
329. Underground nuclear testing has resulted in the contamination of groundwater in the immediate vicinity of anumber of tests. The quality of the groundwater has been
impaired, but only in these limited areas. No radioactive
contamination attributable to USDOE activities has been
detected in monitoring wells outside the NTS. Detection ofsignificant contamination is limited to underground testingareas on the NTS. Tritium-contaminated groundwater exists
in the subsurface as a result of past underground testing of
nuclear weapons performed within the NTS and at two offsite locations, the Project Shoal Area and the Central Nevada
Test Area. On the basis of the combined results of studies
performed by various authors, the estimated range of peaktritium concentrations at the area of uncontrolled use closest
to the NTS varies from 0.02 Bq/L at 150 years after the
beginning of migration to 1.4 × 105 Bq/L in 25–94 years.
The migration of tritium-contaminated groundwater fromthe test location at the Project Shoal Area could result inpeak concentrations ranging from 1 × 104 to 2.7 × 107 Bq/L
at the boundary of the controlled area between 71 and206 years after the test. No public water well currently exists
at this location.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

330. The environmental impacts related to the waste management programme are minor compared with those of theother programmes. Underground nuclear detonations create
underground cavities into which the soil and rock above thecavity then collapse. The final result is a crater on the surface. Low-level waste at the Area 3 Radioactive Waste Management Site is disposed of in subsidence craters formed
from past underground nuclear tests. Waste management
programme operations in Area 5 are more diverse and include
facilities for hazardous and mixed-waste management inaddition to low-level-waste management facilities. After
30 years of waste disposal operations, the USDOE has notdetected any contamination in groundwater monitoring wells
recently completed near this area.
(vii)

Reganne and In Ecker, Algeria

331. Though the Reganne site is at present not sealed off,
access to the area of the test sites has been and continues tobe restricted by military control. There are practically noroads leading to the Algerian test sites, making access very
difficult. A survey has recently been performed at the nuclear
test sites [I32]. External dose rate measurements were madeat 76 locations. A total of 25 environmental samples werecollected. While the number of dose rate measurements wasconsidered adequate, the number of samples collected andanalysed was somewhat small, in view of the size of theareas. Most of the areas at the test sites have little residual
radioactive material except: (a) the ground zero locations of
the Gerboise Blanche and Gerboise Bleue atmospheric tests
at the Reganne test site, where the areas that have elevatedexternal dose rates are only a very small part of the tracts
surveyed and are confined to distances of a few hundred
metres from the four individual ground zero points; and(b) at Taourirt Tan Afella in the vicinity of the E2 tunnel,
where at the opening of one of the partially confined underground tests an accidental release of fission products mixed
with molten rock took place and formed a large bed of­hardened lava.
332. Despite the preliminary nature of the sampling andinvestigation programme, all conclusions indicate that
present-day exposure rates do not justify a requirement
for intervention, in view of the current state of development of the region. However, if the economic conditions
change in the area, the requirement for intervention at the
Gerboise Bleue, Gerboise Blanche and E2 tunnel sites
should be reconsidered. At Reganne, for occasional visitors to the site, exposures to external radiation due toresidual radionuclides from the tests are likely to be low,
i.e. less than a few microsieverts per day, while the area
at Taourirt Tan Afella has been protected from public­intrusion by a security fence.
333. In addition to the above-mentioned sites, at theAdrar Tikertine experimental site, at In Ecker, plutonium
in fine particulate form was spread over a wide area. The
concentration of plutonium in sand was determined from a267

small number of samples that were not sufficient to be representative of the area and which therefore could not be
used for a detailed or precise evaluation of the inventory or
specific distribution of activity in the Adrar Tikertine area.
Nevertheless, the activity concentration of anthropogenic
radionuclides in those samples was generally below laboratory detection limits. Thus it is expected that the residual
surface contamination from the plutonium dispersion
experiments is unlikely to give rise to doses to nomadic
herders or their families exceeding 1 mSv/a [I32].
(viii)

Lop Nor, China

334. Lop Nor, located in central Asia in a vast desert region
in western China, was the location for 34 nuclear weapons
tests conducted between 1964 and 1988; of these, 22 were
atmospheric tests and 12 were underground. Little information is publicly available on doses received by the public orby test personnel in China. It is known, however, that the
trajectory of the cloud carrying radioactive debris was determined for each test. The Ministry of Public Health set up anationwide monitoring network for environmental radio­
activity in the early 1960s, but the Lop Nor test site has neverbeen opened to Western scientists and no information could
be located on present levels of contamination and public
exposure, although available information indicates that thesite was made a reserve for the highly endangered Bactriancamel [S22].
(ix) Amchitka, United States
335. Following a report stating that there was radioactive
leakage from the test site to terrestrial and freshwater environments, recent surveys determined tritium concentrations
in surface water in the range 0.41–0.74 Bq/L at the sites sampled, which included the reported leakage sites. Only at theLong Shot test site, where leakage of radioactive gases to thenear surface occurred in 1965, were higher 3H levels(5.8 Bq/L) still observed in 1997. The mean 240Pu/239Pu value
for all of the Amchitka samples was 0.1991, with values
ranging from 0.1824 to 0.2431.
336. The measured 3H levels and 240Pu/239Pu ratios in freshwater moss and sediments at Amchitka provide no evidenceof leakage occurring at the sites. Deviations from the mean
240
Pu/239Pu ratios for global fallout were observed in marine
algae, sediment and pooled Amchitka samples, and may
suggest another source of plutonium release to the marine
environment; however, uncertainties in analyses and environmental processes need to be fully assessed before any
firm conclusions can be drawn. These results do not necessarily mean that leakage from the Amchitka undergroundnuclear tests is not occurring or will not occur into the North
Pacific Ocean or the Bering Sea. Hydrogeological modelling predicts that leakage of 3H from the test sites into themarine water might be seen beginning 20 to 3,000 years
from now [D2].

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(b)

UNSCEAR 2008 REPORT: VOLUME I

Sites contaminated by non-nuclear tests

and stones may ricochet and be found lying on the surface
some distance from the targeted area [U20].

(i) War sites contaminated with depleted uranium
337. During the enrichment process for natural uranium,
the 235U fraction is increased from its natural level (0.72%
by mass) to 2% or more. The uranium that remains after
the enriched fraction has been removed has reduced concentrations of 235U and 234U. This by-product is known as
depleted uranium (DU). The 235U content in DU is depleted
to 0.2–0.3%, about one third of its original natural fraction[U17]. Since 234U is a lighter isotope, its concentration iscorrespondingly higher in fuel uranium and lower in DU
compared with natural uranium. The fact that DU has
lower concentrations of 235U and 234U than natural uranium
also means that DU is less radioactive than natural uranium. Only traces of isotopes beyond 234Th and 231Th in the
decay chain are present in DU, as the other decay products
have not had time to build up in significant quantities inthe time since the DU was originally produced. The total
specific activity of natural uranium is 25.4 Bq/mg, whilethat of DU is 14.2 Bq/mg. Table 43 gives the main physicalproperties of the three isotopes of uranium and compares
their relative abundance by mass and activity in naturaluranium and DU [U17].
338. DU has been used for both civilian and military purposes for many years. The civilian applications include uses
in radiation shielding or as counterweights in aircraft. DU isalso used for heavy tank armour. Armour made of DU ismuch more resistant to penetration by anti-armour munitions
than conventional hard rolled steel armour plate. Also, owing
to its high density, its high melting point and its property of
becoming “sharper” as it penetrates armour plating, DU isused in anti-tank munitions. DU is pyrophoric; on impact
against its target, a DU penetrator will ignite, breaking up
into fragments and forming an aerosol of particles (“DUdust”) that can ignite spontaneously in air [I24].
339. Both tanks and aircraft can fire DU munitions, withtanks firing larger-calibre rounds (105 and 120 mm) and aircraft firing smaller-calibre rounds (25 and 30 mm). Typicallythe DU round fired by A-10 aircraft has a conical DU penetrator, 95 mm in length and with a diameter at the base of
16 mm, fixed inside an aluminium jacket. The weight of one
penetrator is approximately 300 g [U20]. When the penetrator hits an armoured vehicle, the penetrator continues throughthe armouring while the jacket usually remains outside.
340. A typical burst of fire by an A-10 aircraft occurs for
2–3 s and involves 120 to 195 rounds. These hit the groundin a straight line, 1–3 m apart, depending on the angle of theapproach. Penetrators that either hit non-armoured targets or
miss targets will generally remain intact and become buriedin the ground. The depth depends on the angle of theapproach, the speed of the plane, the type of target and thenature of the ground surface. In clay soils, penetrators usedby A-10 attack aircraft may reach a depth of more than 2 m.
Conversely, penetrators hitting hard objects such as rocks

341. Normally 10–35% (maximum 70%) of the roundbecomes aerosol on impact with armour. Most of the dust
particles are less than 5 µm in diameter and can be dispersedin the environment, spreading according to wind direction.
The amount of dust produced is actually small, because thevast majority of DU munitions miss their targets or hit softtargets and remain intact. The dispersion of the DU dust
leads to resuspended activity in the air and subsequent deposition on the ground. However, such radioactive material
should be limited to within about 100 m of the target. In acombat situation, the main radiological hazard associatedwith DU munitions is inhalation of the aerosols created when
DU munitions hit an armoured target [U20].
342. Small penetrator fragments and DU dust are gradually transported into the upper soil layer by weathering
processes. Wind, rainwater or surface water flow may also
redistribute the dust. Mobilization of DU through the soil
profile and the possible migration of DU into groundwater
will depend on a number of factors, such as the chemistry
and structure of the surrounding soil, rainfall and hydrology[U20].
343. The alpha particles emitted by DU are very energetic
but have a very limited range in tissue. They can barely penetrate the external layer of the skin and hence do not pose ahazard in terms of external irradiation, but internal irradiation is an important consideration. Uranium is not generallytransferred effectively through food chains; therefore, in
environmental assessments, inhalation is the exposure pathway that usually merits primary attention. Processes such as
migration through the soil, deposition of resuspended material on to crops and transfer to groundwater may, however,be of interest in the longer term [I24].
344. The only exposure of concern may arise from externalbeta radiation to the skin if a penetrator is placed in a pocket
or is used as an ornament worn on a neck chain. This could
result in quite high localized radiation doses after someweeks of continuous exposure. Although there will not be
any radiation skin burns, erythema may occur. The resulting
gamma radiation exposure will be insignificant, of the sameorder of magnitude as natural radiation, at most [U17].
345. Although it has been suggested that DU from munitions remaining in Kosovo or other locations may migrate
to groundwater, the uranium concentration arising fromthis source would be undetectable compared with naturally
occurring concentrations in water. Oeh et al. [O3] measured
water samples and the urinary excretion of uranium in aregion of Kosovo where DU munitions were deployed. Morethan 1,300 urine samples from peacekeeping personnel andunexposed controls of different genders and ages were analysed. The urine measurements for 113 unexposed subjects
had a uranium excretion rate of 13.9 ng/d (geometric standard deviation (GSD) = 2.17). The analysis of 1,228 urine

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

269

samples from the peacekeeping personnel resulted in a geometric mean of 12.8 ng/d (GSD = 2.60). No DU could befound in any water samples, and there was no difference
between urine samples from persons potentially exposed andcontrols.

against a hard surface while the penetrator enters the target. Potential exposures arising from jackets are far lower
than from penetrators, because the jackets are made of aluminium rather than DU, of which they have only very low
levels [U17].

346. Metallic DU reacts chemically in the same way as
metallic uranium, which is considered to be a reactive material. Studies carried out on penetrators collected in Kosovo,
Serbia and Montenegro showed that ground impact caused
numerous fine cracks in penetrators. This favours subsequentcorrosion and dissolution [U17]. Corrosion occurs relatively
quickly when the penetrator remains in the ground and issurrounded by soil. A penetrator can be completely corrodedin the 25–35 years following impact. The corrosion products
may in turn dissolve and disperse in water. However, the rateof corrosion depends on the composition of the soil. If thepenetrator is lying on the ground surface, the corrosion rate
is significantly lower. However, the corroded uranium isloosely attached and easily removable. Consequently, if sucha penetrator is picked up, it could easily contaminate the skinand clothing of anyone handling it. Buried penetrators andjackets may inadvertently be brought to the surface in the
future through digging as part of soil removal or construction work. The corresponding exposures would then be thesame as for penetrators and jackets currently lying on thesurface.

350. It has been confirmed that DU munitions have beenused in several recent military conflicts, including the Gulf
War in 1991, the conflicts in Bosnia Herzegovina in 1994
and in Kosovo in 1999. It was probably also used in the 2003
Iraq war. Available estimates of the total munitions used ineach conflict are presented in table 44.

347. There have been reports that the DU in munitions contained small amounts of other radionuclides, such as isotopesof americium and plutonium as well as 236U. The presence of
these man-made radionuclides indicated that some of the
DU had been obtained from uranium that had been irradiatedin nuclear reactors and subsequently reprocessed, or resulted
from contamination of equipment in the processing plant
during the reprocessing of spent nuclear fuel [I24].
348. Doses to members of the public living in areas where
they could be exposed to DU munitions are very low [I24].
There are several possible pathways through which populations in these areas may be exposed to radiation emitted by
DU munitions. The most significant pathway is inhalation of
DU particles that have been resuspended either by the windor by human activities such as ploughing. Fragments of DU
can be brought to the surface during the construction ofhouses, roads, etc. Lumps of DU lying on the ground surface(either complete penetrators or penetrator fragments) can bepicked up by members of the public. Consequently, there isa possibility of people being exposed to external beta andgamma radiation and to internal radiation if dust from corroded DU or DU fragments enter the body. The surface radiation from DU includes beta and gamma radiation from its
decay product, 234Th. The external dose due to direct contact
with DU fragments has been estimated to be 2.3 mSv/h [F7,I24, U17, U20].
349. As mentioned above, the jacket is the non-DU partof a weapon projectile that encases the DU penetrator. The
projectile is designed so that the jacket stops upon impact

351. Kuwait. The 1991 Gulf War was the first conflict inwhich DU munitions were used extensively. The total numberof rounds expended in the Gulf War is estimated to be about
860,600, representing a total weight of DU of about 286 t[I24]. Of the 3,700 Iraqi army tanks destroyed during the
Gulf War, DU munitions accounted for only around 500.
352. A large number of DU munitions were stockpiled on
the United States military base of Camp Doha when a fire
broke out on 11 July 1991. After the immediate clean-up
operations, approximately 300 DU penetrators (corresponding to a total of 1,500 kg of DU) were found to be
missing. The area was fenced and access to it restricted. In
2001, remediation actions were conducted. There was evidence of the presence of DU in environmental samples, but
the concentrations of 238U were more than two orders ofmagnitude lower than the values observed in the soil prior
to remediation. A person spending several hours each day
working on the site could receive a dose of 7.7 µSv over thecourse of a year, mainly from inhalation of resuspended
material. Individuals using the area for recreational purposes would receive doses of about one sixth of this. Accessto the area remains restricted, and actual doses due to DU
to people working or spending time nearby would be lower
still [U24].
353. At the Military Hospital storage site, adjacent to thearea where contaminated tanks had been stored, some DU
was present in the top 5 cm soil layer. However, the highest
concentrations of 238U observed were only about two to fourtimes the value expected from the natural background levels
across Kuwait. A person who worked on this part of the site
could receive an annual dose due to DU of about 3.3 µSv,
almost entirely via inhalation of resuspended material.
Annual doses to members of the public using the area forrecreation would be less than 1 µSv. Doses to members of
the public making use of nearby facilities would be lower
still [U24].
354. The site of Um Al Kwaty is used to store several thousand Iraqi military vehicles destroyed during the war, among
them 105 tanks contaminated with DU. It is estimated thatthe tanks stored at the site have a total of about 1 t of DU
associated with them. The site also contains 366 heaps ofcontaminated soil from Al Doha that contain ash from thefire at Camp Doha, fragments of munitions and other

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metallic debris. The debris is estimated to contain about 1.5 t
of DU. Access to the site is currently restricted [U24].
355. DU rounds were used in an attack on a convoy of Iraqi
vehicles at Al Mutlaa, a major and expanding urban areawith a population of about 50,000. Vehicles destroyed in theattack have been removed and the road has been completely
resurfaced. None of the samples of either soil or vegetation
contained detectable amounts of DU, and the concentrationsof 238U in the soil samples were consistent with the values
expected generally in soil in Kuwait.
356. The Manageesh oilfields cover a very large area southwest of Kuwait City. During the Gulf War they were subjected to repeated air raids involving DU munitions. The
area as a whole is thought still to contain several hundred
unexploded landmines and cluster bombs. Access to thisarea is also restricted.
357. Overall, it cannot be excluded that fragments of DU
penetrators or entire munitions might still be found and collected by members of the public at locations in Kuwait where
DU munitions were used in the 1991 Gulf War. Prolonged
skin contact with these DU residues is the only possible
exposure pathway that could result in exposures of radiological significance. As long as access to the areas remains
restricted, the likelihood that members of the public could
pick up or otherwise come into contact with these residues is
low [U24].
358. Bosnia and Herzegovina. There are 15 target sites
confirmed by the North Atlantic Treaty Organization(NATO) in Bosnia and Herzegovina where DU munitions
were used, of which one is inaccessible because of the presence of mines. There are also six NATO target sites in thevicinity of Sarajevo for which the coordinates are still missing. These sites could therefore not be investigated. Three ofthe 14 sites investigated by the United Nations Environment
Programme (UNEP) clearly showed DU contamination,
confirming the earlier use of DU ordnance. No DU contamination was found at the other 11 sites investigated. None ofthese sites showed signs of widespread contamination of theground surface. Ground surface DU contamination was
typi­cally limited to areas within 1–2 m of penetrators andlocalized points of contamination caused by penetrator
impacts. Almost 300 contamination points were identified
during the mission, but most of them were only slightly
contaminated. Given that several thousand DU rounds werereportedly fired against the target sites investigated, thenumber found is low. It is possible that the majority of thepenetrators are buried deep in the ground [U18].
359. DU could be clearly identified in one of the drinking
water samples. A second drinking water sample from a wellshowed traces of DU contamination, which were detectable
only through the use of mass spectrometric measurements. DU
was found in lichen samples at the three sites mentioned above.
There are no reasons to expect the presence of any DU in food,owing to the low dispersion rate in the ground and the low

uptake factor in food. DU contamination in air was found at
two sites where DU use had been confirmed. The concentrations were very low, and the resulting radiation doses wereminor and insignificant. At distances of over 100 m from­contaminated areas, no DU could be detected in the air [U18].
360. Kosovo. During the Kosovo conflict in 1999, DU
weapons were fired from NATO aircraft; it has been reported
that over 30,000 rounds of DU were used. Because of therisks posed by mines and unexploded ordnance, the sites
investigated by UNEP in 2000 were limited compared withthe total area potentially affected by the use of DU in Kosovo
and represented some 12% of all sites attacked using DU
munitions during the Kosovo conflict [U20].
361. No significant widespread contamination of groundsurfaces or soil was found in Kosovo, although localized
points of concentrated contamination close to penetrator
impact sites or penetrator holes exist. The levels of DU
detected decreased rapidly with distance from impact points,
the maximum distance at which levels were still measurable
being 10–50 m. When a penetrator or a jacket was found on
the surface of the ground, the soil below the penetrator normally had measurable levels of DU. The area of the impact
point was normally small, i.e. less than 0.04 m2, but the relative concentration of DU at such a point could be high. The
absolute concentration of DU in soil varied from a few milligrams of DU per kilogram of soil to about 18 g of DU perkilogram of soil, which corresponded to about 6% of theweight of a penetrator.
362. The depth of soil beneath impact points with measurable DU levels was normally in the range 10–20 cm, with theactivity concentration decreasing with increasing depth. Thisvertical distribution probably resulted from the dissolution
and dispersion of DU following the initial surface contamination or from the penetrator lying on the surface. However,
the amount of DU at the impact points was very low and thecorresponding exposures insignificant.
363. The surface of penetrators was probably subject tooxidation, as part of the radioactive material was easily
removed from the oxidized surface. However, the amount
was very low, about 10‑5 of the mass of the penetrator, i.e. a
few milligrams. As in the case of the penetrators, the soil
beneath a jacket had measurable activity to a depth of
15–20 cm. The potential exposure to radiation arising fromthe jackets is much lower than from the penetrators, because
the jackets are not made of DU and are only slightly­contaminated [U20].
364. It is probable that many penetrators and jackets remain
hidden at some metres depth in the ground. No measurable
levels of DU were found in houses, vehicles or other objects.
Results on the levels in botanical material were not conclusive except for lichen (and possibly bark). No measurable
levels of DU were found in milk samples taken from cows
grazing in fields that potentially might have elevated levels
of DU [U20].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

271

365. Serbia and Montenegro. In Serbia, significant levels
of DU were found at localized points in the immediate vicinity of penetrators lying on the ground and around penetrator
impact marks/holes. The levels of DU detected decreased
rapidly with distance from such points, and beyond a distance of one metre were no longer detectable by field measurements. Laboratory analyses of soil samples, however,enabled activity levels to be traced for several metres further
from the points. More detailed laboratory analyses of soil
samples revealed widespread low levels of DU at five of thesix study sites [U17].

environmental contamination with DU, but contamination isstill anticipated to be found, as many of the destroyed Iraqi
tanks and armoured personnel carriers were hit by DU rounds,
normally 2–7 times per armoured vehicle. These vehicles are
therefore expected to have extensive DU contamination in theform of dust and large fragments [U19]. Urine analysis in
United States personnel who served in the conflict has been
inconclusive regarding exposure to DU [M24].

366. Localized points of increased activity can occur atsites of penetrator impacts or close to a penetrator that hasremained on the surface and been subject to corrosion. The
concentration of DU can be very high at these points, but theextent of the increased activity is very limited, normallywithin a radius of 1 m, and the total amount varies widely,
being in the range 0.01–10 g of DU per kilogram soil.
Beneath these points, the activity levels are measurable insoil down to a depth of 10–20 cm or more, with the activity
concentration decreasing with increasing depth [U20]. The
penetrators recovered had decreased in mass by 10–15%
because of corrosion. This has important implications fordecontamination approaches as well as for possible future
migration into groundwater. DU was not present in any of
the groundwater or drinking water samples [U17].

370. The Russian Federation inherited from the former
Soviet Union several thousand square kilometres of radio­
nuclide-contaminated land and some tens of petabecquerelsof radioactive waste. Environmental contamination began
and was particularly intensive in the early years of the“Atomic Project” activities initiated in the mid-1940s [V10].
At present in the Russian Federation, about 650 million
cubic metres of liquid and solid radioactive waste with a
total activity of approximately 7.4 × 1019 Bq (2 billion curies)
have been accumulated. In addition, approximately 12,000 t
of spent nuclear fuel, with a total activity of about 3 × 1020 Bq(8.2 billion curies), are kept at the sites of Minatom and otheragencies in the Russian Federation [L2].

367. Airborne DU particles were detected at two of the six
sites. While these particles may have become airborne fromon-site digging operations, the finding highlighted the possibility of exposure pathways associated with soil disturbance at DU sites. The overall exposure to DU decreases
with time as the exposure via airborne contamination fromresuspension of DU dust on the ground surface decreases
with time. On the other hand, the probability of DU migration in soil increases with time, owing to the corrosion of DU
penetrators [U17]. Many penetrators were found to be heavily corroded, and given a similar rate of corrosion, those penetrators still on the surface may have more or less disappeared
from the environment (as solid objects) within 10–20 years.
What happens in the case of penetrators buried deep in the
ground is not yet known.
368. Uranium concentrations were within the normal
range for uranium concentration in drinking water. The concentrations of uranium in air samples were also varied
within the normal range, even though they were in the upper
part of that range. The UNEP mission to Kosovo in 2000
found that lichen appears to be a bioindicator of airborne
DU contamination.
369. Iraq. At the time of writing, there were no conclusive
results publicly available from assessments of DU levels in
the environment in Iraq. Also, the amount of DU munitions
used and the sites of impact in the 2003 conflict are unknown.
No conclusions on the current situation regarding public
exposure due to DU in Iraq can be drawn at present. Preliminary surveys of “hot spots” in Iraq have not detected(ii)

Contaminated sites in the Russian Federation

371. The total land area contaminated with radionuclides
as a result of activities of the Minatom enterprises is about
480 km2. About 15% of the total area contaminated withradionuclides has gamma radiation exposure rates of above
2 µGy/h [L2]. More than 90% of this land, i.e. 65.7 km2, wascontaminated as a result of the accident at the Mayak complex in 1957 [V10]. The main sites and contaminated areas
are described in table 45. An area of about 0.26 km2 wasrestored in the period 1996–1999, and rehabilitation of
13.5 km2 of contaminated land is planned for the period
2001–2010 [L2].
372. At uranium ore mining and milling enterprises, morethan 300 million tonnes of solid waste (in dumps of barren
rocks and unspecified ores, etc.) and about 60 million cubic
metres of liquid waste (in tailings dumps) have accumulated
up to the present time. Their total activity (due to radio­nuclides
of uranium and its decay products) is about 7 × 1015 Bq. The
total area occupied by the dumps is 9.871 km2 [L2].
373. Chemical and metallurgical enterprises for nuclear
material and fuel element production have accumulated
more than 600,000 m3 of liquid radioactive waste and about
5 million tonnes of solid radioactive waste, containing radionuclides of uranium, thorium and their decay products withtotal activity of over 1.6 × 1014 Bq (4,200 Ci). The area ofland contaminated with radionuclides is 1.868 km2, including 0.464 km2 with exposure rates in the range 2–10 µGy/h(200–1000 µR/h).
374. In 1999 there were 50 operating nuclear research­reactors and critical or subcritical assemblies in the RussianFederation, 53 facilities whose operation had been suspended

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or that were in the process of decommissioning, and 6 facilities under construction. Spent nuclear fuel from theresearch facilities was concentrated mainly at the following
sites: the Russian Research Centre Kurchatov Institute; theInstitute of Physics and Power Engineering; the ResearchInstitute of Atomic Reactors; the Sverdlovsk Branch of the
Research and Development Institute of Power Engineering;
the St. Petersburg Institute of Nuclear Physics of the­Russian Academy of Sciences; and the Karpov Physicaland Chemical Research Institute’s branch in Obninsk. Theinterim storage facilities for spent nuclear fuel are 80%
filled on average [L2].
375. At the Mayak Industrial Association, studies werecarried out on Karachai Lake, which is being filled with soil.
Beginning in 1951, the lake was used for the discharge of
medium- and high-level liquid radioactive waste. Stage-bystage remediation of the water reservoir was started in 1988.
At present, as a result of the remediation actions, the averagearea of Karachai Lake has been reduced by a factor of over 3(to 100,000 m2), which has significantly reduced the emanation of radioactive aerosols from the water surface and shoreline and their subsequent transport by wind. This work issoon to be completed [L2].
376. At the Mining and Chemical Complex, two of thethree uranium–graphite production reactors have alreadybeen shut down. Many years of reactor operation led to theaccumulation of radioactive silts in cooling and storage
ponds, and also caused contamination of the Yenisei Riverflood plain. Contamination levels in the Yenisei flood plain
began to decline when the reactors with once-through cooling were shut down. Dose rates in the range 0.08–0.4 µSv/hhave been measured in populated areas along the Yenisei
River. As a result of the shutdown of these once-throughreactors, radionuclide discharges into the Yenisei River havedecreased by a factor of over 10, and the present exposure
rate at the water surface does not exceed allowable values,
even at the discharge point [L2].
377. Up to 2000, the Russian Navy had withdrawn 184
nuclear submarines from service. Of these, 108 were in thenorth-west part of the country (the Murmansk and Archangelregions) and 76 were in the Far East (Primorsk and the Kamchatka region). Spent nuclear fuel was not unloaded frommost of the submarines. A number of the nuclear submarines
were withdrawn from service over 10–15 years ago, anddefects in the vessels’ structures have appeared during thislong period afloat. The nuclear submarines with spent
nuclear fuel on board represent a serious potential radiation
hazard to the environment [L2, V10].
(iii) Contaminated sites in the United States
378. The main contaminated sites in the United States areusually related to the mining of uranium and of other products that have uranium associated with the ore (such as phosphate rocks), to the processing of monazite, to industries

dealing with radium, to fuel preparation for nuclear power
plants and to research institutions associated with defence
programmes.
379. The United States Environmental Protection Agency(EPA) coordinates a project aimed at identifying and cleaning up contaminated areas throughout the country. It haslisted 84 sites contaminated with radionuclides; of these, 61
are currently on the EPA’s National Priority List. Of these,
14 sites are directly linked with United States nuclear military programme operations (i.e. USDOE sites): Brookhaven
National Laboratory, New York state; Feed Material Production Center, Ohio; Hanford Areas 100, 200 and 300, Washington state; Idaho National Engineering Laboratory, Idaho;
Lawrence Livermore National Laboratories, California;
Monticello Mill Tailings, Utah; Mound Plant, Ohio; Oak
River Reservation, Tennessee; Paducah Gaseous Diffusion
Plant, Kentucky; Rocky Flats Plant, Colorado; SavannahRiver Site, South Carolina; and Weldon Spring, Missouri.
Four sites are related to the production of radium devices andproducts, and eight sites are related to NORM, mainly phosphate ore processing and heavy-metal smelting. About
25 sites have been contaminated by improper waste disposal
or by the use of waste as landfill; some of these sites are
inside military installations. The main concern for such sites
is related to the public exposure due to possible contamination of groundwater. For the other sites, the origin of theradioactive contamination is not clear, except for one site,
reported to have been contaminated as a result of radiopharma­
ceutical manufacture [E5]. A large national programmecalled the Superfund targets the clean-up of hazardous contaminated sites and is conducting recovery operations at
most of the sites listed; for some of them remedial operations
have already been completed.
(iv)

Contaminated sites in the European Union

380. The Dounreay nuclear site, located on the north coastof Scotland, United Kingdom, was responsible for the releaseof an unknown quantity of approximately sand-sized fragments of irradiated nuclear fuel during the late 1950s, 1960s
and 1970s. The first Dounreay hot particle to be formally
identified was recovered from the Dounreay foreshore in
1983. A further single particle was recovered from Sandside
Beach the following year. Particles have been detected andremoved from the Dounreay foreshore regularly since 1984
and from the offshore sediments since 1997. Over 1,200 individual particles have since been found in the littoral (intertidal) and marine environments in the vicinity of Dounreay,
including Sandside Beach (1 km west of Dounreay), theDounreay foreshore, Dunnet Beach and Murkle Beach (bothapproximately 25 km east of Dounreay), and in marine sediments adjacent to the Dounreay site. In addition, 86 particles
have been found on the Dounreay site itself (table 46).
381. Particles are detected in the environment by their 137Cs
gamma activity, but the total activity is dominated by thebeta emitters 90Sr and its associated 90Y. The particles were

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

produced during the reprocessing of fuel at Dounreay during
the late 1950s, 1960s and 1970s. Two main types of particle,
produced from Materials Test Reactor and Dounreay Fast
Reactor fuel, have been identified. Materials Test Reactor
particles, which make up ~80% of the total recovered, were
produced as a result of fault conditions during milling andcropping operations, prior to reprocessing. These milling
activities stopped at Dounreay in 1973. Dounreay Fast Reactor particles were most likely produced during combustion
incidents in the dissolution cycle during reprocessing. Several such incidents are known to have occurred between
1969 and 1972. Very few particles are found on publicly
accessible beaches, and those which are found are small andare promptly removed. Although the risks to members of the
public from the presence of particles in the environment aresmall, they are a problem of public concern [D5, T8].
382. No information has been found on sites in other countries of the European Union contaminated as a result of military activities, except for those related to former uranium
mining activities.
(v)

Dumping of radioactive waste in the sea

383. Radioactive waste has been dumped in the Arctic Sea,
the North Atlantic, the North Pacific and the West Pacific(figure XXX). At present, the total activity of waste dumped
in these regions is estimated to have decreased to a total of
about 4 × 1013 Bq. This information is contained in the relevant IAEA database [I11]. Doses to critical populationgroups in coastal areas of the Arctic, North Atlantic and Far
East regions of the Russian Federation due to the consumption of seafood products containing radionuclides wereshown not to exceed 10‑4–10‑3 of natural radiation ­backgroundexposure [L2].
384. Kara Sea [I11]. In 1992, it was reported that theformer Soviet Union had dumped radioactive waste in theshallow waters of the Arctic Seas for over three decades (figure XXXI). The International Arctic Seas Assessment Project(IASAP) was launched by the IAEA in 1993 with the objectives of assessing the current environmental situation associated with the radioactive waste dumped in the Kara andBarents Seas and examining possible remedial actions.
385. The total amount of radioactive waste dumped in the
Arctic seas was first estimated by the Russian Federation to
be approximately 90 PBq at the time it was dumped. Items
disposed at sea included: six nuclear submarine reactors
containing spent fuel; the shielding assembly from an icebreaker reactor, which contained spent fuel; ten nuclear
reactors without fuel; and solid and liquid low-level waste.
Of the total inventory, 89 PBq came from high-level wastecomprising reactors with and without spent fuel. Solid
waste, including the above reactors, was dumped in the
Kara Sea, mainly in the shallow fjords of Novaya Zemlya,
where depths at dumping sites ranged from 12 to 135 m, andin the Novaya Zemlya Trough, at depths of up to 380 m.

273

Liquid low-level waste was released into the open Barents
and Kara Seas. On the basis of reactor operating histories
and calculated neutron spectra, the estimate of the total
radio­nuclide inventory of the high-level radioactive waste at
the time it was dumped has been revised to 37 PBq. The corresponding inventory of high-level waste dumped at sea wasestimated to be 4.7 PBq in 1994, of which 86% were fission
products (main radionuclides 90Sr and 137Cs), 12% activation
products (main radionuclide 63Ni) and 2% ­actinides (main­radionuclide 241Pu).
386. The high-level radioactive waste dumped in the Kara
Sea and adjoining fjords was in discrete packages, which are
expected to leak at some time in the future. They therefore
constitute a potential chronic exposure source where theconcern relates to future increments of dose to exposed individuals. The open Kara Sea has relatively low levels of artificial radioactivity compared with some other marine areas.
Measurements of environmental materials suggest that theannual individual doses due to artificial radionuclides in the
Kara and Barents Seas are in the range 1–20 µSv.
387. In two fjords where both high- and low-level wastes
were dumped, elevated levels of radionuclides were detected
in sediments within a few metres of the low-level waste containers, suggesting that some had leaked. However, this leakage has not led to a measurable increase of radionuclides in
the outer parts of the fjords.
388. Calculations of individual doses were undertaken fortime periods covering the projected peak individual dose
rates for three scenarios and for the following populationgroups: (a) groups living in the Ob and Yenisei estuaries andon the Taimyr and Yamal peninsulas, with habits typical ofsubsistence fishing communities in other countries with Arctic coastlines; (b) a hypothetical group of military personnel
patrolling, for 100 hours in a year, the foreshores of thefjords containing dumped radioactive material; and (c) agroup of seafood consumers considered representative of thenorthern Russian population situated on the Kola Peninsula.
The calculated peak doses to members of these groups due toall sources are shown in table 47.

(vi) Accidental losses of radioactive material at sea
389. Besides the reported events of planned dumping ofradioactive material in the sea, there were also several events
that included the loss or the release of radioactive material in
the sea. These events are summarized in table 7 of annex Cof the UNSCEAR 2008 Report and include the following[I17]:(i) ix nuclear submarines have been lost since 1963 at
Svarious sites in the Atlantic Ocean: two from theUnited States Navy—Thresher in 1963 (one nuclear
reactor, 1.15 PBq) and Scorpion in 1968 (one nuclear
reactor, 1.3 PBq, and two nuclear warheads); three
from the Navy of the former Soviet Union—K-8 in

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1970 (two nuclear reactors, 9.25 PBq, and a nuclear
warhead, 30 GBq), K-219 in 1986 (two reactors,
9.25 PBq) and K-278 Komsomolets in 1989 (reactor
core, 3.59 PBq); and one from the Russian Federation—K-141 Kursk in 2000 (two nuclear reactors,
1–2 PBq). With the exception of the accident involving the Russian submarine Kursk, the depth at thesites of the accidents, below 1,500 m, has not permitted the recovery of the ­submarines or their nuclear
reactors.
(ii)

 uclear weapons have been designed to be carriedNon submarines, surface ships, aircraft and rockets.
There are seven recorded accidents that have resulted
in the confirmed loss of one or more nuclear
weapons.

(iii) T here have been four recorded accidental re-entriesof nuclear powered satellites and one recorded accidental re-entry of a spacecraft. Four of these accidents resulted in the actual or potential release ofradionuclides into the environment.
(iv) T here have been two recorded incidents whereradioisotope thermoelectric generators (RTGs)
have been lost at sea, both occurring near the eastern coast of Sakhalin Island in the Sea of Okhotsk
and both involving emergency disposals of the
RTGs during transport by helicopter. In the firstincident, which occurred in 1987, the RTG disposed of contained about 25.3 PBq of 90Sr. The second RTG was disposed of in 1997 and contained
about 1.3 PBq of 90Sr.
390. Sealed radiation sources are widely used in the marine
environment in association with oil and gas exploration andextraction. In some instances the well logging tool and drill
string containing the sealed source become stuck in the drill
hole and tool recovery is not feasible. The equipment is generally left in place and the hole is closed/sealed. This resultsin situations where radioactive material could enter themarine environment. In general, these losses have occurred
deep in the sediment. The nature of the containment as wellas the location of the loss are such that, in general, radio­
nuclide release could occur only after a long period of time.
The IAEA database on sealed radiation sources lost in thesea includes about 150 items [I17].
(vii) Other sources of public exposure391. Since the start of the space age in 1957, radiation
sources have been used on board spacecraft for power generation, for thermal control, and in subsystems and instruments (figure XXXII). While electricity for spacecraft has
predominantly been produced by photovoltaic cells, there
are occasions when, owing to mission criteria (e.g. highpower requirements, insufficient solar energy flux in deep
space or requirements for planetary landing), the use of solar
power is impractical. In such cases, nuclear power sources
have been used. To date, only the former Soviet Union, theRussian Federation and the United States have utilized
nuclear power systems in Earth orbit or beyond [U44].
392. The United States launched one thermoelectric reactor in 1965. The reactor was shut down after 43 days of operation and placed in a long-term “storage” orbit (an orbit withan estimated orbital decay time of longer than 400 years).
The former Soviet Union launched 31 thermoelectric reactors between 1970 and 1988. Their lifetimes ranged from 0.1
to 293 days. Two thermoionic reactors were launched by theformer Soviet Union in the period 1987–1988. Their operational lifetimes were 142 and 343 days. It should be notedthat no nuclear reactors have been launched since 1988. In
addition to nuclear reactors, RTGs have been used as spacecraft power sources. The United States launched 25 missions
using 43 RTGs as power sources, two of them using 210Po
and the others 238Pu. The former Soviet Union launched two
missions with RTGs using 210Po and one mission with four
RTG units using 238Pu [U44].
393. Radioisotope heating units (RHUs) utilize radioactive
decay to provide heat to surrounding satellite systems andinstruments. RHUs have been used on board deep-spaceprobes (i.e. space probes operating beyond the asteroid belt),such as the United States New Horizons probe to Pluto,
launched in 2006, and on board planetary landing craft suchas the Lunokhod lunar rovers of the former Soviet Union andthe United States Mars Exploration Rovers. RHUs are usually small in size and typically produce approximately 1 Wof thermal power. Depending on the size of the spacecraft,the number of RHUs used can vary.
394. The current status of these devices is shown in figure XXXIII. Radioactive sources have also been used on
board satellites and launch vehicles in applications such as
triggering launch vehicle flight termination systems, calibration of on-board instruments and scientific experiments. As
an example, the Mars Exploration Rovers (launched in 2003
and still operational as of April 2008) each carry a Mössbauer spectrometer, which uses a small amount of 57Co, andalso an alpha particle X-ray spectroscope. Sources of thistype are small, and their impact on the environment is­considered minimal [U44].
395. In eight cases all or part of the nuclear system reentered the earth’s atmosphere, and there have been two situations where environmental contamination occurred. The
first was in April 1964, when the United States SNAP 9A
satellite burned up during re-entry. In August 1964, plutonium was detected in the stratosphere (at a height of 32 km),and in May 1965, it was detected at aircraft altitude. In
November 1970, it was estimated that some 5% of the original plutonium was still in the earth’s atmosphere. Plutonium
was eventually detected on all continents and at all altitudes—the concentration in the southern hemisphere wasabout four times higher than in the northern hemisphere. The
second event was in 1978, when the Cosmos-954 satellite of
the former Soviet Union came down over Canada, leading toa track of radioactive residues some 500 km long. Some

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

50 other objects have been recovered. Other satellites orparts of satellites have fallen into the oceans, and one wasrecovered intact [E1].
396. Reports of accidents involving unconventional orphan
sources in the new States that resulted from the dissolutionof the Soviet Union have caused particular security concerns.
The new States, some of which were not even aware of theexistence of such sources, exercised no control over them.
Many orphan sources have also been found on former military bases. The resulting exposures are described in annex C.
Notable cases of particular concern are abandoned thermoelectric generators containing powerful radioactive sourcesof 90Sr, which were introduced in the 1970s for dual civilian
and military use. RTGs were used in various civilian andmilitary applications, for example to power navigational
beacons and communications equipment in remote areas.
They usually hold over 1.5 PBq of 90Sr. RTGs were widelyused in the former Soviet Union for such applications as
generating electricity, heat and battery power for remotecommunication systems. These types of generator have also
been built in the United States, and their radioactive content
is more or less of the same order of magnitude. A large
number of navigational beacons powered by these RTGs
were operated in the Arctic area from Novaya Zemlya to the
Barents Straits. In Alaska, United States, several generators
were located in the Burmont area. Many RTGs are now being
recovered and their sources are being recycled. The firstabandoned RTG was found and recovered from the riverbedof the Ingury River in the Republic of Georgia. Two other
RTGs were recovered by the IAEA in a remote forested areaof north-west Georgia in 2001. RTGs were also found in
Tajikistan, dumped in an abandoned building and completely
unsecured. A number of RTGs have also been recovered in
Belarus [G13, I35].
397. In the period of 2004–2005, a bilateral project between
Norway and the Russian Federation decommissioned 96
RTGs from north-western Russia. It is estimated that about
760 RTGs primarily used as lighthouse energy sources still
remain along the northern Russian coast. An analysis of radiation protection issues related to decommissioning RTGs
has shown that a safe decommissioning practice is unlikely
to result in significant radiation exposure of human populations, with a worst case scenario being direct contact with an
exposed 90Sr heat source [S34].
(c)Summary on public exposure due to military uses ofatomic energy

398. Activities, practices and events involving militaryand defence uses of sources of radiation have led toreleases of radioactive material into the environment withresulting exposures of human populations. The main contribution to the global collective dose resulting from manmade sources has come from the testing of nuclear
weapons in the atmosphere. This practice occurred
between 1945 and 1980. Each nuclear test resulted in an

275

unconstrained release to the environment of substantial
quantities of radioactive material. These were widely­dispersed in the atmosphere and eventually deposited­everywhere on the earth’s surface.
399. Historically, the Committee has given special attention to the evaluation of exposures due to atmosphericnuclear weapons testing. Numerous measurements of the
global deposition of 90Sr and 137Cs and the presence of theseand other fallout radionuclides in the human diet and thehuman body have been made since the time tests took place.
The worldwide collective dose resulting from this practice
was evaluated in the UNSCEAR 1982 Report [U9], and asystematic listing of transfer coefficients for a number offallout radionuclides was given in the UNSCEAR 1993
Report [U6].
400. Although the total explosive yields have been divulged
for each test, information concerning the fission and fusion
yields remains suppressed for the most part. Some generalassumptions have been made to estimate the fission andfusion yields of each test in order to estimate the amounts ofradionuclides produced in the explosions. The estimatedtotal fission yields from all individual tests is in agreementwith the estimate of global deposition of the main fission
radionuclides 90Sr and 137Cs, as determined by worldwide
monitoring networks [U3].
401. With improved estimates of the production of eachradionuclide in individual tests and using an empiricalatmospheric transport model, it has been possible to determine the time course of dispersion and deposition of radionuclides and to estimate the annual doses due to various
pathways in each hemisphere. In this way it has been estimated that the world average annual effective dose reacheda peak of 110 μSv in 1963 and has since decreased to about
5 μSv (and now results mainly from residual levels of 14C,
90
Sr and 137Cs in the environment). The average annual
doses are higher than the global average by 10% in thenorthern hemisphere (where most of the testing took place)and are much lower in the southern hemisphere. Although
there was considerable concern at the time of testing, exposures in fact remained relatively low, reaching at mostabout 5% of the background level due to natural radiation
sources.
402. Exposures of local populations living in areas aroundthe test sites have also been assessed using available information. The level of detail is still not sufficient to document
the exposures with great accuracy. Attention to local conditions and consideration of the potential for exposure were
not great in the early years of the test programmes. However,dose reconstruction efforts are proceeding to clarify thisissue and to document the local and regional exposures that
occurred. Local and regional doses may have been very different from the exposure of global fallout. An example isshown in figure XXXIV, where results for 137Cs deposition
are presented for the tests in Nevada and for the contribution
from global fallout [S23].

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403. Underground testing caused exposures beyond thetest sites only if radioactive gases leaked or were vented.
Most underground tests had a much lower yield than atmospheric tests, and it was usually possible to contain the debris.
Underground tests were conducted at the rate of 50 or more
per year between 1962 and 1990. Although it is the intentionof most countries to agree to ban all further tests, both atmospheric and underground, the treaty to this effect has not yet
come into force. Further underground testing occurred in1998 in India and Pakistan, and in 2006 in the DPRK. Thus
it cannot yet be stated that the practice has ceased. Underground testing resulted in a large global burden of radio­
active material, and in particular of plutonium, albeit inunderground environments. The contribution of this material
to future population exposure is uncertain. Currently theseresidues are not expected to expose members of the public,
because they are buried deep underground, and, because of
the high temperature reached during the tests, they werefused within the matrix of host rock in an apparently stable
and insoluble form.
404. At present there is great concern regarding the reuseof nuclear test areas, since some are being reoccupied.
Residues in some environments, for example in localized
areas at the Semipalatinsk test site, may be considerable,
while in others, such as the Mururoa and Fangataufa Atolls,
the residues will not contribute more than a fraction of thenormal background exposure to a population eventuallyoccupying the site. For other sites still, such as the Marshall Islands and Maralinga, exposures will be highly
dependent on the habits of the populations occupying thearea.
405. During the time when nuclear weapons arsenals werebeing built up, and especially in the earlier years (1945–1960), there were releases of radionuclides and exposures of
local populations downwind or downstream of the military
nuclear installations. Since monitoring of releases was limi­
ted and there was little recognition of the potential risks,present evaluations of exposure must be based on dosereconstructions. Results are still being obtained that document this experience. Practices have greatly improved andarsenals are now being reduced.
406. The military use of DU has led to the contaminationof large areas with residues from munitions in several locations, for example Kosovo, former Serbia-Montenegro,­Bosnia and Herzegovina, Kuwait and Iraq. This fact has created serious concern that members of the public could be
exposed to such residues. A large international effort toassess the consequences of this contamination has been performed, and the main conclusion is that, except for a few
specific scenarios (such as the long-term handling of lumps
of DU), exposures are expected to be low. It is very unlikelythat the long-term behaviour of DU with regard to the leaching and transport of corroded DU lodged in the ground andits potential migration could cause any impact on underground water sources. An assessment of the DU residues in
Iraq has not yet been performed.

E. Historical situations
407. Some experiments using atomic weapons were carried out that were not related to military activities. However,these operations would not be allowed today under current
international conventions.
408. Nuclear explosions for peaceful purposes. Over a
period of 24 years, 128 nuclear explosions for peaceful purposes were conducted at 115 sites in the former Soviet
Union—in Russia, Kazakhstan, Uzbekistan, Turkmenistan
and Ukraine. The first was in 1965 at the Semipalatinsk test
site, in the Chagan River channel, to create a water reservoir,
and the last was in 1988, near the town of Kotlas. The overall
quantity of fission fragments was about 100 kg. Of 108 camouflet explosions,3 76 were fully contained. In 26 cases there
was radioactive gas leakage (blasts showed pressure efflux),and one explosion, Kraton-3, resulted in the release of radioactive products. The explosion sites and their technical purposes are shown in figure XXXV. The total energy yield ofpeaceful nuclear explosions in Russia reached 0.75 Mt, or
2% of the value for all underground nuclear explosions in theformer Soviet Union.
409. Radioactive traces and contamination of soil andvege­tation cover are very rare. Some of these 128 eventswere single excavation explosions. Five of these, such as theTaiga test, led to the contamination of adjacent areas, requiring remediation. The Taiga test was an attempt to create acanal; this resulted in a radioactive trace 25 km in length. An
accidental release from the Kraton-3 test caused the formation of a trace 31 km in length [V10]. The undergroundnuclear explosion Kristall took place in 1974. Its purpose
was to construct a reservoir dam for diamond enrichment
plant tailings. Explosions of this type are accompanied by
the formation of craters and are characterized by significant
releases of radioactive products into the environment.
Because of the heavy radioactive contamination, all further
work at the Kristall site was stopped. In 1990 a water-filledcrater, 60 m in diameter and 6 m deep, still existed at thelocation of the explosion. During clean-up operations in
1992, the crater was filled with barren rock from the Udachnaya diamond field and was covered with an artificial moundabout 100 m in diameter and 7–20 m in height [G5].
410. Kazakhstan’s low population density, vast territories
that are unsuitable for farming and considerable reserves ofminerals made the country a convenient location for thedevelopment and production of defence technology andarmaments. Apart from the Semipalatinsk test site, there arethree other test sites in Kazakhstan where undergroundnuclear explosions were conducted for peaceful purposes.
The radioecological situation at the three sites is not considered serious for the population or the environment. However,
the radioecological situation at the Koshkar-Ata storage
facility for waste is of major concern [C14].

3

Camouflet: a cavern caused by a subterranean explosion.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

F.

Exposure from accidents

411. Several accidents have included the release of nuclear
or radioactive material to the environment, leading to exposure of members of the public. In the present report, theChernobyl accident, which occurred in 1986, is described inannex D, “Health effects due to radiation from the Chernobyl accident”, and other accidents, such as the Kyshtym
accident of 1957, the Windscale accident of 1957, the Three
Mile Island accident of 1979 and the Tomsk accident of
1993, are described in annex C, “Radiation exposures in
accidents”. There have also been accidents with orphan
sources that involved exposures and fatalities among members of the public; these accidents are also described inannex C.
G. Summary on public exposure412. Exposure to natural sources of radiation is an unavoidable fact of the human condition. The single main
source of exposure is the inhalation of radon gas. The estimates of the global average per caput values of exposure tonatural sources of radiation are essentially the same as in
the UNSCEAR 2000 Report. The estimated value of worldwide average annual exposure to natural radiation sources
remains at 2.4 mSv. The normal range of exposures to thevarious components is presented in table 12. As describedearlier in this annex, the dose distribution worldwide is
expected to follow approximately a log-normal distribution,and most exposures would be expected to fall in the range
1–13 mSv/a.
413. The interest in exposures to NORM is increasing as
new situations are identified and corresponding dose assessments are performed for specific scenarios. Doses of up to afew millisieverts per year may be expected for some specificscenarios, such as the use of sludges from water treatment as
fertilizers, or the use of wastes and other materials as landfill
or building materials. There is not yet a consistent approach
to characterize inventories of sources and to estimate potential and actual exposures in order to extrapolate to a worldwide dose assessment. The Committee encourages thecontinued development of inventories and methodologies fordose assessment in order to make possible a broader view ofthe scenarios in a global context.
414. Residues due to conventional mining operations also
lead to huge amounts of material with enhanced levels of
NORM, and these represent a challenge regarding both thedisposal of the residues and site restoration. The large diversity of ores containing low levels of nuclides from the uranium and thorium families, which may be concentrated in
products, by-products and wastes, complicates the problem.
The detailed picture of worldwide exposure is far from complete. As with other contaminated sites, the main radioactive
materials are still under the control of operators, and mostsituations pose mainly potential exposure for members of
the public. Although the public exposure is not expected to277

be high, some areas with enhanced levels of NORM may
involve the low-level exposure of large numbers of people. A
large effort is needed to reach an international consensus on
ways of addressing this situation to keep the public exposures under control at levels compatible with exposures toother sources.
415. One continuing practice is the generation of electricalenergy by nuclear power reactors. During the routine operation of nuclear installations, releases of radionuclides are
low and radiation exposures must be estimated using environmental transfer models. For all fuel cycle operations(mining and milling, reactor operation and fuel reprocessing), the local and regional exposures are estimated to be
0.72 man Sv/(GW a). For the present world nuclear energy
generation of 278 GW a, the collective dose per year of practice is of the order of 200 man Sv. The assumed representative global value for the local and regional populations of
nuclear installations is about 250 million persons, and theannual per caput dose to this population is less than 1 μSv.
The collective doses due to globally dispersed radionuclides
are delivered over very long periods and are expressed forthe projected maximum future population of the world. If thepractice of nuclear power production were to be limited to
100 years at the present capacity, the maximum annual per
caput effective dose to the global population would be lessthan 0.2 μSv. This dose rate is minute compared with that
due to natural background radiation.
416. Releases of isotopes produced and used in industrial
and medical practices have been discussed and appear to be
associated with rather insignificant levels of exposure of thegeneral public. Except in the case of accidents, in whichmore localized areas can be contaminated to significant levels, there are no practices that result in important exposures
as a result of radionuclides released to the environment.
417. While doses due to nuclear power production havebeen extensively described and reported, this is not the casefor military uses and activities. Furthermore, some historicalestimates assigned doses to nuclear power production (suchas those due to the generation of radioactive waste and touranium mill tailings, among others) that were in part also
related to military activities.
418. The main contribution to the global collective dose
due to man-made sources has come from the testing of
nuclear weapons in the atmosphere. This practice occurred
between 1945 and 1980. These tests have led to local,
regional and global exposure because of the worldwide dispersion of radioactive material in the atmosphere, material
that was subsequently deposited everywhere on the earth’s
surface. It has been estimated that the worldwide averageannual per caput effective dose reached a peak of 110 μSv in
1963 and has since decreased to about 5 μSv (mainly due toresidual levels of 14C, 90Sr and 137Cs in the environment). The
average annual doses are higher than the global average by
10% in the northern hemisphere, where most of the testing
took place, and are much lower in the southern hemisphere.

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The underground testing also left an environmental legacy ofplutonium in the subterranean environment of all the sites
involved in such tests. Although currently any exposure tothese sources is low, exposure scenarios for the distant future
are very uncertain.
419. Besides areas related to atomic bomb production andtesting, early uses of radiation also left a legacy of numerous
small contaminated sites around the world. Efforts to decontaminate these sites and return them to public use have beena focus of attention in many countries. Several types of contamination are involved, many related to industrial uses ofnaturally occurring radionuclides or to old mining areas.
Exposures and collective doses are site-specific; once theareas are defined, exposures can be constrained. There is a
general tendency for exposures to fall with time because ofclean-up procedures, although for some sites there will be aneed for long-term follow-up because of the long half-livesof the radionuclides involved. In the United States alone, justover 5,000 remediation projects have been completed to dateat various USDOE facilities, and another 5,400 remain.
Some 1,186 sites are currently under decommissioning. As
site release criteria are usually developed with a focus on
critical group exposure, real doses to the public will depend
on whether released sites are actually occupied. In general,
individual doses estimated for a hypothetical critical groupare in the range 0.3–1.0 mSv. Regional average individual
doses will be at least one order of magnitude lower, and thecontribution to the worldwide population doses will mostprobably be negligible.
420. The enrichment process for natural uranium generates
a large amount of by-products containing DU. Owing to theproperties of this dense metal, it has found civilian and military uses. Military use led to pockets of contamination overlarge battlefield areas on the territory of the former Yugoslavia and in Kuwait. This has led to great public concern, andconsequently considerable work has been done to assess current and potential exposures due to these residues in theenvironment. Although most areas were cleaned up before
release to public access, uncertainties remain on the longterm exposures to specific individuals. This is because of thepossibility of penetrators presently buried underground being
found following human actions such as digging or ploughing, and of the enhanced corrosion rates observed for penetrators, which could ultimately lead to migration of DU into
underground water. However, no significant collective doses
are expected to result from either of these pathways.
421. Historically contaminated sites related to the peacefuluses of atomic energy are primarily related to the radiumindustry. These areas, mainly located in the United States,
the European Union and Canada, have already been identified, and most of them have been isolated from the public or
have been the subject of decommissioning programmes.
Residual exposures are thereby constrained to levels that arecompatible with current operational practices. There are alsoa large number of sites with mining residues associated with
nuclear power production worldwide. Large environmental

restoration programmes are being undertaken in order tobring the level of exposure in these areas within the range of
those considered acceptable for ongoing practices.
422. Possible future practices (such as weapons dismantling, decommissioning of installations and waste management projects) can be reviewed as experience is acquired,
but these are all expected to involve little or no release of­radionuclides and consequently little or no exposure.
423. A large number of smaller accidents have also resulted
in the exposure of members of the public, and many have led
to fatalities. Annex C of the UNSCEAR 2008 Report, “Radiation exposures in accidents”, discusses this subject in moredetail. Most of these accidents resulted in the exposure ofsmall groups of people to radiation from industrial and medi­
cal sources that had left institutional control. These accidents
have mostly involved relatively small numbers of persons,usually family, close friends or neighbours, but individual
doses were in some cases very high.
424. There were also a few situations where this type of
accident led to more widespread environmental contamination and to the exposure of larger numbers of people. These
include: the Goiânia accident in 1987, with the dispersion
within an urban area of a medical 137Cs source; the accident
in Mexico in 1983, where a cobalt source for medical purposes found its way into the production of steel used inbuilding material and other objects; and the accident in­Taiwan, China, where several residential buildings usedmaterial with contamination from a cobalt source. Such accidents led to widespread exposures, and although the collective doses resulting from such events are not high, thoseindividuals who personally manipulated the sources were
subject to doses that led in some cases to deterministic effects
or even death.
425. Exposure of members of the public to the various
sources discussed in this annex has a very wide variability inactual doses and in the contribution of different sources to theoverall exposure. As an example, figure XXXVI shows theestimated contribution of different sources to the population
exposure of different countries. In describing exposure fromdifferent sources, there is no standard pattern followed by different countries. For example, most countries do not havespecific data on exposures from consumer products, andtherefore such data are not included on their overall assessments. Also, exposures to sources have different time trends
in different countries. For example, while in United Kingdom
it has been verified that the contribution from various sources
has not changed significantly since the 1970s, with natural
sources dominating public exposure, in the United States, theaverage annual per caput dose from medical exposure has
increased from 0.54 mSv in 1982 to about 3 mSv in 2006,
making medical exposure the largest source of radiation
exposure to United States population [J5, M23].
426. A better understanding of the components of the total
exposure from different sources on a geographical basis

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

could change the current exposure assessment and lead tomore precise estimates of the distribution of exposures
worldwide. Up to now, only the variability associated with
exposures to individual sources has been taken into accountin worldwide dose estimates. There are, however, circumstances in which the distribution of doses due to one source
affects the overall distribution of doses due to other sources.
This can be the case, for example, for certain locations where
there are high levels of natural radionuclides in the environment and where higher doses due to radon inhalation may becorrelated with high doses due to external exposure or foodingestion. Another possible situation is the uneven distribution of nuclear power plants worldwide, which is broadly
correlated with the distribution of population.

279

427. An example of different dose distributions affecting
public exposure is given in figure XXXVII, which shows
several maps related to different sources of exposure of the
public in the United States. It can be seen that concentrations of uranium and thorium are closely correlated witheach other and also with external dose rates and radon concentrations. Also, the distributions of nuclear installations
and of population density appear to be correlated. The distribution of collective dose contributions may be very different from current estimated distributions, considering
specific distributions among the individual quantities
involved. This could indicate a need for future revision of
the methodology for estimating averages and ranges of­population doses worldwide.

III. Occupational radiation exposure428. The International Labour Organization (ILO) [I62]and the International Basic Safety Standards [I7] define
occupational exposure as “all exposure of workers incurredin the course of their work, with the exception of exposure
excluded from the Standards and exposures from practices
or sources exempted by the Standards” [I62].
429. Various national authorities or institutions have useddifferent methods to measure, record and report the occupational data included in this annex [I25]. The main features ofthe method used by each country that responded to the
UNSCEAR Global Survey of Occupational Radiation Exposures are summarized in table A-15. The procedures for therecording and inclusion of doses differ from practice to practice and from country to country. It must be recognized thatdifferences in monitoring and reporting practices do exist,
and these differences may, in particular cases, lead to spurious conclusions being drawn from comparisons between
reported data.
430. The criteria applied in different countries to selectworkers who should be monitored differ considerably. Somecountries monitor only the exposed workers, while others
also include non-exposed workers in their individual monitoring programmes for various reasons. This can lead to spurious results when attempting to compare levels of exposure
in different countries and practices. Moreover, the exposuredue to radon is often underreported, since many countries
record the dose only when radon concentrations of above
1,000 Bq/m3 in air are found. There are likely to exist workplaces where radon exposure can deliver significant doses
but which have not yet been identified [F15].
431. Occupational radiation exposures have been evaluated by the Committee [U3, U6, U7, U9, U10] for six
broad categories of practice: practices involving elevatedlevels of exposure to natural sources of radiation, thenuclear fuel cycle, medical uses of radiation, industrial
uses, military activities and miscellaneous uses (which

includes educational and veterinary uses of radiation). The
Committee has evaluated five-year average exposuresbeginning in 1975. The data presented in this annex are forthe periods 1995–1999 and 2000–2002. The data from theprevious periods are provided for comparison. Table 48
presents the practices for which the occupational exposure
has been evaluated.
432. The data in this annex were obtained in much the same
way as the data for the UNSCEAR 2000 Report [U3], i.e. bymeans of a questionnaire, the UNSCEAR Global Survey ofOccupational Radiation Exposures. For the current period, a
new questionnaire (requesting more detailed information forthe period 1995–2002) was distributed to Member States of
the United Nations by the UNSCEAR Secretariat. The data
have been supplemented by other (usually published) sourcesof information. For the nuclear power industry, for example,
a principal source is the joint databank of the Organisation
for Economic Co-operation and Development/Nuclear
Energy Agency (OECD/NEA) and the IAEA—the Information System on Occupational Exposure (ISOE) [O14, O19,
O20], which serves as a main source of data on occupational
exposure resulting from reactor operations for the period
1995–2002. Table A-15 presents the complementary information provided by those States that responded to the
UNSCEAR survey.
433. Differences may exist in the procedures used in various countries to categorize workers according to their occupations. This limits the validity of direct comparisons
between data compiled in different countries. Where theselimitations may be important, they are identified. The extent
to which valid comparisons between countries can be madeis also influenced by differences in the approaches used tomeasure and report occupational exposures, e.g. the type of
dosimeter used, its minimum detectable level (MDL), thedose entered into records when the measured dose is lessthan the MDL, and the dose assigned when dosimeters are
lost. The approaches used in measuring and reporting

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occupational exposures in each of the countries for whichdata were reported are summarized in table A-15. Where
important differences in approach are apparent, caution
should be exercised in making direct comparisons between
data.
434. In the UNSCEAR 2000 Report, the Committee evaluated occupational exposure for each practice in each country using average values for all workers over five-year
periods. The purpose of this annex is to provide moredetailed information on occupational exposure related to thedifferent practices, for example to identify job functions andcategories of work within each practice that lead to moresignificant exposures, to identify the contributions of external versus internal exposure to the total effective dose, andto obtain information about the reliability of measurements
associated with the accreditation or authorization of­monitoring services.
435. About 70% of the countries that reported data havetheir external dosimetry services accredited or authorized bysome national or international regulatory authority. The situation is the very different for internal dosimetry, for which
about 25% of the countries have reported that their services
are accredited or authorized.

A.

Assessment methodology
1.

Dose recording

436. In most countries, dose recording and reporting practices are governed by regulations and may differ for various
categories of workers depending on the anticipated levels of
exposure. The IAEA, in its publications [I7, I13, I14, I16,
I27], has provided guidelines on how monitoring data andresults should be reported, what dose levels should berecorded, and what documents and records of radiation
exposure should be maintained. Although there are guidelines for dose recording, there may be variations from country to country that may significantly affect the reported
values of collective dose. The most important differences
arise because of the following factors:− The recording of dose values less than the MDL;
− The technique used for measurement of externalradiation exposure, for example thermoluminescent dosimeter (TLD), film, electronic dosimeter,­optically stimulated dosimeter or glass dosimeter;− The assignment of dose values to fill missing ­periodsin the records;− The evaluation of anomalous results, such as­unexpectedly high or low dose values;− The subtraction of background radiation doses;− The protocol for determining who in the workforce
should be monitored and for whom doses should berecorded in particular categories;− Whether or not internal exposures are included or
are treated separately;− The reliability of the individual monitoring data.
437. In order to ensure the reliability of dose assessments,
some countries have implemented systems to authorize
monitoring services based on a set of requirements established by the national regulatory authority, while others
apply criteria based on the quality management system foraccrediting individual monitoring services [M19].
2.

Characteristics of dose distributions

438. The dose distributions presented in this annex follow
the same approach as the one described in the UNSCEAR
2000 Report [U3]. The Committee is interested in comparing dose distributions and in evaluating trends. For these
purposes, four characteristics of the dose distributions are
identified as being particularly useful:− The average annual effective dose (i.e. the sum of
the annual dose due to external irradiation and thecommitted dose due to intakes in that year), E;− The annual collective effective dose (i.e. the sumof the annual collective dose due to external irradiation and the committed collective dose due tointakes in that year), S;− The “collective dose distribution ratio”, SRE (forvalues of E of 15, 10, 5 and 1 mSv), provides anindication of the fraction of the collective dose
received by workers exposed at various levels of
individual dose;− The “distribution ratio for the number of exposed
workers”, NRE (for values of E of 15, 10, 5 and1 mSv), provides an indication of the fraction of thetotal number of workers exposed at various levels
of individual dose.
439. The annual collective effective dose, S, is given by:N

S = ∑ Ei
i =1

where Ei is the annual effective dose received by the ith
worker and N is the total number of workers. In practice, S isoften calculated from collated dosimetry results using thealternative definition:r

S = ∑ N jEj
j =1

where r is the number of effective dose ranges into which thedosimetry results have been collated and Nj is the number ofindividuals in the effective dose ranges for which Ej is themean annual effective dose. The average annual effective
dose, E, is equal to S/N. The number distribution ratio, NR, isgiven by:NRE =N (> E )N

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

where N(>E) is the number of workers receiving annual
doses exceeding E mSv. Similarly, the annual collective dose
distribution ratio, SR, is given by:SRE =S (> E )S

where S(>E) is the annual collective effective dose deliveredat annual individual doses that exceed E mSv.
440. Depending on the nature of the data reported and subject to the objectives of the evaluation (or the topic of interest), the “number of workers” may be those monitored, those
who work in workplaces classified as controlled areas, thosemeasurably exposed, the total workforce or some subset
thereof. Therefore these derived quantities will always be
specific to the nature and composition of the workforce
included in the estimation; when making comparisons, caution should be exercised to ensure that like is being ­comparedwith like.
3.

Estimation of worldwide exposures

441. Inevitably, the data provided in response to the
UNSCEAR Global Survey of Occupational Radiation
Exposures were insufficient for estimating worldwide levels of dose. Procedures were therefore developed by theCommittee to derive estimates of worldwide doses fromthe data available for particular occupational categories.
Two procedures were developed, one for application to
occupational exposures arising at most stages in the commercial nuclear fuel cycle and the other for general application to other occupational categories. For the occupationalgroups involved in practices other than the nuclear fuelcycle, the approach to derive estimates of worldwide doses
used in the UNSCEAR 2000 Report is no longer used here.
This is because the available data for the last two periods,
1995–1999 and 2000–2002, are not sufficient to derive areliable number that reflects the worldwide level of exposure. For medical exposure, the number of workers wasestimated on the basis of the information from the
UNSCEAR Survey of Medical Radiation Usage and Exposures. The Committee has decided to evaluate the worldwide level of occupational exposure for the different
practices in the industrial and miscellaneous fields on thebasis of the trends in the countries for each practice. The
worldwide level of exposure was estimated on the basis of
the quantile regression using the median estimated valuesof the data reported by the countries [K15].
442. In general, the reporting of exposures arising in thecommercial nuclear fuel cycle is more complete than that of
exposures arising from other uses of radiation. Hence thedegree of extrapolation from reported to worldwide doses isless, and this extrapolation can be carried out more reliably
than for other occupational categories. Moreover, worldwide
statistics are generally available on the capacity and production in various stages of the commercial nuclear fuel cycle.
Such data provide a convenient and reliable basis for

281

extrapolating to worldwide levels of exposure. Thus theworldwide annual collective effective dose, Sw, due to a given
stage of the nuclear fuel cycle (e.g. uranium mining, fuel fabrication or reactor operation) is estimated from the total of theannual collective effective doses reported by countries multiplied by the reciprocal of the fraction, f, of the world production (uranium mined, fuel fabricated, energy ­generated, etc.)
accounted for by these countries, namely:

1 n
∑ Sc
f c=1

SW =

where Sc is the annual collective dose arising in country cand n is the number of countries for which occupational
exposure data have been reported. The fraction of the total
production can be expressed as:n

f = ∑ Pc / Pw
c =1

where Pc and Pw are the production in the country, c, and in
the world, w, respectively.
443. The number of monitored workers worldwide, Nw,
in a given year is estimated by a similar extrapolation.
Because the data are more limited, the worldwide distribution ratios, NRE(w) and SRE(w), are simply estimated asweighted averages of the reported data. The extrapolations
to worldwide collective effective doses and numbers ofmonitored workers and the estimation of worldwide average distribution ratios are performed for each year. Valuesof these quantities have then been averaged over five-year
periods, except for the last period (2000–2002), which
included only three years, and the average annual values
are reported in this annex. The Committee has also madeprojections for exposures for the period 2002–2006 basedon extrapolating the trends for each practice over the six
periods previously analysed.
B.

Natural sources of radiation

444. Enhanced levels of natural background radiation areencountered in many occupational settings, especially inunderground mines. Mining involves a large number ofworkers, and although the data are more limited than thosefor occupational exposures to man-made sources, the annual
collective effective dose has been estimated to be approximately twice as large [U6]. Until implementation of theInternational Basic Safety Standards [I7], most countries
had not been particularly concerned with assessing occupational exposure to natural sources of radiation. Over the last
few years, exposures to enhanced levels of natural radiation
have become a focus of attention in the field of radiation
protection. Title VII of the European Basic Safety Standards[E11] and related guidance [E14] cover those work activities
where the presence of natural radiation sources that lead to asignificant increase in the exposure of workers and membersof the public cannot be disregarded. Besides the European
Union countries, others have already implemented radiation
protection legislation for NORM.

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445. The great majority of the workers exposed to natural
sources of radiation are not individually monitored. They
include aircrew, workers involved in mineral extraction andprocessing, and workers exposed to radon in workplaces
other than mines. The doses of aircrew are estimated frommeasurements in the aircraft and also by numerical simulation with computer codes. The occupational exposure of aircrew is controlled through limiting their time in flight [U41].
The workers involved in mineral extraction and processing
represent by far the largest occupational group exposed tosources of ionizing radiation. Only a few countries have monitored these workers on a routine basis. Besides mines, there
are several other workplaces where workers may receive very
high doses due to radon exposure; this has been highlighted
in the results of survey programmes conducted in some ofthese workplaces. Since for many countries these data are notroutinely recorded, an extensive review of the literature has
been conducted in order to present a more comprehensive
picture of occupational exposure to natural sources.
1.
(a)Cosmic ray exposures of aircrew and space crew

Aircrew

446. Exposure to cosmic radiation is influenced by manyfactors, as was discussed in section II.A.1 of this annex. The
International Commission on Radiological Protection(ICRP), in its Publication 60 [I47], has identified airline
flight crews as an occupationally exposed group. By theearly 1990s, the European Commission had agreed that acomprehensive survey should be undertaken of the radiation
environment produced by cosmic rays at aviation altitudes,
and an extensive programme of experimental and theoreticalstudies was supported [E9, S31]. The European Union has
established standards for the protection of workers exposed
to natural radiation [E10]. These standards explicitly
include flight personnel, who could receive an annual dose
due to cosmic rays of over 1 mSv. Since 2002, the European Union countries have recorded the associated doses inan ­occupational exposure database on a regular basis.
447. In recent years, new experimental studies have been
conducted of the monitoring methodology for estimating thelow- and high-linear-energy-transfer (LET) components of
the radiation field on board aircraft [B5, S10, S11, S31]. The
tissue equivalent proportional counter (TEPC) is the only
direct-reading dosimeter that measures both absorbed dose
to tissue and radiation quality in terms of linear energy [L14,
T1]. Several studies have been carried out to compare thedose estimated on the basis of the results of on-board measurements with the ones estimated by calculations using thecomputer codes. Good agreement has been observed between
the measured values and the calculated ones [B15, B17, B44,
F6, L8, L15, O2, S32].
448. A number of computer codes have been developed toestimate aircrew doses according to specific parameters
related to the flight routes. A new version of the Civil

Aerospace Medical Institute (United States Federal Aviation
Administration) computer program CARI-6M calculates, on
the basis of an anthropomorphic phantom, the effective doseof galactic cosmic radiation received by an individual on an
aircraft flying a user-specified route [N3]. The European
Program Package for the Calculation of Aviation RouteDoses (EPCARD) is a tool to calculate the effective dose or
the ambient dose equivalent and to determine the contribution of the different field components [M8]. The Predictive
Code for Aircrew Radiation Exposure (PCAIRE) estimatesvalues for the total ambient dose equivalent or the effective
dose. The PCAIRE program is based on experimental resultsfrom measurements on board aircraft, and its predictions
should agree with the associated measurement results [L6].
SIEVERT, a computerized system for the assessment of
exposure to cosmic radiation in air transport, is also a very
useful tool [B42, B44].
449. The different programs have been used to calculateroute doses for 28 different flights that took place during theperiod from May 1992 until September 2001. Calculations
were performed for both effective dose and ambient dose
equivalent. There are relatively larger differences (up to30%) between the results of the different transport codes for
effective dose than between the results for ambient dose
equivalent. For the latter quantity, the agreement is within
10–15%. This can be explained by the different assumptions
about the galactic proton distribution and the use of a proton
radiation weighting factor of 5 in the calculation of effective
dose, whereas the corresponding mean quality factor is 1.5
in the calculation of ambient dose equivalent [L15].
450. Since August 2003, 45 airline companies in Germany
have routinely assessed the exposure of their personnel by
application of computer codes. For the first year of dose registration, from August 2003 to July 2004, the national dose
registry for occupational exposure includes data on a total of
31,000 crew members. The collective dose to the group of
60 man Sv contributes more than 50% to the total of the collective dose of all workers in Germany. About the same proportion of collective dose to aircrew is reported by the
Netherlands [V3]. As seen in table 49, Germany and theUnited Kingdom report the largest number of flight personnel among the European Union countries. The averageannual dose of the flight personnel varies from 1.3 to 2.5 mSv.
None of the reported annual dose values exceeded 6 mSv.
The frequency distribution of the individual dose values isbimodal; however, the dose distribution observed for theother categories of work is characterized by an exponential
decrease in the number of observations with increasing values of the dose. Table 50 presents the dose estimates fors­ pecific flight routes leaving Frankfurt [S38].
451. The number of flight personnel in the United States isapproximately 150,000 [U27]. Radiation doses due to individual commercial flight segments typically range from 0.3 to>60 μSv per flight, depending on latitude, altitude and duration. Annual doses range from 0.2 to 5 mSv, depending on
flight routes and number of hours flown per year [W4, W5].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

452. There are a large number of females in the workforce.
Female flight attendants flying both a large number of hours
during pregnancy (e.g. 100 hours per month) and only theroutes with the highest dose rates (e.g. 0.006 mSv per blockhour) would exceed 0.5 mSv to the embryo/foetus ­(excluding
natural background and medical exposures) [W4].
453. Data on the occupational exposure of crew members
are presented in the first part of table A-16 and also intable 49. Most of the limited number of data refer to the year
2002. The number of reported monitored workers is 90,540.
The reported collective effective dose is 165 man Sv. The
reported average effective dose is about 1.8 mSv. The
reported average effective dose data are in agreement withthe data presented in table 49. In this table, information isprovided for the United States, with a workforce of approximately 150,000 [U27]. No changes in terms of the total
number of crew in the worldwide workforce have occurredsince the UNSCEAR 2000 Report. Assuming that the countries reported in tables 49 and A-16 represent about 80% of
the worldwide workforce, the total would be 300,000 workers. The average effective dose for the European countries isabout 2 mSv. The average annual flying time is estimated as
600 hours for aircrew in European countries and about 50%
more for aircrew in the United States. On the basis of thesevalues, it is assumed that 50% of the workforce is exposed to
2 mSv/a and 50% is exposed to 3 mSv/a. Under theseassumptions, the estimated collective effective dose is about
900 man Sv. This value for the collective dose is about thesame as that estimated in the UNSCEAR 2000 Report,
800 man Sv [U3]. These doses could be slightly underestimated, if it is assumed that the crew members are also frequent flyers, since most of them receive free air tickets fortravel with their airlines. Couriers represent a separate group;
they may spend greater total times in flight in the course of ayear, but even so are unlikely to incur a dose exceeding
10 mSv in a year.
(b)Space crew

454. At altitudes of between 200 and 600 km and at low
inclinations, the major contribution to the absorbed dose isdelivered inside the South Atlantic Anomaly (SAA) by thegeomagnetically trapped protons and electrons of the radiation belt. The SAA is an area where the radiation belt comesclosest to the earth’s surface owing to a displacement of themagnetic dipole axes from the earth’s centre. In this region,
fluxes vary extremely rapidly with altitude, because of interactions of the charged particles with the nuclei of the atomsof the upper atmosphere. The flux in the SAA is anisotropic,
with most of the flux arriving perpendicular to the magnetic
field lines [R8].
455. The dose measurements for the on-board crew of various missions (first United States Spacelab mission (SL1),
Dedicated German Spacelab missions (D-1 and D-2), International Microgravity Laboratories (IML-1 and IML-2),
German Mir-92 flight to the Russian space station) show that283

the doses were in the range 1.9 mSv to around 27 mSv,
depending on the mission, as shown in table 51. The main
contribution to the dose came from the protons of the SAA;
its value increases with altitude and decreases with ­increasing
solar activity and mass shielding [R6, R7, R8].
456. The fraction of the dose on the Mir space station due
to the SAA on an orbit inclined at 51.6° and at an altitude ofabout 400 km was determined during the Euromir ’95 mission. The measurement was performed using an hourly
measuring period for 170 h. It was found that the maximum
dose due to crossing the SAA was equal to 0.055 mGy. Averaging all the measurements, it was calculated that the meandose rate inside Mir varied from 0.012 to 0.014 mGy/h, andthat half of this value was due to the SAA [D4].
457. Measurements of the cosmic radiation dose inside theMir space station and the additional dose to two astronauts inthe course of their extravehicular activity (EVA) were performed. During an EVA lasting 6 h, the ratio of dose rates
inside and outside Mir was measured. During the EVA, Mircrossed the SAA three times. Taking into account the influence of these three crossings, the mean outside/inside dose
rate ratio was 3.2. The absorbed dose rate inside Mir was
0.023 mGy/h, while the mean absorbed dose rate during the
EVA was 0.073 mGy/h [D12].
458. The dose assessments for various space missions of theformer Soviet Union and the United States show that the daily
absorbed dose varied between 0.32 and 0.57 mGy, and thedaily dose equivalent between 0.62 and 1 mSv. The dose
assessment was based on data from dosimeters placed in different locations in the space station. The value for the radiation weighting factor was about 3 at high latitude and decreased
to about 1.5 near the equator. This effect is due to the greater
geomagnetic protection at low latitudes, where only highenergy particles penetrate the atmosphere. Nearer the poles,
there is a higher particle flux with lower mean energy. Variations could be explained by differences in the mass shielding
properties at the locations of these detectors [B41].
459. The second flight of IML-2 on Space Shuttle flight
STS-65, which was launched on 8 July 1994, was sustainedin a 28.45º by 296 km orbit for a duration of 14 days, 17 hours
and 55 minutes. The crew doses varied from 0.94 to 1.2 mGy.
A reasonable agreement was found between the galactic cosmic ray dose, dose equivalent and LET spectra measured
using the TEPC flown in the payload bay and those ­calculatedusing models [B3].
460. The Mir-18 mission began in March 1995 [B4]. The
absorbed dose measurements for the Mir-18 crew showedthat the dose depended on the tasks the crew performed.
Estimates were 3.76 + 0.18 mGy, 2.87 + 0.15 mGy and3.53 + 0.24 mGy for the commander, flight engineer andflight researcher, respectively. Dosimeters were worn at least
80% of the time. The dose values are not corrected for theloss of high-LET particles. The Mir space station was in a
51.65º inclination orbit from 1986 to 2001.

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461. Evaluation of individual doses using cytogenetic
dosimetry techniques has shown that the yields of dicentrics
and centric rings scored after long-term space flights are considerably higher than those scored prior to the flights. In thisstudy, a total of 22 cosmonauts were examined. Some of themwere examined after repeated flights. The missions lasted for
4–6 months on average. Individual doses measured using biodosimetry to cosmonauts who showed a reliable increase in
the yields of chromosomal-type aberrations after their firstflights were estimated to be from 0.02 to 0.28 Gy [F3].
2.

Exposures in extractive and processing industries

462. The extraction and processing of radioactive ores arecarried out in a number of countries throughout the world.
The extractive industries include all forms of mining. Minerals and other natural materials that are not normally regarded
as being radioactive may nevertheless contain significant
levels of natural radionuclides from the uranium and thorium
decay chains. These raw materials, their by-products fromprocessing and the end products produced may lead to exposures in workplaces where there is often no perception, letalone appreciation, among workers of the various relevant
radiation protection problems. The main source of exposurein most mining operations is radon. Exposure due to longlived radionuclides in mineral dusts can, however, be­important in certain mining and other situations.

465. The main potential sources of occupational exposurein the extractive industries are the natural radionuclides arising from the radioactive decay of the 238U and 232Th series.
Exposures may arise via three main routes: (a) the inhalationof radon, thoron and their respective progenies; (b) the inhalation and ingestion of ore dust; (c) external irradiation withgamma rays.
466. Radon is the main source of radiation exposure in
most underground mining operations. While several isotopes of radon exist in nature, one (222Rn) dominates interms of dose to workers. Under some circumstances, 220Rn(thoron, a decay product of the 232Th chain) may also be
important. For convenience, unless stated otherwise, “radon”
is taken here to mean 222Rn. The short-lived decay products
or progeny of radon, rather than the gas itself, are the maincause of exposure, although for control purposes it is oftenthe ­concentration of the gas that is referenced [C13].

463. Mining is an extensive industry. Employment in themining industry is changing in several ways for a variety ofinterrelated reasons: commercial, political, technological,
demographic and social. The net effect, however, has been asteady fall in the number of people employed in mining.
According to the International Labour Organization, since
the early 1990s, when about 25 million people were estimated to be employed in mining (including some 10 million
in coal mining), the decline in employment has ranged fromsteady to more rapid at different times in different regions.
By the year 2000, a decline in the number of workers in mining ranging from 32% to about 45% was estimated to haveoccurred, which would give an estimate of about 6.8 million
workers involved in coal mining operations [I39]. The estimated number of underground coal mine workers in China is
about 6.05 million [L20]. Mining is still a male-dominatedindustry. Although more women are now working in all
aspects of mining in some countries, any increase in female
employment is generally from a very low base.

467. Continuous radon and thoron gas measurements alongwith several particle size distribution measurements weremade at 20 locations in a rare earth (monazite) pilot processing plant near Bangkok, Thailand. The measurements wereconducted from February 2001 to November 2006. A miniature alpha track detector combining both radon and thoron
measurements was used. The radon and thoron concentrations ranged from 15 to 100 Bq/m3 and 150 to 1,550 Bq/m3,
respectively. The measured thoron range was large at any single location. Near a monazite digesting tank, for example, thethoron concentration in air ranged from 80 to 1,500 Bq/m3
over the five-year period. The UNSCEAR 2000 Report’s conversion factors of 9 nSv/(Bq h m‑3) for effective dose fromradon and 40 nSv/(Bq h m‑3) for effective dose from thoron
were used. Several studies document the equilibrium fraction, Feq, for thoron indoors, as 0.02–0.03. The value 0.02
was used in the thoron dose calculation. The Feq used forradon indoors was 0.4. The calculated bronchial dose forindividuals who worked in the same location in the rare earth
processing facility for 2,000 hours in a year could lead to a
calculated annual lung dose of up to 0.7 mSv due to radonand 2.4 mSv due to thoron. The particle size distributions
taken over intervals of 1 to 2 months with a miniature integrating particle size sampler showed four peaks, at 5, 150,
400 and 5,000 nm, with 50% of the activity associated withthe 150 nm mode [H6]. The results of the SMOPIE project(Strategies and Methods for Optimization of Internal Exposures) indicate that rare earth processing may give rise toannual doses of over 20 mSv [V1].

464. By far the largest category of workers exposed to ionizing radiation are those employed in the extractive andprocessing industries. A rough estimate of the total numberof workers potentially exposed to internal radiation in nonnuclear industry in the European Union is 5,000–10,000.
Exposure situations for workers in these industries differ
considerably with respect to the type of industry, the conditions in the workplace, the radionuclides involved, and thechemical and physical forms of the matrices in which theradionuclides are incorporated [V1].

468. The natural radionuclides involved in any processing
technology for natural raw material end up either in the finished products or in the liquid, solid or gaseous waste generated. Depending on their chemical properties, theradionuclides are concentrated or distributed in the end products and in the waste [B19, S20]. The grinding of raw materials may generate fine particles of dust and also make iteasier for radon to escape into the workplace air. Processing
materials rich in uranium or thorium decay products at hightemperatures (e.g. coal combustion) could enrich airborne

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

dust in some radionuclides of the uranium and thorium
series, e.g. 210Po and 210Pb. At very high temperatures(3,000ºC or greater), other radionuclides of the uranium or
thorium series may also sublimate. For example, 228Ac may
sublimate during welding from welding rods doped with232
Th [B19, I18, I26].
469. In a survey programme involving six undergroundcoal mines in Baluchistan, Pakistan, radon measurements
were carried out to estimate the workers’ doses due to radon
exposure. Radon concentrations varied from 121 to 408 Bq/
m3 in the mines under study. The dose estimate was based on
the conversion factor of 5 mSv/WLM on the assumption thatthe occupancy time in the mines is 4,000–4,500 h/a. Consequently the annual doses for workers were within the range
2.1–7.0 mSv [Q10]. An evaluation of occupational exposurein three underground coal mines in Turkey (Kozlu, Karadonand Üzülmez) indicated average annual effective doses of
4.9 mSv. The total workforce was 12,510 and the collective
effective dose was estimated to be 61.5 man Sv [F10]. Evaluation of occupational exposure due to intakes of long-lived
radionuclides from the radon decay series by workers in coalmines in Brazil indicated average annual committed effective doses of less than 1 mSv [L17]. In an assessment ofoccupational radiation exposure carried out in Polish coalmines in 1997, it was estimated that the maximum value ofthe dose equivalent received by any miner during the periodof an entire year of work under such conditions would notexceed 3.5 mSv [I18, S24]. In a survey programme carriedout in three underground coal mines in Western Australia
employing 297 workers, the estimated average annual­effective dose was 2.9 ± 1.5 mSv [H22].
470. An occupational exposure assessment of some
80 coal mines in China was carried out during the period
2002–2004. The results indicated that the average annual
dose to the staff of the underground mines is 2.4 mSv, withthe largest dose being over 10 mSv [C12]. The effective
doses for Chinese coal mine workers seem to have adecreasing trend; the average value was reported as 4.8 mSv
for 1999 [T4]. Of the 6 million underground coal miners
countrywide, 1 million are working in large coal mines,
1 million in medium-sized coal mines, 4 million in small
coal mines and 50,000 in bone-coal mines. Bone-coal is an
impure coal that contains much clay or other fine-grained
detrital mineral matter; it is hard and compact. On thisbasis, the collective dose is estimated to be about
14,600 man Sv (table 52). Most of the occupational­exposure is due to radon and its progeny [C12].
471. In the Islamic Republic of Iran, there are about 150
underground mines, of which 60% are coal mines and 40%
metal mines. To assess the possible presence of high radonlevels in these mines, a radon survey programme of nonuranium mines was started in early 2000. The evaluation ofoccupational exposure in the ten mines gave the following
results: 35 workers incurred an average effective dose of
8.3 mSv, and 235 workers an average effective dose of
0.06 mSv in the two manganese mines; 235 workers in the285lead mine received an average effective dose of 1.2 mSv; and8,772 workers in the seven coal mines received an average
effective dose of 2 mSv [G9].
472. An assessment was undertaken of occupational exposure in 27 underground non-uranium mines in Western
Australia. These mines employed 2,173 workers, which
represented nearly 80% of the underground workforce atthe time of the survey. The average annual effective dose
across all mines was estimated to be 1.4 mSv, ranging from
0.4 mSv for a nickel mine to 4.2 mSv for a coal mine.
Radon progeny exposure contributed approximately 70%of the total ­effective dose [H22].
473. The average annual effective dose to workers in four
metal ore mines in Poland was 2.5 mSv (maximum value
9.6 mSv). Annual doses for workers in two lead and zincmines were estimated at about 4 mSv (maximum value
8.7 mSv), and for workers in two copper mines at about
2.8 mSv (maximum value 7.0 mSv); these doses were also
due to radon exposure [I18]. Workers at a commercial underground lead and zinc mine in Ireland have been monitored
for radon exposure; 11 workers received annual doses due toradon inhalation in the range 1–6 mSv [C25].
474. The estimated average annual doses received byunderground gold mine workers in South Africa were
6.3 mSv in 1997, 4.9 mSv in 1998, 5.4 mSv in 1999 and7.0 mSv in 2000. The data are presented in table 53. A survey programme carried out in the gold mines during 1993–
1994 found that 71% of the dose was due to inhalation ofradon gas and its short-lived progeny, 25% due to external
gamma exposure and the remaining 4% due to inhalation ofdust [I18, W17].
475. An evaluation of the occupational radiation exposure
to NORM in surface and underground mining operations ina gold mine in the Ashanti Region of Ghana showed that theannual effective dose is about 0.26 ± 0.11 mSv for surface
mining and 1.83 ± 0.56 mSv for the underground mines. The
total number of workers was 4,439 [D1].
476. A dose assessment for 45 workers in five different
areas of the largest underground phosphate mine in Egypt,
the Abu-Tartor phosphate mine, was conducted taking into
account measurements of radon, its short-lived decay products, thoron and external dose (using TLDs). The calculated
effective dose due to airborne radionuclides was the main
contributor to the occupational exposure and exceeded
20 mSv/a, especially at locations in the side tunnels (wherelevels were higher by a factor of up to 4 because of inadequate ventilation). The average annual effective dose was
11.66 mSv. The mean value of external dose as measured by
TLDs was 8.97 mSv/a. These results are presented intable 54. The dose estimate calculated from workplace measurements underestimates the annual dose by around 25%[K11]. The average annual effective dose (due to radon,
radon progeny and thoron progeny) in other Egyptian phosphate mines was 70.2 mSv, with a range of 12.2–136.9 mSv

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[H31]. More recent evaluation has been conducted in threephosphate mines in Egypt, located in the Eastern Desert
about 500 km south of Cairo. The average annual effective
doses for workers from the mines, due to inhaled radon­progeny, are in the range 107–182 mSv [E3].
477. The estimated annual doses for workers in surface
copper mines in Poland were about 1 mSv resulting frominternal exposure due to radium and about 0.5 mSv fromexternal exposure [I18]. The exposure of surface workers ingold mines in South Africa is generally very low, except forworkers in acid plants, where the radium originating fromthe pyritic ore can become very highly concentrated during
the formation of scales and give rise to substantial external
gamma and dust inhalation exposures. A survey of occupational exposure carried out in 1999 in South African mines
and mineral processing facilities (other than those associatedwith gold production) showed that around 98% of the9,955 workers received doses of less than 5 mSv. The workers with the highest exposures are in copper mining [W17].
478. Another group of exposed workers are those in diamond mines in Africa. Security measures are implemented
to reduce diamond thefts. These measures are explicitly
authorized through national regulations and cover a large
spectrum, from access control to the use of special equipment to prevent the employees having direct contact with the
diamonds. Personal searching, including searching by hand
and X-ray searching, which is practised in some conditions
in some countries, is one of the security measures. Personalsearching has two main functions: to recover diamonds thathave been concealed with the intention to steal, and to deterand prevent theft. The radiation dose is about 5 μSv per scanin screening workers to detect if they have swallowed or hidden diamonds in their bodies. There is no estimate of thenumber of workers involved in these diamond mines and of
how often they are exposed [I6].
479. A fluorspar mine that operated in St. Lawrence, Newfoundland, Canada, from the early 1930s until 1978 was estimated to have radon progeny concentrations of 2–130 WL.
The source of radon was eventually identified as the waterthat poured into the mines [D3]; the radon itself apparently
originated from the host granite. Mechanical ventilation was
introduced in all levels of the mine that were still operating,and the radon daughter levels subsequently fell below thesuggested limit of 1 WL in 1960 [M31]. The last fluorspar
mine was closed in St. Lawrence in 1978. The average annualinternal and external effective doses received by workers in
the phosphate fertilizer plant were 0.75 mSv and 0.88 mSv,
respectively [B39].
480. In most of the extractive and processing industries in
Brazil, average annual effective doses were somewhat
greater, above 1 mSv [L17]. An evaluation of internal exposure of workers at a thorium purification plant in Brazil
showed that the annual effective dose ranged from 0.12 to
1 mSv. In this facility, thorium sulphate is converted in the
purification process into concentrated thorium nitrate,

Th(NO3)4, which is then used in gas mantle production [C2].
The average annual effective dose of workers in an electrothermal plant in the Netherlands for producing elemental
phosphorus is about 1 mSv [E4]. A radiological survey was
conducted and a radiation protection system implemented
during the site remediation and decommissioning of an old
and abandoned Greek phosphate fertilizer industry. The initial estimate of the effective dose to workers involved in thedecontamination process, for a worst-case scenario, wasestimated to be up to 9 mSv [K17].
481. Various assessments of annual effective doses received
by workers in zircon milling plants have been reported. The
results of these assessments are summarized in table 55,
from which it would appear that, in most zircon milling
operations, workers do not receive annual doses exceedingabout 1 mSv. Except for bagging operations, this is likely tobe the case even if respiratory protection is not used [I41].
However, the results of the SMOPIE project indicate that zircon milling may give rise to annual doses of between 6 and20 mSv, in workplaces where protection measures are poor
or non-existent [V1].
482. Data from the UNSCEAR Global Survey of Occupational Radiation Exposures on the occupational exposure ofworkers involved in extractive and processing industries areincluded in table A-16. For coal mines, only the United
Kingdom has reported data on occupational exposure. The
workforce consists of 5,000 workers, who represent about
10% of the number reported in the previous period. The
average annual effective dose has remained constant at
0.6 mSv. For other mineral mines, five countries havereported data, representing about 1,300 workers. The ­average
effective dose is 1.2 mSv.
483. The level of exposure depends on a number of factors,
including the type of mine, the geology and the working
conditions, particularly the ventilation. In general, the occupational exposure is distinguished by the type of mine
(underground versus above ground). The range of typicalvalues of annual effective dose for underground coal mines
is 0.5–4 mSv. The typical average effective dose for coalmining operations is considered to be 2.4 mSv. The range oftypical values of annual effective dose for other mineral mining is 1.3–5.0 mSv. The typical average effective dose forother mining operation is considered to be 3.0 mSv. In orderto have a rough estimate of the worldwide level of exposuredue to the extractive mining industry, it is assumed that the
total workforce comprises about 11.5 million workers, that
60% of this workforce (i.e. 6.9 million workers) receive an
average annual effective dose of 2.4 mSv, and that 40% of
the workforce (i.e. 4.6 million workers) receive an averageannual effective dose of 3.0 mSv. This results in an estimate
for the annual collective effective dose of about 16,560 man Sv
for coal mines and 13,800 man Sv for other mines. This
makes a total of some 30,360 man Sv annually for the mining industry as a whole. It has been very difficult to distinguish the level of exposure and the numbers of workers
engaged in mining and mineral extraction. In this annex, theANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

collective dose estimated for workers involved in mineral
extraction includes those involved in mineral processing.
The level of exposure to radon may be underestimated, since
the doses for workers in workplaces where the radon concentration is below 1,000 Bq/m3 may not be reported. The
number of workers, the average effective doses and the­collective effective doses are presented in table 57.
484. The UNSCEAR 1988 Report [U7] estimated the collective doses for coal mining as 2,000 man Sv. This was
based solely on exposures in mines in the United Kingdom
and on the worldwide production of coal. The UNSCEAR
2000 Report [U3] estimated the collective dose as about
2,600 man Sv, which was about 16% of the current estimateof 16,560 man Sv. For non-coal mines, the collective dose
estimate has also increased considerably. The UNSCEAR
2000 Report [U3] estimated the collective dose as about
2,000 man Sv, which is about 14% of the current estimate of
13,800 man Sv. The overall estimate for mining activities is30,360 man Sv, which is about seven times higher than theprevious estimate [U3].
3.

Gas and oil extraction

485. Naturally occurring radioactive material (NORM)
found in the earth’s crust, largely in the form of 226Ra and228
Ra and their associated radionuclides, is brought to thesurface during gas and oil production processes. The NORM
represents a potential internal radiation exposure hazard toboth workers and members of the public through the inhalation and ingestion of radionuclides. In addition, a gamma
exposure rate higher than normal background has been
observed in the oil and gas industry.
486. The mixed stream of oil, gas and water associatedwith the production process also carries the noble gas 222Rn,
generated in the reservoir rock through the decay of 226Ra.
This radioactive gas emanating from the production zone
travels with the gas/water stream and then preferentiallyfollows the dry export gases. As a consequence, equipmentfrom gas treatment and transport facilities may accumulate
210
Pb formed from the short-lived progeny of 222Rn, whichplate out on to the inner surfaces of gas lines. These 210Pb
deposits are also encountered in liquefied natural gas­processing plants [G7].
487. NORM in the oil and gas industry has the potential to
give rise to external exposure during production owing to theaccumulation of gamma-emitting radionuclides. Moreover,it can give rise to internal exposure to workers and other persons through the inhalation or ingestion of radionuclides,
particularly during maintenance, the transport of waste andcontaminated equipment, the decontamination of equipment
and the processing of waste. The short-lived progeny of theradium isotopes, in particular of 226Ra, emit gamma radiation
capable of penetrating the walls of internally contaminatedpipes and vessels. Therefore the deposition of contaminatedscales and sludge inside these components produces287

enhanced dose rates outside them as well. The values dependon the amount and activity concentrations of radionuclides
present inside the components and the degree of shielding
provided by the pipe or vessel walls. Maximum dose rates
usually range up to a few microsieverts per hour, but in a few
cases dose rates of up to 100 µSv/h (about 1,000 times
greater than the normal background values due to cosmic
and terrestrial radiation) have been reported ­outside­production equipment [M14, T2, W9].
488. At the Omar oilfield in Syria, the highest equivalentdose rates were 30 µSv/h on the surface of the well-head and25 µSv/h on the surface of some piping containing scale
deposits, especially in valve and bend areas. In the Gulf ofSuez oilfield in Egypt, the maximum equivalent dose ratemeasured at the surfaces of separator tanks and piping, anddue to scale precipitate, was 33 µSv/h [A9]. Dose rates
observed in oil production and processing facilities varyfrom 0.1 µSv/h to 300 µSv/h [I23].
489. The IAEA has published information concerning concentrations of NORM in oil, gas and by-products that may
result in occupational radiation exposure. The concentrations of 226Ra, 228Ra and 224Ra in scales and sludge range fromless than 0.1 Bq/g to 15,000 Bq/g. The activity concentrations of radium isotopes are lower in sludge than in scales.
The opposite applies to 210Pb, which usually has a relatively
low concentration in hard scales but may reach a concentration of over 1,000 Bq/g in lead deposits and sludge. Althoughthorium isotopes are not mobilized from the reservoir, thedecay product 228Th grows in from the decay of 228Ra after
deposition of the latter. As a result, when scales containing
228
Ra age, the concentration of 228Th increases to about
1.5 times the concentration of 228Ra still present [I23].
490. An assessment of the occupational exposure t due topetroleum pipe scales has been performed for three oilfields.
Four radiation exposure pathways were investigated: inhalation of pipe scale dust generated during pipe rattling; incidental ingestion of the pipe scale dust; external exposure
resulting from uncleaned pipes; and external exposure resulting from pipe scale dispersed on the ground. The estimatedannual effective dose for the operator and the assistant was
0.11–0.45 mSv for inhalation and 0.02–0.1 mSv for sporadic
ingestion. The annual effective dose due to external exposure
from uncleaned pipes ranged from 0 to 0.28 mSv. The annual
effective dose due to external exposure from pipe scale dispersed on the ground was estimated to be 2.8 mSv for theoperator and 4.1 mSv for the assistant [H5].
491. According to an estimate based on assuming the inhalation of 5 µm AMAD (activity median aerodynamic diameter) particles incorporating 226Ra (with its complete decay
chain in equilibrium), 228Ra and 224Ra (also with its complete
decay chain in equilibrium), each at a concentration of
10 Bq/g, a committed effective dose per unit intake of about
0.1–1 mSv/g would be delivered. The exact value depends onthe extent of ingrowth of 228Th from the decay of 228Ra andon the lung absorption types assumed. For 1 µm AMAD

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particles, the committed effective dose per unit intake would
be 25–30% higher [I23].
492. Available data from the UNSCEAR Global Survey ofOccupational Radiation Exposures for gas and oil extraction are included in table A-16. The data have been presented for only two countries. The total number of monitored
workers was 500 for the period 1995–1999 and 600 for
2000–2002, and the average effective dose was 1.3 mSv for
both periods. It is difficult to estimate the collective dose forthis practice since the total number of workers exposed toionizing ­radiation is not known.
4.

Radon exposure in workplaces other than mines

493. The levels of radon in workplaces are exceptionallyvariable, and high doses to workers can arise in places otherthan uranium mines. Regulatory authorities have recognized
the importance of controlling radon exposure in workplaces
other than mines. The European Guideline 96/29/Euratom[E10], which formulated basic safety standards for the protection of the health of workers and members of the generalpublic against the hazards of ionizing radiation, included
consideration of areas where the presence of natural radiation sources would increase exposures to employees ormembers of the public to levels that could not be ignoredfrom the standpoint of radiation protection. ICRP Publication 65 [I48, I61] indicated a planning value in the range
500–1,500 Bq/m3, above which radiation protection measures are required; orientation values are available fora­ pplication to health protection.
494. The radiation protection regulations applied in­Switzerland since the promulgation of 1994/SSS-94 [S33]established a radon concentration limit of 3,000 Bq/m3 forindustrial areas. Orientation values of 200 Bq/m3 and400 Bq/m3 were indicated for new buildings and for therenovation of buildings, respectively. These workplaces arevaried in nature. They include industries (food industries,
breweries, laundries, etc.), waterworks, shops, public buildings and offices, schools, subways, spas, caves and closed
mines open to visitors, underground restaurants and shopping centres, tunnels (construction and maintenance) andsewage facilities [I21, S39, S41].
495. An occupational exposure survey in over 500 of the
2,600 water supply facilities in Bavaria showed that, in all
geological regions, exposure levels giving rise to over
6 mSv/a can occur. About 2% of the staff is subjected to
exposure levels that give rise to over 20 mSv/a [S40, T9,
T10]. A survey of occupational exposure was conducted inten drinking water supply plants in Slovenia. The annual
doses were found to be below 0.5 mSv at six of the workplaces and in the range 0.6–3.0 mSv at the other four [V16].
496. Occupational exposure in radon therapy rooms isrelated to the different treatment procedures, which affect thetemporal variation of radon and its progeny. An evaluation ofoccupational exposure due to radon and its progeny in thetreatment facilities of the radon spa Bad Gastein in Austria
produced different dose ranges for each of the four treatment
rooms monitored. The estimated annual effective doses were
9.4–32 mSv, 1.8–2.4 mSv, 1.3–1.7 mSv and 0.2–0.3 mSv[L7]. The annual individual effective doses to the employees
of a therapeutic dry carbon dioxide spa in Hungary, due toinhalation of 222Rn, ranged from 0.9 to 4.2 mSv. The highest
dose to a staff member was received by an attendant who
spent much of his time in the treatment room watching overthe patients in the “pit” [C30]. The results of the dose assessment for a therapeutic cave in Hungary showed that staff
received doses of up to 20 mSv/a when working 4 hours per
day in the cave [K6]. Annual effective doses of between 1 and
44 mSv were estimated for workers in Spanish spas [S29].
Bath attendants were the working group subject to the highest
doses. In Slovenia, a dose assessment was performed in fivespas; the radon ­concentration in indoor air rarely exceeded
200 Bq/m3 [V13].
497. In Slovenia, there are more than 3,000 caves locatedin the Karst regions. Some 50 professional guides and otherworkers are employed in the Postojna and Skocijanske caves,and many volunteers from local cave associations work orserve as guides for visitors in about 20 other caves. Annual
doses, estimated on the basis of various lung models, ranged
from 10 to 85 mSv [J7]. A survey carried out in 2002 ofoccupational exposure in three Irish caves showed that 13
workers received annual doses due to radon inhalation in therange 1–6 mSv, and one worker received an estimated annual
dose of 12 mSv [C25]. A dose assessment was carried out in
2004–2005 in the Lantian Xishui karst cave of Shaanxi,
China. The average annual effective dose to tour guides wasfound to vary between 1.2 mSv and 4.9 mSv [L23].
498. A radiation survey of seven archaeological sites inside
Egyptian pyramids or tombs, conducted in the ­Saggara area,
obtained measurements of radon (222Rn) and its short-lived
decay products, thoron (220Rn) progeny and gamma radiation. In seven of the pyramids and tombs, workers could
receive annual doses ranging from 2 to 13 mSv; in the ­others,
annual doses were less than 1 mSv [B26]. The dose assessment for the workers at two archaeological sites in Alexandria, Egypt, has shown that the effective doses are in therange 0.05–5 mSv/a at both sites [H2]. The estimated average annual effective dose to tour guides at the great pyra­midof Cheops was 0.05 mSv, and estimates for the ­pyramid
guards varied from 0.19 to 0.36 mSv [H1].
499. A programme of radon measurements in Irish schools
has been conducted since 1998. A total of 45,000 individual
radon measurements were made in 3,444 primary and postprimary schools. The average radon concentration was
93 Bq/m3, comparable with the 89 Bq/m3 observed forhomes; the highest concentration measured was 4,948 Bq/
m3. In 74% of the schools, no classrooms had radon concentrations of greater than 200 Bq/m3, while in 9% of theschools, the radon concentration in at least one classroom
exceeded 400 Bq/m3. A total of 591 schools (17% of thoseANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

measured) had radon concentrations of between 200 and
400 Bq/m3. A total of 898 schools (26% of those measured)
will require some degree of remediation to reduce indoor
radon concentrations [C25]. The radiation survey performed
in 25 classrooms in the capital city of Kuwait between September 2003 and March 2004 showed that the annual dose
was about 1 mSv [M1]. In a radon survey in a school withelevated levels of radon in Slovenia, the annual effective
doses received by the staff were estimated to range from 1.3
to 12.6 mSv [V15]. Another radiation survey, in schools on
the territory of an abandoned uranium mine in Slovenia,
found that the annual doses for the staff ranged from 0.07 to
0.27 mSv [V14]. An extensive radon survey was performed
in 890 schools in Slovenia, and radon concentrations with anarithmetic mean of 168 Bq/m3 and a geometric mean of
82 Bq/m3 were found. In 67% of the schools, indoor radon
concentrations were below 100 Bq/m3, while in 8.7% ofthem the concentration exceeded 400 Bq/m3. The averagevalue of the gamma dose rate measurements was 102 nGy/hand the geometric mean was 95 nGy/h [V11, V12].
500. The average annual effective dose to the workers in
94 offices in Hong Kong has been estimated to be 0.35 mSv[Y3]. In the United Kingdom, a study was undertaken
throughout British Telecom underground workplaces during
1993–1994 to assess occupational exposure due to radon.
The study concluded that no British Telecom staff received
an annual radiation dose of greater than 5 mSv [W13]. In
Venezuela (Bolivarian Republic of), the average effective
dose received by the employees of the Caracas subway system has been estimated as about 1 mSv/a [L9] A radiation
survey in 201 rooms of 26 major hospitals in Slovenia gavean estimate of the annual effective doses for 966 staff (94.2%)of less than 1 mSv, but for 10 workers the doses were between
2.1 and 7.3 mSv [V17].
501. Available data from the UNSCEAR Global Survey ofOccupational Radiation Exposures for radon in workplaces
other than mines are included in the last part of table A-16.
Five countries have reported data for the period 2000–2002.
These data show considerable variation for the average effective dose, from 0.7 to 5 mSv. Germany has reported separate
data for spas, waterworks and tourist caves. The average
effective dose for people working in spas, 4 mSv, is twicethat in the other workplaces, 2 mSv, as shown in table 56.
502. Elevated levels of radon have been found in a numberof countries, but the levels of exposure vary considerablyaccording to the workplace. So far the UNSCEAR reports
have performed only crude estimates of the worldwide levelsof exposure, owing to a lack of information. Although thenumber of data available for the last two periods has increased
compared with the previous periods, the sample sizes arestill very small and the levels of exposure depend on factorsthat vary from country to country, such as geology, building
materials and regulatory regimes. There are clearly very few
data on which to base an accurate estimate of worldwide
exposure. Since the scenario of exposure throughout theworld has not changed dramatically since the UNSCEAR

289

2000 Report, the number of exposed workers is estimated as1.250 million, the collective effective dose as about
6,000 man Sv and the average effective dose as 4.8 mSv(table 57). The level of exposure is the same as estimated in
the UNSCEAR 2000 Report [U3]. Clearly this estimate isvery crude.
5.

Conclusions on occupational exposure tonatural sources of radiation

503. After the implementation of the International Basic
Safety Standards [I7] and subsequently the implementationof the European Union standards for the protection of workers exposed to natural radiation (European Union Directive
96/29/Euratom) [E10], data on levels of occupational exposure to natural sources of radiation have become available,
mainly in the European Union countries. Other qualifying
data are needed on specific issues for each category of exposure in order to be able to derive an accurate estimate for theworldwide average levels of exposure to natural sources ofradiation. The highest level of occupational exposure comesfrom exposure to natural sources of radiation.
504. Data have indicated that aircrew are one of the mosthighly exposed occupational groups. In Germany, the collective dose to this group, 60 man Sv, contributes more than 50%of the total collective dose to all workers in the country. The
estimated worldwide collective effective dose to aircrew isabout 900 man Sv. This value is about the same as that­estimated in the UNSCEAR 2000 Report, 800 man Sv [U3].
505. Work activities with materials containing NORM caninvolve significant exposure of workers through internal
contamination by inhalation. However, there can be considerable differences in workplace conditions, the radionuclides
involved and the physical and chemical matrices in which
the radionuclides are incorporated.
506. The level of exposure in mines depends on a numberof factors, including the type of mine, the geology and theworking conditions, particularly the ventilation. The
UNSCEAR 1988 Report [U7] estimated the global collective dose for coal mining as 2,000 man Sv. The UNSCEAR
2000 Report [U3] estimated the collective dose as about
2,600 man Sv, which is about 16% of the present estimate of
16,560 man Sv. For coal mines, the estimated number ofworkers is 6.9 million and the average effective dose is
2.4 mSv. The increase is due to taking into consideration thecontribution of the coal miners in China. The current estimate of the exposure levels for coal miners seems to be
more realistic than the previous ones, since it is based ondata obtained from a comprehensive survey performed in
China, which represents the great majority of the global
workforce. On the basis of the survey programme carriedout in China, the level of exposure appears to be declining,since the annual effective dose fell from 4.8 mSv in 1999 to
2.4 mSv in 2000–2002 [C12, T4]. For non-coal mines, thecollective dose estimate has also increased considerably.

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The UNSCEAR 2000 Report [U3] estimated the collective
dose as about 2,000 man Sv, which is about 14% of the current estimate of 13,800 man Sv. The estimated number ofworkers in non-coal mines is about 4.6 million, and theaverage effective dose is 3.0 mSv. However, for non-coalminers the worldwide estimate is still only rough, since thedata need to be qualified with regard to their completeness,in particular for the number of workers engaged in underground and above-ground mines. The overall estimate formining activities is 30,360 man Sv, which is about seventimes higher than the previous estimate [U3].
507. The SMOPIE project, which dealt with occupational
internal exposures from practices and work activities in
NORM industries in European countries, covered a broad
variety of practical issues, including: the generation of andexposure to dust; whether the exposure is continuous or discontinuous; whether the exposure is worker-induced or­process-induced; and the variation of doses between workers. Several studies have been reviewed, but they do not provide the information required for a scientifically soundevaluation of the problem. The results of the project haverevealed that there still is a severe lack of information on thenumber of exposed workers in NORM industries and on theassociated occupational doses. The number of 85,000 exposed
workers, as derived in this project, warrants more research.
The largest group of exposed workers (70,000) appears to bewelders using thoriated welding electrodes. The available
data suggest that the grinding of welding rods may give riseto annual doses of between 6 and 20 mSv [S4, V1]. There issome evidence that alternative (non-radioactive) welding
rods are increasingly being used. This means that the numberof exposed workers should decrease in the future. A survey
programme in ­Denmark has shown that the annual committed effective dose from the inhalation of 232Th, 230Th, 228Th
and 228Ra, for a full-time TIG (tungsten inert gas) welder, isbelow 0.3 mSv in a realistic case and around 1 mSv or lower
with conservative assumptions. The contribution from grinding electrodes was lower, 0.010 mSv or less [G2]. Again,­precise details on this trend were not available.
508. According to the SMOPIE project, the second largest
group of exposed workers (10,000) are those trading or using
phosphate fertilizers (The data originate from only one country.). The results indicate that, like the grinding of thoriatedwelding rods, zircon milling may also give rise to annual
doses of between 6 and 20 mSv in workplaces where protection measures are poor or non-existent. Rare earth processing may even give rise to annual doses of greater than
20 mSv. In both industries, the number of exposed workers
is small [V1].
509. The results of the occupational exposure survey performed in nine European Union countries from 1996 to 2000
have shown that the average annual effective dose declined
from 6 to 3 mSv during that period. The annual collective
dose fell from 70 to 39 man Sv; therefore the mean value
may be influenced by the increasing number of monitored
workers. The reduction of 71% in the number of workers

receiving annual doses of over 20 mSv is the largest for all
work sectors. There was a substantial change in the dose distribution towards lower values in almost all dose bands.
However, substantial differences exist between the countries
where monitoring was undertaken. There are some uncertainties in this evaluation, since the registered doses may
include uranium miners as well as non-uranium miners or
workers in tourist caves and at drinking water facilities, i.e.
they include external exposures as well as doses from radon
inhalation. The recommendations of ICRP Publication 65[I48] changed the dose calculation substantially by introducing conversion factors and detriment coefficients, as a consequence of which the values of the calculated doses fell
considerably. However, the declining values of the annual
doses may also be a result of modified work management
and workplace conditions [F15]. In conclusion, a declining
level in reported occupational exposures to natural sourcesof radiation in European countries has been seen, althoughsubstantial differences exist between the countries where
monitoring is undertaken [F15].
510. Elevated levels of radon have been found in a numberof countries, but the levels of exposure vary considerably
depending on the workplace. The level of exposure to radon
may be underreported, since the doses for workers in workplaces where the radon concentration is below 1,000 Bq/m 3
may not be reported. So far the UNSCEAR reports have performed only crude estimates of the worldwide levels of
exposure, owing to a lack of information. Although thenumber of data available for the last two assessment periods
has increased compared with the previous periods, the sample sizes are still very small and the levels of exposure
depend on many factors that vary from country to country,
such as geology, building materials and regulatory regimes.
There are clearly very few data on which to base an accurate
estimate of worldwide exposure. Since the scenario of exposure throughout the world has not changed dramaticallysince the UNSCEAR 2000 Report, the same value for theworldwide annual collective effective dose of 6,000 man Sv
is assumed. As in the UNSCEAR 2000 Report [U3], thenumber of workers is estimated to be 1.250 million and theaverage effective dose to be 4.8 mSv. These estimates areclearly very crude.
511. The worldwide level of exposure for workers exposed
to natural sources of radiation has increased considerablycompared with the UNSCEAR 2000 Report [U3]. The estimated number of workers is about 13 million. The estimated average effective dose is 2.9 mSv and the estimatedc­ ollective effective dose is 37,260 man Sv.
C. Man-made sources for peaceful purposes
1.

Nuclear power production

512. A significant source of occupational exposure is theoperation of nuclear reactors to generate electrical energy.
This involves a complex cycle of activities, including theANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

mining and milling of uranium, uranium enrichment, fuelfabrication, reactor operation, fuel reprocessing, waste hand­ling and disposal, and research and development activities.
Exposures arising from this practice were discussed andquantified in the UNSCEAR 1972 [U11], 1977 [U10], 1982
[U9], 1988 [U7], 1993 [U6] and 2000 [U3] Reports, withcomprehensive treatment in the UNSCEAR 1977, 1982 and2000 Reports. In comparison with many other sources of
exposure, this practice is well documented, and considerable
quantities of data on occupational dose distributions areavailable, in particular for reactor operation. This annex considers occupational exposure arising at each main stage ofthe fuel cycle. Because the final stage—treatment and disposal of the main solid wastes—is not yet sufficiently developed to warrant a detailed examination of potential exposures,
it is given only very limited consideration. However, for theperiod under consideration, occupational exposures due towaste disposal are not expected to add a significant amountto the collective exposure of workers to radiation due to theother stages in the fuel cycle.
513. Each stage in the fuel cycle involves different types ofworkers and work activities. In some cases, for example forreactor operation, the data are well segregated, while in others the available data span several activities, e.g. uranium
mining and milling. Where the data span a number of activities, this is noted in footnotes to the tables. The data on occupational exposures for each of the activities are derived
primarily from the UNSCEAR Global Survey of Occupational Radiation Exposures [U3, U6, U7, U9, U10] but also
from other sources, particularly the joint OECD/NEA and
IAEA Information System on Occupational Exposure(ISOE) [O14, O19, O20], which serves as a main source ofoccupational exposure data for reactor operations in the
period 1995–2002.
514. For each stage of the fuel cycle, this annex provides
estimates of the magnitude of and temporal trends in theannual collective and per caput effective doses, the numbersof monitored workers and the “distribution ratios”. The collective doses are also expressed in normalized terms, i.e. perunit practice relevant to the particular stage of the cycle. For
uranium mining and milling, fuel enrichment, fuel fabrication and fuel reprocessing, the normalization is initially presented in terms of unit mass of uranium or fuel produced orprocessed. An alternative way to normalize is in terms of theequivalent amount of energy that can be (or has been) generated by the fabricated (or enriched) fuel. The bases for thenormalizations, i.e. the amounts of mined uranium, the separation work during enrichment and the amount of fuelrequired to generate a unit of electrical energy in various
reactor types, are given in section II.C of this annex. For
reactors, the data may be normalized in several ways,
depending on how they are to be used. In this annex, normalized collective doses are given for each reactor type and perunit electrical energy generated.
515. To allow proper comparison between the doses arisingat different stages of the fuel cycle, all the data are ultimately

291

presented in the same normalized form, in terms of the electrical energy generated (or the amount of uranium mined orof fuel fabricated or reprocessed, corresponding to a unit ofenergy subsequently generated in the reactor), which is theprincipal measure of output of the nuclear power industry.
This form of normalization is both valid and useful whentreating data averaged over a large number of facilities orover a long time. It can, however, be misleading when applied
to data for a single facility for a short time period. This isbecause a large fraction of the total occupational exposure ata facility arises during periodic maintenance operations,when the plant is shut down and not in production. Such difficulties are, however, largely circumvented in this annex,
since the data are presented in an aggregated form for­individual countries and are averaged over five-year periods.
516. Various national authorities or institutions have useddifferent methods to measure, record and report the occupational data included in this annex. The main features of themethod used by each country that responded to the
UNSCEAR Global Survey of Occupational Radiation Exposures are summarized in table A-15. Data collected under
ISOE are provided by participants according to standardized
reporting formats, although the details requested haveincreased over time, and not all countries report to the samelevel of detail. Additionally, the data provided under ISOE
are based on operational data collected from the participating utilities, and may differ slightly from official doserecords. The reported collective doses and the collective
dose distribution ratios are largely insensitive to the differences identified in table A-15, so these quantities can generally be compared without further qualification. The average
doses to monitored workers and the number distribution
ratios are, however, sensitive to the decisions and practices
concerning which workers in a particular workforce are to bemonitored. Differences in these areas could not be discerned
from responses to the UNSCEAR Global Survey of Occupational Radiation Exposures, and they therefore cannot be
discerned from table A-15. However, because the monitoring of workers in the nuclear power industry is in general
fairly comprehensive, comparisons of the average individual
doses (and number distribution ratios) reported here are
judged to be broadly valid. Nonetheless, it must be recognized that differences in monitoring and reporting practices
do exist, and they may, in particular cases, affect the validityof comparisons among reported data. As mentioned before,
the criteria applied in different countries to select workers
who should be monitored differ considerably. Some countries
monitor only the exposed workers, while others also include
the non-exposed workers in their individual ­monitoring­programme for various reasons.
(a)Uranium mining and milling

517. Most natural uranium is mined for energy production
in fission reactors, but it is also used in nuclear research reactors and in military activities. Commercial uranium use is
primarily determined by the fuel consumption requirements

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UNSCEAR 2008 REPORT: VOLUME I

of power reactors and continues to increase steadily, while
the requirements for research reactors remain modest bycomparison.
518. The mining of uranium is similar to that of any othermaterial. It mainly involves underground or open-pit techniques to remove uranium ore from the ground, followed byore processing, usually performed at a location relatively
close to the mine. The milling process involves the crushingand grinding of raw ores, followed by chemical leaching, theseparation of uranium from the leachate and precipitation ofyellowcake [K14], and the drying and packaging of the final
product for shipment.
519. Uranium mining has been conducted in 24 countries(see table 14 for annual uranium production worldwide andsection II.C of this annex) over the period 1998–2003. This
practice has ended in some countries. Between 1990 and
1997, 34 countries were involved in uranium mining, andover the whole nuclear era, some 37 countries [U3]. The
major producer is Canada, which is responsible for about
30% of the world production, followed by Australia with
about 14%, and Niger with about 10%. About 93% of theworld’s production comes from only ten countries: Australia,
Canada, Kazakhstan, Namibia, Niger, the Russian Federation, South Africa, Ukraine, United States and Uzbekistan.
520. In the mining and milling of uranium ores, the workers incur both internal and external radiation exposures.
Mining operations such as drilling, blasting, loose-dressing,mucking, crushing, boulder-breaking, loading and dumping,etc., generate ore dusts of different particle sizes, whichbecome dispersed in the mine environment and give rise to
an inhalation hazard. Concentrations of these ore dusts arequite variable with time and location. Extremely high values
can be reached during blasting and ore dumping. In general,these workplaces are very dusty, and consequently there is apotential risk for inhalation of aerosol particles containing
radionuclides from the 238U decay chain. The internal dose
depends on workplace conditions, which vary considerablyaccording to the type of mine (underground or above ground),the ore grade, the airborne concentrations of radioactive particles (which vary depending on the type of mining operationand the quality of ventilation) and the particle size distribution. In underground mines, the main source of internal
exposure is likely to be radon and its decay products. Becauseof the confined space underground and practical limitations
to the degree of ventilation that can be achieved, the totalinternal exposure is of greater importance in undergroundmines than in open-pit mines. In open-pit mines, the inhalation of radioactive ore dusts is generally the largest source ofinternal exposure, although the doses tend to be low. Higher
doses resulting from this source would be expected in the
milling of the ores and the production of yellowcake. Internal exposure makes by far the greatest contribution to thetotal exposures resulting from underground mining.
521. Exposure data for the mining and the milling of uraniumores from the UNSCEAR Global Survey of Occupational

Radiation Exposures for 1995–2002 are given in tables A-17
and A-18, respectively, and trends for the six periods 1975–
1979, 1980–1984, 1985–1989, 1990–1994, 1995–1999 and
2000–2002 are given in figure XXXVIII.
522. Over the four previous five-year periods (1975–1994),
the average annual amounts of uranium mined worldwide
were 52, 64, 59 and 39 kt. For the periods 1995–1999 and
2000–2002, the average annual amounts mined were 34 kt.
This represents a reasonably constant level of production for
the first three periods and a reduction by about one third forthe last three periods. The average annual amount of ­uranium
mined remained constant over the last three periods.
523. Germany has ceased mining operations; its reported
doses relate to the decommissioning of mines. Other countries, e.g. France and Spain, are in the same situation. Stillother countries, e.g. Argentina, Belgium, Gabon and Hungary,
have completely stopped their uranium production in the last
several years (see table 14).
524. The estimate of worldwide levels of exposure resulting from uranium mining has been derived by scaling up to
the total world uranium production from the 36% of production for which data were reported. For the reported data,
Canada dominates, accounting for about 30% of the world
uranium production. On this basis, the average annual
number of monitored workers worldwide has decreased dramatically over time: 240,000, 310,000, 260,000, 69,000 in
the first four periods (1975–1979, 1980–1984, 1985–1989
and 1990–1994), compared with 22,000 and 12,000 in thelast two periods (1995–1999 and 2000–2002). These reductions by a factor of 3 and 6 in the last two periods are also
seen in the values for average annual collective effective
doses. For the first four periods the worldwide estimates
were 1,300, 1,600, 1,100 and 310 man Sv, but for 1995–1999
and 2000–2002 the values fell to 85 and 22 man Sv, respectively. Similarly, the average collective dose per unit mass ofuranium extracted was 26, 23, 20 and 8 man Sv/kt for the
first four periods and declined to 2 and 1 man Sv/kt for
1995–1999 and 2000–2002, respectively. However, the estimated average annual effective doses have been high over
the years, even though they started to decrease in the last two
periods: they decreased from 4.5 mSv in 1990–1994 to
3.9 mSv in 1995–1999 and to 1.9 mSv in 2000–2002. The
average effective dose for measurably exposed workers has
decreased significantly as well. The data are consistent witha worldwide reduction in underground mining activity coupled with more efficient mining operations. The trends arepresented in table 58 and are represented graphically infigure XXXVIII.
525. In order to evaluate the occupational exposure inunderground and above-ground mines, a new questionnaire
was distributed requesting the data to be provided separately.
Canada and Germany have reported data separately forabove-ground and underground mines. The data are presented in table 59. The effective doses were in the range
0.3–1.3 mSv for above-ground mines and 1.0–3.1 mSv forANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

underground mines. Effective doses for the workers in theunderground mines are at least twice as high as those in theabove-ground mines. The data reported by Germany forunderground mines are related to the decommissioning ofmining facilities. The doses reported by Canada show adecreasing number of monitored workers and decreasingcollective dose and average effective dose for undergroundminers. The major reason for the reduction in the level ofoccupational exposure in Canada is that uranium mining
moved from the conventional cut-and-fill method used tomine ore grades of around 0.1% U in northern Ontario to themore advanced, non-entry type of method used to mine thehigher-grade ores (some exceeding 20% U) in northern Saskatchewan. These non-entry mining methods significantly
reduced gamma radiation exposures and greatly restricted
exposure to radon progeny and uranium ore dust.
526. The contribution of internal and external exposure to
the total effective dose has been analysed in this annex on thebasis of data provided by Canada, the Czech Republic andGermany for each type of mine. The percentage dose contributions from radon and ore dust inhalation and from external
exposure are given in table 60. The contribution of eachsource varies according to the type of mine and the ore grade.
However, internal exposure is the main contributor to thetotal effective dose, independent of the type of mine, and its
overall contribution is about 70%.
527. According to the Canadian Occupational Radiation
Exposures reports [H9, H10, H11, H12, H13, H14], radiation exposure is significantly higher for underground mining
workers than for surface mining workers. It also differs considerably according to job function. The contribution ofradon exposure to the total effective dose is about 60%, independent of the type of mine. As shown in table 61, the annual
effective dose to the more exposed miner job category in
Canada, averaged over the period 1995–2001, fell from
11 mSv to 2 mSv for underground mines and rose from
1 mSv to 2 mSv for above-ground mines.
528. For the period 1996–2000, the average annual doses
received by workers at three underground uranium mines inIndia were around 8 mSv. The main contribution to the effective dose came from inhalation of 222Rn and its short-lived
progeny [K10].
529. The assessment of exposure of miners to the longlived α-emitting radionuclides associated with respirable
ore dusts in the Jaduguda uranium mine in India, where theU3O8 concentration is less than 1%, has shown that theinhalation of ore particles has contributed only about 5% of
the annual effective dose limit, indicating that in this mine
it is not a significant source of exposure [J3]. At ore gradesof up to about 3% U3O8, limitation of airborne silica willusually place a stricter constraint upon dust concentration
than does radioactivity. However, at ore grades in excess of
3% U3O8, and when the ore is not high in silica, radiation
exposure resulting from inhalation of ore dust could become
important.

293

530. Data on exposure of workers due to uranium millingare presented in table A-18. The Committee assumes thatthe amount of uranium milled is equal to the amount mined.
The estimated worldwide level of exposure has decreased
over the six periods: (a) the average annual number of monitored workers was 3,000 for the periods 1995–1999 and
2000–2002, which is substantially fewer than in the previous four periods: 12,000, 23,000, 18,000 and 6,000; (b) the
average annual collective effective dose was 4 man Sv and3 man Sv for 1995–1999 and 2000–2002, respectively, compared with 124, 117, 116 and 20 man Sv for the previous
four periods; (c) the average annual effective dose was
1.6 mSv and 1.1 mSv for 1995–1999 and 2000–2002,
respectively, compared with 10.1, 5.1, 6.3 and 3.3 mSv forthe previous four periods. The data are ­presented in table 62
and figure XXXIX.
(b)

Uranium conversion and enrichment531. Uranium conversion is the process by which UO2,
which is the chemical form of uranium used in most commercial reactors, is produced for the fabrication of reactor
fuel. Some reactors use fuel slightly enriched in 235U (generally about 3% enrichment, in contrast to natural uranium,
which contains about 0.7% 235U). The U3O8 from the milling
process is converted to UO2 by a reduction reaction with H2.
The UO2 is converted to UF4 by the addition of hydrofluoric
acid (HF) and then to UF6 using fluorine (F2). The gaseous
product, uranium hexafluoride (UF6), is then enriched in
235
U. Most of this is performed by the gas diffusion process,but gas centrifuge techniques are being used increasingly.
Once the enrichment process has been completed, the UF6
gas is reconverted into UO2 for fuel fabrication [U3].
532. In 2003 there were 29 uranium conversion/recovery
facilities and 21 uranium enrichment facilities in operation.
The enrichment capacity of these facilities and a few othersmall producers is presented in section II.C of this annex.
The greater part of the enrichment services came from five
suppliers: the United States Department of Energy, Eurodif(France), Techsnabexport (Russian Federation), Urenco(Germany, Netherlands and United Kingdom) and China[X1]. Most thermal reactors use enriched uranium with typically a 3% level of enrichment. Four types of uranium fuelwill be considered: unenriched metal fuel, used in Magnox
reactors; low-enriched oxide fuel, used in AGRs and LWRs;
unenriched metal fuel, used in HWRs; and mixed oxide fuel,used in FBRs. Mixed oxide (uranium–plutonium) fuels are
increasingly being developed for use in LWRs.
533. Exposure data for this practice are given in table A-19.
The average annual number of monitored workers increased
from 12,600 in 1990–1994 to about 18,000 in 2000–2002.
The average annual collective dose has increased from 1.28
to 1.70 man Sv. The average annual effective dose to monitored workers was low, 0.1 mSv, in 1995–2002, and has notchanged since 1985–1989. The absence of data from theRussian Federation would suggest that these figures are

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UNSCEAR 2008 REPORT: VOLUME I

underestimates. Even taking this into account, the individualand collective doses arising from enrichment are low. The
trends in this practice are presented in table 64 andfigure XL.
534. Occupational exposure occurs during the enrichment
and conversion stages of the fuel cycle. External radiation
exposure is more important than internal radiation exposure,but workers may be exposed to internal radiation, particularly during maintenance work or in the event of leaks. The
workers may be exposed to UF6, classified as a soluble compound and assigned as Type F for lung retention, according
to the ICRP [I51, I57]. In these situations, the occupational
exposure to daily intakes of these uranium compounds ofany isotopic composition would be limited by considerationsof chemical toxicity rather than radiation dose [I55, S40]. A
new questionnaire was distributed to Member States toobtain information about the contribution of internal exposure to the total effective dose. Data from China show that
64% of the dose is due to external exposure. The contribution of each source varies according to the level of exposure.
The data provided by China show that, for effective doses
lower than 1 mSv, the contributions of internal and external
exposure are about the same. For effective doses higher than
1 mSv (1–5 mSv), the contribution of internal and external
exposure is about 17% and 83%, respectively.

537. The increase in the average effective dose may have
two possible reasons: the inclusion of new countries that
contributed higher levels of exposure in these last two periods, and the fact that some countries began to include thedose due to internal exposure in their dose records. There are
two main sources of exposure in the fabrication of nuclear
fuels: external exposure to gamma radiation and internal
exposure resulting from the inhalation of airborne material.
China has provided information that the doses below 1 mSv
are entirely due to external exposure. However, for the doses
above 1 mSv, there is an important contribution from internal exposure (between 30% and 80%). According to an NRC
report on fuel fabrication facilities, internal exposure contributes most of the total effective dose, up to about 99% of
the total dose [U29, U30, U31, U32, U33, U34, U36, U37].
However, the internal dose component depends on the typeof nuclear fuel. The occupational exposure in the productionof nuclear fuel is expected to be lower for fuel that involves
only natural uranium than for fuels that involve enriched uranium or plutonium. In conclusion, the type of dose that isrecorded in the national databases can be a source of discrepancy among countries. Some countries record only the doses
from external exposure and others record the doses due toboth internal and external exposure. Some countries also
include in their individual monitoring programme workers
who do not work in controlled areas. The variation in typesof nuclear fuel also influences the comparison of doses
between countries.

(c) Fuel fabrication
535. The characteristics of fuels that are relevant here are
the degree of enrichment and the form, either metallic oroxide. The majority of reactors use low-enriched fuel (typically 3–5% 235U). The main exceptions are the gas-cooled
Magnox reactors and the HWRs, which use natural uranium.
Some older research reactors use high-enriched uranium (upto 98%); however, for security reasons this material is being
used ever less frequently. The principal source of exposure
during fuel fabrication is uranium (after milling, enrichment
and conversion, most decay products have been removed).
536. Exposure data for fuel fabrication are given in table
A-20. The average annual number of monitored workers has
been reasonably constant over the six periods at about 20,000
but with a small peak of 28,000 in the 1985–1989 period. The
worldwide average annual number of measurably exposed
workers has been approximately 10,000, about half thenumber of monitored workers. The estimated average annual
collective dose showed a decline, from 36 to 21 man Sv,
between the first two five-year periods, showed little changeover the next two periods, with the value for 1990–1994 being
approximately 22 man Sv, and then increased to about
30 man Sv for the last two periods. The average annual effective dose to monitored workers showed an initial decline,from 1.8 to 1.0 mSv, between the first two periods, and thevalue for 1990–1994, 1.0 mSv, is very similar to that for
1980–1984. For the last two periods the average effective
dose increased by about 60%. The trends in occupational
exposure are presented in table 65 and figure XLI.

(d)Reactor operation

538. The types of reactor used for electrical energy generation are characterized by their coolant system and moderator:
light-water-moderated and -cooled pressurized- or boilingwater reactors (PWRs, BWRs); pressurized heavy-watermoderated and -cooled reactors (HWRs); gas-cooled,graphite-moderated reactors (GCRs), in which the gas­coolant, either carbon dioxide or helium, flows through asolid graphite moderator; and light-water-cooled, graphite-­moderated reactors (LWGRs). These are all thermal reactors,
in which the moderator material is used to slow down fast
fission neutrons to thermal energies. Fast-breeder reactors(FBRs) at present make only a minor contribution to energy
production. Between 1990 and 1994, the number of operating
reactors remained relatively stable, increasing slightly from
413 to 432 by the end of the period. A listing of nuclear reactors in operation during the period 1990–1997, the installedcapacities and the electrical energy generated is given inannex C of the UNSCEAR 2000 Report [U3], “Exposures tothe public from man-made sources of radiation”. At the endof 1997, there were 437 nuclear power reactors operating in
the world, with a capacity of about 352 GW(e) (net gigawattsof electrical power) [I8]. For the period 1998–2002, thenumber of nuclear reactors in operation, the installed capacities and the electrical energy generated are given in section
II.C.1 of this annex. The average number of power reactors
operating in the world over the period 1998–2002 was 444,with an average capacity of about 278 GW(e).

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

539. In addition to data provided in response to the
UNSCEAR Global Survey of Occupational Radiation Exposures, data on exposures of workers at nuclear power reactors are also available from the ISOE database [O14, O19,
O20]. The ISOE occupational exposure database includes
information on occupational exposure levels and trends for
401 operating reactors in 29 countries, covering about 91%
of the world’s operating commercial reactors [O22]. The
ISOE data on occupational exposures at nuclear power reactors for 1990–2002 [O19, O20] and data from the UNSCEAR
Global Survey of Occupational Radiation Exposures combined with information provided in the UNSCEAR 2000
Report [U3] for the various types of reactor are given intable A-21.
540. Occupational exposures can vary significantly fromreactor to reactor and are influenced by such factors as reactor size, age and type. Several different broad categories ofreactor are currently in operation, including PWRs, BWRs
and GCRs (which include older Magnox reactors), as well as
a newer generation of reactors, AGRs, HWRs and LWGRs.
Within each category, there is much diversity in design andin refuelling schedule, which may contribute to differencesin occupational exposure. In addition, changes in operating
circumstances can alter the exposure at the same reactor
from one year to the next. Some of these variations will be
discussed in this section.
541. The type of reactor is only one of the factors influen­cing the doses received by workers. Other basic features of thereactor play a role, including the piping and shielding configuration, fuel failure history, reactor water chemistry, and theworking procedures and conditions. All of these can differ
from site to site, even among reactors of the same type, contributing to the differences seen in occupational exposures. At
all reactors, external irradiation by gamma rays is the mostsignificant contributor to occupational exposures. The exposures occur mostly during scheduled maintenance and/or refuelling outages. For the most part, such exposures are due toactivation products (60Co, 58Co, 110mAg); however, when fuelfailures occur, fission products (95Zr, 137Cs) may also contribute to external exposures. At BWRs, workers in the turbinehall incur some additional external exposure due to 16N, an
activation product with an energetic gamma ray that is carriedby the primary circulating water through the turbines. In
HWRs, heavy water is used as both coolant and moderator.
Neutron activation of deuterium produces a significant amountof tritium in these reactors, so in addition to the usual external
exposures, workers may also receive ­internal ­exposures due totritium, which is a pure beta emitter.
542. Throughout the world, occupational exposures atcommercial nuclear power plants have been steadily decreasing over the past decade, and this trend is reflected in the datafor 1995–2002. Regulatory pressure (particularly after theissue of ICRP Publication 60 [I47] in 1991), technologicaladvances, improved plant designs, installation of plant
upgrades, improved water chemistry, improved plant operational procedures and training, the involvement of staff in the295

control of their own doses, and international sharing of
ALARA data and experience have all contributed to thisdecreasing trend. Globally, ISOE includes the world’s largest database on occupational exposures at nuclear power
plants and provides an international forum for radiation protection experts from both utilities and national regulatory
authorities to discuss, promote and coordinate international
cooperative undertakings in the area of worker protection at
nuclear power plants [O5, O6, O7, O8, O9, O10, O11, O12,
O13, O14, O15, O18, O19, O20].
543. Data on occupational exposures for reactors of eachtype are detailed by country in table A-21, and a worldwide
summary by reactor type is given in table 66. Worldwide levels of exposure have been estimated from the data provided;
the extrapolations are based on the total energy generated incountries providing data. Very little extrapolation wasneeded, as the data provided were substantially complete(about 96% for PWRs, 99% for BWRs, 63% for HWRs,
100% for GCRs and 13% for LWGRs). Data provided
through ISOE for 1995–2002 are included as provided by
ISOE participants. With a few exceptions, the ISO programme includes essentially all reactors worldwide. The
annual data reported in response to the UNSCEAR Global
Survey of Occupational Radiation Exposures have been
averaged over five-year periods, which provide the average
effective dose and the number of monitored workers. The
ISOE data provided from 1995 to 2002 are given as averages
over the periods 1995–1999 (five years) and 2000–2002(three years), and provide estimates for the collective effective dose. Figures XLII and XLIII illustrate some of thetrends. Previous UNSCEAR reports treated FBRs and hightemperature graphite reactors (HTGRs) separately. No data
were provided on these in either the ISOE database or the
responses to the UNSCEAR Global Survey of Occupational
Radiation Exposures, and in the main these types of facility
are no longer operational. The UNSCEAR 1993 and 1988
Reports [U6, U7] concluded that they make a negligible contribution to occupational exposure, and therefore they are
not considered further.
544. The UNSCEAR 1993 Report [U6] identified the needfor more data on measurably exposed workers, as these provide a better basis for the comparison of average doses toindividuals than is possible using the monitored worker data.
The UNSCEAR Global Survey of Occupational Radiation
Exposures now provides good data on measurably exposed
workers for PWRs, BWRs and HWRs (see table A-21). The
vast majority of the GCRs are in the United Kingdom, andwhile data matching the definition of “measurably exposed”
are not readily available, a good data set showing dose distribution is available from the United Kingdom’s Central Index
of Dose Information (CIDI) [H8].
545. The procedures for the recording and inclusion ofdoses incurred by transient or contract workers may differ
from utility to utility and country to country, and this may
influence the statistics in different ways. In some cases, transient workers may appear in the statistics for a given reactor

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UNSCEAR 2008 REPORT: VOLUME I

several times in one year (whereas they should rather appear
only once, with the summed dose being recorded). If appropriate corrections are not made, the statistics so compiled
will inevitably overestimate the size of the exposed workforce and will underestimate the average individual dose, as
well as the fraction of the workforce receiving doses above
the prescribed levels and the fraction of the collective dose
arising from these doses. This will only be important where
extensive use is made of transient workers and where no­centralized reporting database is used.
546. Countries also differ in how they present informationon the exposures of workers at nuclear installations. The
majority present statistics for the whole workforce, i.e.
employees of the utility and contract workers, often withseparate data for each category. Other countries provide datafor utility employees only, whereas still others present thecollective dose for the total workforce but individual doses
for the utility employees only. Where necessary and practicable, the data provided have been adjusted to allow them tobe fairly compared with other data; these adjustments areindicated in the respective tables.
(i)549. The average number of PWRs worldwide increased
from 78 in 1975–1979 to 266 in 2000–2002. The corresponding increase in average annual energy generated has beensomewhat greater, from 27 to 191 GW a. The number ofmonitored workers at PWRs increased from about 63,000 in
1975–1979 to 283,000 in 2000–2002 (see figure XLII andtable 66). Between the first two periods, the average annual
collective effective dose increased by a factor of about 2,
from 220 to 450 man Sv. A further small increase to
500 man Sv occurred in the third period, followed by areduction to 415 man Sv in the fourth period. The dose
increased again to 506 man Sv and finally decreased to
415 man Sv. Although the number of reactors increased by afactor of around 2 between 1980 and the last period, the collective dose has remained between 400 and 500 man Sv. To
see the underlying trend in the efficiency of radiological protection measures in both design and operational procedures,
it is more instructive to look at the normalized annual collective dose. Per reactor this increased from 2.8 to 3.3 man Sv
over the first two periods but has since dropped to about 2.0
in the last four periods (2.3, 1.7, 2.0 and 1.6 man Sv). The
corresponding values for collective effective dose divided by
the energy generated are (in chronological order to 2002)
8.1, 8.0, 4.3, and 2.8, 3.0 and 2.2 man Sv/(GW a).

Light-water reactors

547. PWRs constitute the majority of the installed nuclear
generating capacity for the period 1998–2002, followed by
BWRs. Averaged over the whole period, about 91% of the totalenergy was generated in LWRs (of this, about 67% was from
PWRs and 24% from BWRs), with contributions of about
4.5% for HWRs, 1% for GCRs and 3.5% for LWGRs. FBRs
contribute only about 0.1% of the total energy generated. Experience has shown that there are significant differences between
occupational exposures at PWRs and those at BWRs. Each
type of reactor is therefore considered separately.
548. PWRs. External gamma radiation is the main sourceof occupational exposure at PWRs. Since in general only asmall contribution comes from internal exposure, the latter isonly rarely monitored. The contribution of neutrons to theoverall level of external exposure is insignificant. Most occupational exposures occur during scheduled plant shutdowns,
when planned maintenance and other tasks are undertaken,
and during unplanned maintenance and safety modifications.
Activation products, and to a lesser extent fission products
within the primary circuit and coolant, are the main sourceof external exposure. The materials used in the primary circuit, the primary coolant chemistry, the design and operational features of the reactor, the extent of unplanned
maintenance, etc., all have an important influence on themagnitude of the exposure resulting from this source. The
significant changes that have occurred with time in many ofthese areas have affected the levels of exposure. One of themost important non-standard maintenance operations that is
associated with significant dose is the replacement of steam
generators. Data on the collective doses associated withmaintenance have been collected by the OECD/NEA [O9]and are given in table 67.

550. The average annual effective dose to monitored workers fell consistently over the first four periods, being 3.5, 3.1,
2.2 and 1.3 mSv, and then increased to 1.9 and 1.7 mSv inthe last two periods, an overall reduction of about one half.
Overall, the average annual effective dose to measurably
exposed workers was about 2.7 mSv for 2000–2002. The
dose distribution data also parallel the downward trend indoses, with both NR15 and SR15 consistently dropping to <0.01
and 0.06, respectively, for the period 2000–2002.
551. There is considerable variation in the worldwide average values with respect to both the trends and the levels ofdose in individual countries. In some cases this variationreflects the age distribution of the reactors and the build-upof activity in the cooling circuits. In other cases the reason
for it is less obvious. More detailed analysis is contained in
the various OECD annual reports [O5, O6, O7, O8, O9, O10,
O11, O12, O13, O15, O18, O20].
552. BWRs. External radiation is also the main source ofoccupational exposure in BWRs, with most exposures arising during scheduled shutdowns, when planned maintenanceis undertaken, and during unplanned maintenance and safety
modifications. By far the largest numbers of BWRs arelocated in the United States and Japan.
553. Worldwide, the average number of BWRs increased
from about 50 in 1975–1979 to about 90 in 2000–2002; thecorresponding increase in the average annual energy generated worldwide was somewhat greater, from about 15 to67 GW a. Overall, 40% of this energy was generated by
BWRs in the United States, 25% in Japan, 16% in Germany
and Sweden, and the remaining 19% in other countries. On
the basis of the UNSCEAR Global Survey of Occupational

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Radiation Exposures, the number of monitored workers at
BWRs worldwide increased from about 59,000 to about
160,000 at the end of period four, and then decreased to
144,000 and 113,000 in the last two periods (see figure XLII
and table 66). The average annual collective effective dose
increased from about 280 to about 450 man Sv between thefirst two five-year periods. It subsequently decreased in thethird and fourth periods to about 330 and 240 man Sv, notwithstanding a twofold increase in the energy generated overthe same period. For the last two periods the values were 237
and 160 man Sv. The normalized average annual collective
effective dose per reactor initially rose from 5.5 to 7.0 man Sv
over the first two periods, but dropped to 4.0 and then to
2.7 man Sv in the next two periods, remained constant at
about 2.6 man Sv in 1995–1999, then finally decreased to
1.8 man Sv per reactor. The corresponding values normalized to the energy generated were 18, 18, 7.9, 4.8, 3.8 and2.4 man Sv/(GW a). Both sets of values indicate significant
reductions over the six periods.
554. The average annual effective dose to monitored workers over the six periods has consistently fallen: 4.7, 4.5, 2.4,
1.6, 1.7 and 1.4 mSv. There has been a reduction by a factor
of about 3 overall. The worldwide average annual effective
dose to measurably exposed workers, 2.1 mSv, is about 50%
higher than that to monitored workers. The declining trend indoses is also seen in the values of NR15 and SR15, with thefraction of the collective dose delivered at individual doselevels of above 15 mSv having been 0.09 in 2000–2002.
555. There is considerable variation in the worldwide
average values with respect to both the trends and the levelsof dose in individual countries. However, the differences
seem to be decreasing over time, and for the vast majority of­countries reporting, a downward trend is apparent.
(ii) Heavy-water reactors
556. The worldwide average number of HWRs increased
from 12 in 1975–1979 to 39 in 2000–2002. The corresponding increase in the average annual energy generated worldwide was somewhat greater, from about 3 to 13 GW a. The
number of monitored workers in HWRs worldwide increased
from about 7,000 to about 20,000 over the first four periods,
and remained about the same over the last two periods, as
shown in figure XLII and table 66. The average annual collective effective dose increased from about 32 man Sv in thefirst five-year period to about 46 and 60 man Sv in the second and third periods. In the fourth and fifth periods, however, it decreased significantly, to 35 and 29 man Sv, before
increasing again to 38 man Sv in the period 2000–2002. The
normalized average annual collective effective dose per reactor decreased slightly, from 2.6 to 2.3 man Sv, over the first
three periods, and then dropped to 1.1 man Sv and remained
constant. The corresponding values normalized by the energy
generated fell by a factor of almost 4, from 11 to 2.9 man Sv/(GW a), over the six periods. Both sets of values indicate
significant reductions over the six periods.

297

557. The average annual effective dose to monitored workers fell from 4.8 to 3.2 mSv over the first two periods but
remained about the same for the third period. For the fourth
period it fell significantly to 1.7 mSv, and became steadyagain in the last two periods at 1.6 mSv. As before, the reduction overall was by a factor of about 2. The doses due to theintake of tritium (as tritiated water) may have provided an
important contribution, around 20%, to the total effective
dose [H15].
(iii)

Gas-cooled reactors

558. There are two main types of GCR: Magnox reactors,
including those with steel pressure vessels and those withprestressed concrete pressure vessels; and AGRs. Anothertype, HTGRs, reported previously [U7], is no longer inoperation. Most of the experience with GCRs has beenobtained in the United Kingdom, where they have been
installed and operated for many years. Initially the GCRs
were of the Magnox type, but throughout the 1980s, thecontribution of AGRs, in terms of both installed capacityand energy generated, became more important. The relative
importance of AGRs will increase as Magnox reactors aredecommissioned.
559. The UNSCEAR 1993 Report [U6] investigated thedifferences in occupational exposures between Magnox
reactors and AGRs. These arise mainly from the use of concrete (as opposed to steel) pressure vessels in the AGRs (andthe later generation of Magnox reactors) and from theincreased shielding they provide against external irradiation,
the dominant source of occupational exposure. The
UNSCEAR 1993 Report identified significant differences
between the various types, with the average annual effective
dose in first-generation Magnox reactors steel pressure vessels remaining uniform at about 8 mSv, whereas the values
for Magnox reactors with concrete pressure vessels and for
AGRs were less than 0.2 mSv. During the 1990–1994 period,
significant dose reductions were made at the Magnox reactors, with further reductions during the last two periods.
More detailed information can be found in the reviews ofradiation exposures in the United Kingdom [H26, H27]. In
this annex no distinction has been made in table 66 between
the various types of GCR.
560. The worldwide number of GCRs averaged over fiveyear periods has decreased significantly from 40 in 1975–
1979 to 23 in the last period (2000–2002). Some reactors
have been shut down. The number of monitored workers asprovided by the UNSCEAR Global Survey of Occupational
Radiation Exposures increased overall, from 13,000 in thefirst period to 30,000 in the fourth, and then decreased to
21,000 and 18,000 in the last two periods, as shown in figure XLII and table 66. The average annual collective effective dose dropped over the six periods, being 36, 34, 24, 16,
7 and 4 man Sv. The normalized collective dose per reactor
also decreased over this period, from 0.9 to 0.2 man Sv. The
corresponding values for energy generation decreased in the298

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first five periods from 6.6 to 0.7 man Sv/(GW a) and thenincreased to 2.6 man Sv/(GW a). The worldwide averageannual effective dose to monitored workers, averaged over
five-year periods, fell progressively from 2.8 mSv in the first
period to 0.2 mSv in 2000–2002. The fraction of the monitored workforce receiving annual doses in excess of 15 mSv
has been small, falling from 0.02 by a factor of over 100.
Between 1992 and 1994 there was only one instance of aworker at a United Kingdom GCR incurring a dose of over
15 mSv in a year, and only ten workers received doses ofover 10 mSv in a year [H27].
(iv) Light-water-cooled, graphite-moderated reactors
561. LWGRs were developed in the former Soviet Unionand have only been installed in what are now the RussianFederation, Ukraine and Lithuania. Only data equivalent to
13% of the total energy generated by LWGRs were provided in response to the UNSCEAR Global Survey ofOccupational Radiation Exposures, ISOE [O19] and othersources [R22].
562. There is no information available to estimate theworldwide level of exposure. The overall number of LWGRs
increased from 12 in the first period to 20 during 1990–1994.
The corresponding average annual energy generation
increased from 4.4 to 9 GW a. The number of monitored
workers increased over the first three periods from about
5,000 to 13,000, but no data are available for the last three
periods. The average annual collective effective dose
increased significantly, being 36, 62, 173 and 190 man Sv
for the four periods from 1978 to 1994. The value for theoccupational exposure given here corresponds to the period
from 1978 to 1994. There are insufficient data for reliable
extrapolation to subsequent periods.
563. It was suggested in the UNSCEAR 1993 Report [U6]that the large increase in collective dose between the second
and third periods (62 to 170 man Sv) was artificial in that thedata included a significant component from the after-effects
of temporary work at Chernobyl. However, the data for
1990–1994 show another increase in exposure. Also, thedata from Lithuania tend to support the overall high levels ofoccupational exposure.
(v)

Summary

564. Data on occupational exposure at reactors worldwide
are summarized in table 66. The worldwide number of powerreactors averaged over the six periods increased from about
190 in the first period to 444 in 2000–2002. The corresponding increase in average annual energy generation was from55 to 278 GW a. Averaged over the whole period, about 91%
of the total energy was generated in LWRs (of this, about
67% was from PWRs and 24% from BWRs), with contributions of about 4.5% for HWRs, 1% for GCRs and 3.5% for
LWGRs. The number of monitored workers increased fromabout 150,000 to 530,000 in the fourth period and decreased
to about 440,000 in the last period. The period 1990–1994 is
the first for which a reasonably robust estimate of the numberof measurably exposed workers, some 290,000, is available;
this value dropped to 170,000 for the period 1995–2002.
565. The annual collective effective dose averaged for eachof the six periods increased over the first three periods (600,
1,000 and 1,100 man Sv) but fell back to 900, 800 and
600 man Sv in the last three periods. The trends in annual
values are shown in table 66 and figure XLIII. About 93% of
the collective dose was received by workers at LWRs. Averaged over all the periods, the contribution from workers at
HWRs was 6%, at GCRs 1% and at LWGRs about 13%.
LWGRs were not considered in this evaluation.
566. The normalized collective effective dose per reactor
averaged over all reactors rose over the first two periods,
from 3.1 to 3.7 man Sv, but dropped to 2.8, 2.1, 1.5 and1.1 man Sv over the remaining periods. The corresponding
figures per unit energy generated are 10.9, 10.4, 5.7, 3.9, 2.5
and 2.5 man Sv/(GW a) for the six periods. A generally
decreasing trend is apparent for both normalized figures formost reactor types. The exception is LWGRs, for which aroughly threefold increase was seen over the first four
periods.
567. The annual effective dose to monitored workers averaged over all reactors fell steadily, from 4.1 to 1.0 mSv. This
number may be an underestimate, since the data for LWGRs
are not included in the dose estimate of the last periods. This
downward trend in annual dose to monitored workers is evident for each reactor type except LWGRs, although there aresome differences between reactor types in the magnitudes ofthe doses and in their rates of decline.
568. Data on the distribution ratios NR15 and SR15 are lesscomplete than data for other quantities, but for 1990–1994
more dose profile information is available for dose bands up
to 1, 5 and 10 mSv. Values of NR15 and SR15 averaged over allreported data are given in table A-21. They show the fractionof monitored workers receiving doses in excess of 15 mSv to
be about 0.08 in the first period, decreasing to 0.02 in 2000–
2002. The corresponding fraction of the collective dose arising from doses in excess of 15 mSv decreased from 0.60 to
0.13.
569. Information on doses according to job category has
been provided through the ISOE programme (table 67). To
account for some inhomogeneity in the statistical recording
systems, these data have been aggregated into five broad categories: refuelling, maintenance, inspection, servicing and a
fifth category covering all other tasks. The data available atthe job level in the ISOE database cover European reactors,
as well as those in China, South Africa and Brazil.
570. In general, the annual doses associated with most jobs,
regardless of reactor type, decreased from the 1995–1999
period (five years) to the 2000–2002 period (three years).

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

For PWRs, the standardized annual dose due to refuelling
showed a 20% decrease. Decreases in doses due to maintenance jobs and servicing jobs of 30% and 10%, respectively,
were also observed. A 38% increase in the annual dose due
to inspection for all countries over the period 2000–2002
resulted partially from an increase in the number and frequency of controls in France and Germany. For BWRs, doses
associated with annual maintenance, inspection and servi­
cing decreased by about 30%, while the annual dose due torefuelling (which is more standardized) remained stable oreven showed a small increase.
571. The evolution of doses due to maintenance and servicing for both PWRs and BWRs is the consequence ofmany factors, including reductions in the duration of refuelling outages and in the number of tasks performed, theimplementation of ALARA programmes and better job
preparation.
(e)Decommissioning

572. Some nuclear power plants are already in a phase of
decommissioning. The workers involved in this process may
receive internal and external exposures. When nuclear reactor facilities are dismantled as part of decommissioning a
nuclear power station, radioactive dust is generated. This
radioactive dust is likely to diffuse in the working environment, resulting in internal exposure. However, most of thedose comes from external exposure. There are few publisheddata available on doses due to decommissioning. Data from
13 nuclear power plants in the United States show that about
2,000 workers were involved in this process in the period
1995–2002. The average annual effective dose for the measurably exposed workers was around 2 mSv and the averageannual collective dose was around 4 man Sv. These data arepresented in table 68. The available data are not sufficient forevaluating the worldwide level of exposure.
(f)Fuel reprocessing

573. The principal reason for reprocessing has been to
recover unused uranium and plutonium in the spent fuel elements. A secondary reason is to reduce the volume of material to be disposed of as high-level waste. In addition, thelevel of radioactivity in such “light” waste after about 100
years falls much more rapidly than in spent fuel itself. The
practice is conducted in only a few countries: France and theUnited Kingdom have with commercial-scale facilities,
Japan and India have experimental facilities, and the RussianFederation has been reprocessing fuel for reactors developedin that country [U3]. In the last decade, interest has grown inseparating (“partitioning”) individual radionuclides both to
reduce long-lived radionuclides in residual waste and to beable to transmute separated long-lived radionuclides into
shorter-lived ones. Reprocessing to recover uranium andplutonium avoids wasting a valuable resource, because mostof the spent fuel (uranium at less than 1% 235U and a little

299

plutonium) can be recycled as fresh fuel, saving some 30%of the natural uranium that would otherwise be required. It
also avoids leaving the plutonium in the spent fuel, where ina century or two the radiological hazard from other components will have diminished significantly, possibly allowing
the plutonium to be recovered for use in weapons.
574. Spent fuel assemblies removed from a reactor are
highly radioactive and produce heat. They are therefore put
into large tanks or “ponds” of water, which cools them and,with three metres of water over the assemblies, shields theradiation they emit. They remain for a number of years either
at the reactor site or at the reprocessing plant, and the levelof radioactivity decreases considerably with time. For mosttypes of fuel, reprocessing occurs at any time from 5 to
25 years after the fuel is unloaded from the reactor.
575. The exposure data for 1995–1999 and 2000–2002
are given in table A-22. In the earlier period, the contribution of 33.9 man Sv from the Russian Federation accounted
for over 50% of the worldwide average annual collective
effective dose.
576. The estimate for the worldwide level of exposure wasbased on the trends in the data from the reporting countries.
The number of monitored workers has increased over the six
periods, from 8,000 in the first period to 76,000 in the last.
The collective effective dose dropped from 53 man Sv in
1975–1979 to 36 man Sv in 1985–1989, increased to
67 man Sv in the following period, and remained steady inthe last two periods. The effective dose has decreased progressively, from 7.1 mSv in the first period to 0.9 mSv in thelast period. The trends are presented in table 69 and figure XLIV. The increase in collective dose is associated withincreased numbers of workers.
(g)

Research related to the nuclear fuel cycle

577. It is difficult to estimate the levels of occupational exposure that can unequivocally be attributed to research and development related to the commercial nuclear fuel cycle. Few data
are available separately for this category; even when they are,uncertainties remain as to their proper interpretation.
578. Occupational exposures arising in nuclear research
are presented in table A-23. There is considerable variationin the levels of collective dose associated with research
activities in each country, reflecting, among other things, therelative role of nuclear energy in the national energy supply
and the extent to which nuclear technology was developed
domestically or imported. The estimate for the worldwide
level of exposure was based on the trends in the data fromthe reporting countries. In the last three periods, the numberof monitored workers decreased by about 25%, from 120,000
in 1990–1994 to about 90,000 in 2000–2002. The annual
collective effective dose dropped by a factor of 4 over the six
periods, from 170 man Sv in 1975–1979 to 36 man Sv in
2000–2002. This fall is a consequence of the reduction in the

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effective dose, which fell from 1.4 to 0.4 mSv from the first
period to the last. The trends are presented in table 70 andfigure XLV.
(h)

Waste

579. Radioactive waste arises at all stages of the nuclear fuelcycle in the process of producing electricity from nuclear
material. The cycle comprises the mining and milling of theuranium ore, its processing and fabrication into nuclear fuel,its use in the reactor, the treatment of the spent fuel taken fromthe reactor after use and finally the disposal of the waste.
Radioactive waste is classified as low-level, intermediate-level
and high-level waste: (a) low-level waste is generated by hospitals, laboratories and industry, as well as the nuclear fuelcycle. It comprises paper, rags, tools, clothing, filters, etc.,
which contain small amounts of mostly short-lived radioactive
material. In order to reduce its volume, it is often compacted
or incinerated (in a closed container) before disposal. Worldwide it makes up 90% of the volume but only 1% of the radioactivity of all radioactive waste; (b) intermediate-level waste
contains higher amounts of radioactive material and may
require special shielding. It typically includes resins, chemicalsludges and reactor components, as well as contaminatedmaterials from reactor decommissioning. Worldwide it makes
up 7% of the volume and 4% of the radioactivity of all radioactive waste; (c) high-level waste may be the spent fuel itself
or the principal waste from its reprocessing. While making up
only 3% of the volume of all radioactive waste, it contains
95% of the radioactive material. It includes the highly radioactive fission products and some heavy elements with long-lived
radioactivity. It generates a considerable amount of heat andrequires cooling, as well as special shielding during handlingand transport. If the spent fuel is reprocessed, the separated
waste is vitrified by incorporating it into borosilicate (Pyrex)
glass which is sealed inside stainless steel canisters for eventual disposal deep underground. On the other hand, if spent
reactor fuel is not reprocessed, all the highly radioactive isotopes remain in it, and so the whole fuel assemblies are treatedas high-level waste. This spent fuel takes up about nine times
the volume of the vitrified high-level waste that would result
from reprocessing and encapsulating an equivalent amount ofspent fuel, which is then ready for disposal.
580. The doses of the personnel managing radioactive
waste depend on the scope of the activities performed. The
average annual effective dose for the workers involved in thesafe management of spent fuel is in the range 0.2–11 mSv.
The level of exposure is lower for the workers involved inwaste management (disposal facilities), where the averageannual effective dose is in the range 0.2–3 mSv [I38, T4].
Some of these data are presented in table 71.
(i)

Summary

581. The trends in worldwide occupational exposures arising from each stage of the commercial nuclear fuel cycle aresummarized in table 72 and are illustrated in figures XLVIand XLVII. The data are annual averages over five-year periods, except for the last period, which covers only three years.
During the first three periods, the number of monitored
workers in the commercial nuclear fuel cycle rose, fromabout 560,000 to 880,000, but in 1990–1994 it started to fall.
The figures for the last three periods are 800,000, 700,000
and 660,000 (figure XLVII). This decrease was largely due
to the drastic reduction in the estimated number in the mining sector, from 260,000 to 12,000. For the three last periods, this may be an underestimate, owing to the limitationsof the data set, but all other indicators support a significant
reduction in this component of exposure of the monitored
workforce. In the first five-year period, mining accounted forover 40% of the workforce, but over the subsequent periods,
reactor operation became the dominant sector with respect tothe number of monitored workers, and at 440,000 it now
accounts for about 70% of the total.
582. The average collective effective dose, averaged
over five-year periods, initially increased from 2,300 to
3,000 man Sv but in the last four periods decreased to2,500, 1,400, 1,000 and 800 man Sv (figure XLVII).
583. The average annual effective dose received by monitored workers in the fuel cycle has fallen progressively overthe course of the six periods: the values are 4.4, 3.7, 2.6, 1.8,
1.4 and 1.0 mSv. There is considerable variation in these
averages for the different stages of the fuel cycle, but overall
the downward trend is evident in all nuclear fuel cycle stages.
The fraction of monitored workers receiving annual doses inexcess of 15 mSv (NR15) averaged over five-year periods has
decreased from about 0.20 to about 0.02; the corresponding
decrease in the fraction of the collective effective dose (SR15)
has been from about 0.63 to about 0.06.
2.

Medical uses of radiation

584. Radiation is used in medicine for both diagnostic andtherapeutic purposes. Irrespective of the level of health caresystem, medical uses of radiation increase yearly as thebene­fits of procedures become more widely disseminated.
The medical use of ionizing radiation remains a rapidlychanging field, stimulated in part by the high level of innovation by equipment supply companies. The wide range ofapplications and of procedures and techniques employed in
the context of patient exposure are described in annex A,“Medical radiation exposures”, which also discusses changes
in practice and current trends. Consideration here is limited
to the occupational exposures that arise from the applicationof these medical procedures. The physicians, technicians,
nurses and others involved constitute the largest single groupof workers occupationally exposed to man-made sources ofradiation. Occupational doses received by staff can be differentiated according to the source of exposure, thereby
characterizing different occupational groups. There is a needto have an evaluation of these occupational doses according
to the main procedures that cause them.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

585. There exists a group of individuals who support or
comfort patients undergoing radiation treatment or diagnostic procedures. Individuals in this group are generally considered as members of the public, though they can also be
workers. The doses for this group of “comforters” are notconsidered in this annex.
586. The Committee has evaluated occupational exposurefor each practice using average values for all workers over
five-year periods, without having taken into account the
influences of job function and medical procedure on staff
exposure. One of the purposes of this annex is to provide
significant information on occupational exposure related tothe different practices. This will be done by: identifying thejob functions and categories of work within each practicethat give rise to the more significant exposures; evaluating
the contribution of external and internal exposure to the total
effective dose; and indicating extremity doses (equivalentdoses in hands and lens of the eye).
587. Data from the UNSCEAR Global Survey of Occupational Radiation Exposures for workers involved in all medical uses of radiation are presented in table A-24. The
estimation of the worldwide level of occupational exposure
was based on the trends from the data reported by countries.
588. The doses used in this annex are those provided by thecountries. For this analysis it was assumed that the dose had
been estimated taking into account the design of any leadapron used, its thickness and the position of the dosimeter, inparticular whether the dosimeter was worn outside the apron
or under it, or whether one dosimeter was worn under theapron and a second worn outside it [N11].
(a) Diagnostic radiology
589. Diagnostic examinations with X-rays have been usedin medicine for over a century, although with increasing
sophistication and new techniques. Medical imaging has
experienced a technological revolution resulting in the
improved imaging of anatomy, physiology and metabolism[H21]. Steady advances in the quality of X-ray images and inpatient protection have ensured a continuing role for diagnostic X-rays in health care, even though alternative modalities
for some diagnoses (such as ultrasound and endoscopy) arebecoming increasingly available, particularly in developed
countries. An increasingly wide range of equipment and techniques are employed to meet a diversity of diagnostic clinicalpurposes. Variations in occupational doses among six identified subgroups in diagnostic radiology (computed tomo­
graphy technologists, general radiographers, fluoroscopy
technologists, radiologists, nurses and radiologic technology
interns) were evaluated. More than 80% of computed tomo­
graphy (CT) technologists and general radiographers do nothave measurable exposure [A8]. On the other hand, the average individual effective dose for interventional procedures issignificantly higher than for conventional diagnostic radio­
logy. Medical doctors performing interventional procedures301

are the most exposed occupational group in diagnostic radiation [K1]. On the basis of such findings, it is no longer appropriate, for example, to treat doses from diagnostic radiology
and from interventional procedures together, as the dose­average ­disguises significant differences between them.
(i)Conventional diagnostic radiology

590. Conventional X-ray examinations involve static imaging; the various techniques applied (radiography, CT, mammography and bone mineral densitometry) are described inannex A, “Medical radiation exposures”. For radiography,
which is the most widely used X-ray application, the average
doses depend on the equipment used. For CT, occupationaldoses are very low and the technique does not represent asignificant source of occupational exposure. For mammo­
graphy, the doses are generally similar to those in CT. In
general, the techniques used in conventional radiography do
not represent a significant source of occupational exposure.
591. During radiography with fixed installations, the radio­
grapher would normally stand in a control booth that typically is shielded as a secondary barrier against X-ray tube
leakage and scattered radiation from the room and thepatient. Depending on room size and barrier thickness, thedose to a radiographer in the control booth area is typicallyless than 1 µSv for a single film taken with a technique of
80 kVp and 40 mA s. Mobile units, however, operate in an
unshielded environment and are therefore of greater concern[N10].
592. Occupational exposure arising from the use of CT isusually low, because the primary X-ray beam is highly collimated, and scattered radiation levels are low. In all such CT
units, leakage of radiation has been reduced to near zero. For
staff in the control room of a properly designed facility, CT
does not represent a significant source of exposure.
(ii)

Interventional procedures

593. The past four decades have witnessed immense technological advances in radiology. The introduction of image
intensification led to the development of interventional
radio­logy. Dotter and Judkins described the first percutaneous treatment of arteriosclerotic vascular obliterations in
1964 [D13], and the range of interventional procedures has
dramatically increased since then. This has been accompanied by considerable equipment development. Because of
the great advantages of interventional radiology, it is not surprising that both the number and the variety of interventional
procedures have grown significantly over the years. Fluoroscopic guidance is frequently utilized in performing many
interventional techniques, including precision diagnostic
and therapeutic injection procedures. In 29 European countries, the number of coronary angiographies (CA) and PTCA
(percutaneous transluminal coronary angioplasty) procedures increased by 264% and 416%, respectively, between

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1992 and 2001 [T5, V20]. It is estimated that approximately
1–4 million interventional procedures are performed annually in the United States, with at least 50% performed underfluoroscopy [M5].
594. The development of CT equipment has made possible a variety of clinical applications. One example is CT
fluoroscopy, which allows the observation of real-time CT
images [K4]. In addition, biopsy examinations performedby CT fluoroscopy offer better accuracy and easier manipulation than conventional examinations. Although the beam
for CT is narrow, the tube voltage and current are relatively
high. Exposures are unavoidable for medical staff whocarry out the examinations. Various surface doses for operating and assisting physicians are shown in table 73. Since
examination times differed greatly from case to case, thedoses were averaged per minute of fluoroscopy for each setof ­fluoroscopy conditions [N15].
595. These procedures require the surgeon and assisting personnel to remain close to the patient and thus close to the primary beam of radiation. The advent of complex and prolonged
coronary interventional procedures has further increased levels of radiation exposure, although proper procedures andexperience can decrease exposures per case. The use of digital
imaging, although potentially capable of reducing radiationdose, requires additional constraints on input dosing, fluoro­
scopy and personnel exposure. Data on the occupational exposure of paediatric cardiologists are sparse, but suggest thattechnical limitations in working with the smaller patient may
adversely affect radiation exposure to physicians and the­medical personnel who assist them [K3, L10].
596. Occupational exposures of primary medical doctors
involved in interventional procedures vary considerablyaccording to the procedure used. Doses were assessed for
different parts of the body according to the type of procedure; results are shown in table 74. The doses resulting fromthe coronariography procedure are about three times higher
than those for angioplasty, arteriography and valvuloplasty.
The average time of fluoroscopy can be lower that for theother procedures, but the number of frames is higher [S21].
597. A total of 1,000 consecutive patients with chronic
pain undergoing interventional procedures performed by one
physician were studied. Two fluoroscopy units were utilized
and operated by two certified radiological technologists. The
procedures performed included caudal and interlaminar epidural injections, facet joint nerve blocks, percutaneous adhesiolysis, intercostal nerve blocks, sympathetic blocks,
transforaminal epidural injections and other procedures. The
results showed that these 1,000 patients had undergone
1,729 procedures with an average duration of radiation exposure of 13.2 ± 0.33 seconds per patient and 7.7 ± 0.21 seconds per procedure. Dosimetry measurements indicated aneffective dose of 13 mSv (1,345 mrem) outside the leadapron; the measurement inside the apron was below thedetection limit. The levels of exposure were significantly
below the annual limits recommended [M6].

598. The radiation exposures to three vascular surgeons
performing 47 consecutive endovascular aortoiliac aneurysm (EAIA) procedures were determined over a one-year
period. The total fluoroscopy time was 30.9 hours (mean of
39.4 minutes per case). The time spent using high-level
fluoroscopy varied between 5% and 37%, although in one
case it was 60%. The current ranged from 2.1 to 4.7 mA andthe tube potential from 65 to 105 kV. Annual effective doses
for the primary surgeon, first assistant and second assistant,respectively, were: 1.5, 1.6 and 0.9 mSv under the leadapron, and 13.8, 12.6 and 5.2 mSv outside it. The estimatedequivalent doses to the eyes were 7.8, 5.7 and 2.0 mSv forthe primary surgeon, first assistant and second assistant,respectively. The estimated equivalent doses to the hands
were 18.7, 16.0 and 5.4 mSv for the primary surgeon, firstassistant and ­second assistant, respectively [L16].
599. Occupational doses in interventional cardiology andradiology services at a university hospital are shown intable 75 for the period 1999–2001. According to the data
presented in this table, the doses are about the same for all
staff if protective measures are undertaken [V9].
600. An evaluation of the occupational exposure to personnel performing cardiac catheterization (dose per procedure) has shown that the range of effective doses fordiagnostic catheterizations was 0.02–38 µSv, for percutaneous coronary interventions 0.17–31.2 µSv, for ablations
0.24–9.6 µSv and for pacemaker or intracardic defibrillator
implantations 0.29–17.4 µSv. The authors have estimated a
reduction in dose of a factor of 4 from 1971 to 2006 for thestaff involved in diagnostic catheterizations. For percutaneous coronary interventions, an increasing pattern wasobserved over time, but this was not statistically significant.
The higher radiation exposure for percutaneous coronary
interventions was primarily due to the long fluoroscopy
times. The contribution of fluoroscopy to the total dose wasabout 30% for diagnostic catheterizations and 60% for­percutaneous coronary ­interventions [K13].
601. An evaluation of the typical occupational dose levels
in interventional radiology and cardiology installations covered a sample of 83 procedures performed by ten specialists
in six laboratories [V7]. The monitored staff wore nine thermoluminescent chips sited next to the eyes and on the forehead, neck, hands, left shoulder, left forearm and left arm
during each individual procedure. Doses for interventional
radiologists and cardiologists are presented in tables 76 and77, respectively. Radiologists were occasionally out of theroom controlling the image acquisition from the system console, and had the lowest dose values. For cardiologists, doses
were divided into values measured with and without the useof a protective lead screen. The lowest values correspond tostaff who made regular use of the protective screen. The meandose values for interventional radiologists and interventional
cardiologists are presented in figure XLVIII. A more homogeneous distribution was observed for vascular radiologists
than for interventional cardiologists. This is because of thevariable positions usually adopted by a radiologist with

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

respect to the patient, while for interventional cardiology procedures, doses are mainly received on the left side, which is
closest to the scatter volume throughout the procedure [V7].
The main difference between radiology and cardiology isthat, for the former, the X-ray tube is less likely to be rotatedand moved during the procedure, and the position of the operator is more variable, depending on the procedure [R5]. Thereis exposure to scattered radiation not only in the upper part ofthe operator’s body, but to the lower part as well, even though
this exposure is at low levels [B33, M5, S21].
602. Different investigators have observed substantial differences in doses received for the same type of procedure.
These differences may be as large as an order of magnitude.
Many factors influence occupational radiation exposuresduring fluoroscopy use. No single standardized method has
evolved to permit easy comparison of dosimetry results
among studies [K13, P1, T11, T12]. The doses vary considerably according to the procedure, operator training andquality assurance. As shown in table 78, occupational doses
related to interventional procedures are strongly dependent
on several parameters, including: dose rate gradients in thevicinity of the patient; the technique selected (kV, mA or
mA s per pulse); the different filtrations available in modern
equipment; field size; TV monitor; intensifier size; operational modes (continuous or pulsed fluoroscopy); the numberof frames; dose rates; locations inside the room; the typicalinterventions performed; patient weight and size; design andmaintenance of the facility; and the existence and use of protective tools, especially spectacles and ceiling-suspended
screens [C1, K3, M10, P12, S37, V8, W15, Z6].
603. National data on occupational exposures arising from
X-ray diagnostic radiology over the six periods are presented in the first part of table A-24. Most of the countries
have not distinguished between data from conventional andinterventional procedures. There is a wide variation in theeffective dose and the percentage of measurably exposed
workers. This variation may be explained by many factors,
including the way data are recorded in the national database, the variety of procedures performed by the medical
staff and the protective measures implemented by eachcountry. The worldwide level of occupational exposure isevaluated on the basis of the analysis of trends for the countries. However, the number of workers is derived from the
UNSCEAR Global Survey of Medical Radiation Usage andExposures. The data are presented in table 79 and figure XLIX. They indicate that overall there has been an
increase in the number of monitored workers employed inthis practice, from 630,000 in 1975–1979 to 6.7 million in
2000–2002, although the number dropped considerably
between the third and fourth periods, from 1,350,000 in
1985–1989 to 950,000 in 1990–1994. For the last two periods the estimated value is a factor of 7 higher than for theprevious period. The number of workers involved in interventional procedures represents about 0.1% of the total
workforce employed in diagnostic radiology. The latest
value is more reliable than the previous ones. The collective
effective dose increased from 600 to 760 man Sv over the303

first three periods, dropped by about 40% in the fourth
period (1990–1994) and then increased by a factor of 7 to
3,370 man Sv in 2000–2002, following the same pattern as
the number of workers. The average effective dose decreased
from 0.94 mSv in 1975–1979 to 0.50 mSv in 1990–1994,
and has since remained constant.
604. Only a few countries have provided data that distinguish between the workers engaged in conventional techniques and in interventional procedures; the data arepresented in table A-25. On the basis of the reported data,
about 85% of the monitored workers are involved in conventional radiology techniques. For conventional techniques,
the average annual effective dose is about 0.5 mSv (range
0.02–1.24 mSv) for the monitored workers and 1.2 mSv(range 0.33–3.14 mSv) for the measurably exposed workers.
For interventional procedures, the average annual effective
dose is about 1.6 mSv for the monitored workers and 3.1 mSv
for the measurably exposed workers, with a range of 0.4–
29.5 mSv. Considerable variation exists in the effective dose
and the percentage of measurably exposed workers. Thisvariation may be explained by differences between countries. Greece has provided data for various job categories forinterventional radiology: 1) medical doctors—cardiologists;
2) medical doctors—orthopaedists, surgeons, gastroentero­logists; auxiliary staff of similar categories; 3) nurses; and4) others (table A-25 and figure L). It can be seen that cardio­
logists are the group most exposed. Their average effective
dose is 4 mSv, about a factor of 6 higher than the doses forother medical doctor specialists, nurses and other workers.
These data are presented separately to indicate that, althoughthese workers are small in number compared with those
involved in the conventional diagnostic use of X-rays, their
exposure levels are high. As the interventional techniques
are being widely applied, the Committee expects that there
will be an increasing trend in the number of workers exposedand consequently in the annual collective effective dose.
605. During interventional procedures, the hands of the
medical doctors are in the field of radiation, resulting in highexposure to the hands and arms. According to the literature,
the equivalent dose to the skin can considerably exceed
500 mSv. During cardiac and abdominal intravascular angiography, a surgeon may receive annual doses to the hands of
440 mSv and 360 mSv, respectively [J1, S35]. There is alarge range in the doses reported in the literature. Only a few
countries have reported data on extremity doses. The reported
values are low, the highest value being around 10 mSv.
(b)

Dental practice606. Diagnostic X-ray machines are widely available andare used frequently in almost every dental office or clinic.
The total number of X-ray devices used in dentistry is thus
extremely large. Their range of energy is between 20 and
60 keV. Occupational exposure in dentistry is due to scattered radiation from the patient and leakage from the tube
head (although the latter should be insignificant with modern

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equipment). The general trend over the last 30 or more years
has been a dramatic increase in the number of personnel
involved in dental radiology coupled with a steady decreasein collective dose. A majority of dental practitioners do notreceive measurable doses, and indeed some regulatory
authorities do not require routine individual monitoring
except where workload is high [U3].
607. National data on occupational exposures arising fromdental practice over the period 1995–2002 are given in thesecond part of table A-24. The worldwide level of occupational exposure is evaluated on the basis of the analysis oftrends for the countries. The data are presented in table 80
and figure LI; they indicate a progressive decline in thenumber of monitored workers involved in this practice. The
value first increased from 370,000 to 500,000 in 1980–1984,
decreased to 265,000 in 1990–1994 and increased again to404,000 in 2000–2002. The annual collective effective dose
also fell, from 120 man Sv in 1975–1979 to 24 man Sv in
2000–2002. The average annual effective dose decreased
from 0.32 mSv in 1975–1979 to 0.06 mSv in 1990–1994,
and then remained steady. The percentage of measurably
exposed workers has been about the same over the last three
periods at around 5%. About 1% of the workforce received
doses higher than 1 mSv; there were no recorded doses
higher than 5 mSv. Following the trends from the six periods, the predicted number of monitored workers for 2007
would be about 350,000, the collective effective dose would
be about 14 man Sv and the average effective dose would be
about 0.05 mSv.
(c)

Nuclear medicine

608. A broad aim in nuclear medicine is the investigationof physiological processes, with most procedures involving
some form of measurement to quantify organ function. The
use of radionuclide generators, particularly 99mTc generators,
requires handling tens of gigabecquerels of radioactive material during the elution process. The magnitude of exposureswhile performing clinical nuclear medicine procedures
depends on the precautions taken, including the use of
syringe shields when administering injections. Personnel
must be close to the patient when giving injections and while
positioning the patient and the camera. Usually the imaging
process makes the largest contribution to the exposure ofstaff [B8]. Internal exposures of personnel are usually much
lower than external exposures and are controlled by monitoring work surfaces and airborne concentrations, althoughsome medical centres also conduct routine bioassays [N10].
609. Radionuclides used for organ imaging emit penetrating gamma radiation and give rise to exposure of nuclear
medicine staff. While some therapeutic procedures are carried out, nuclear medicine departments can generally becharacterized by various diagnostic examinations involving
intravenous administration of radiopharmaceuticals. Databases on occupational exposure in nuclear medicine rarely
distinguish between diagnostic and therapeutic applications.

In nuclear medicine, because of the possibility of internal
exposure, higher values of annual effective dose are expected
for personnel involved in the preparation and assay of­radiopharmaceuticals than for medical doctors and nurses.
610. Radionuclides used for organ imaging, for example
Tc, emit penetrating gamma radiation and cause exposureof nuclear medicine staff and other persons in the vicinity ofpatients undergoing diagnosis or treatment. The dose rate at
1 m from a typical diagnostic patient is about 10 µSv/h after
the administration of 0.74 GBq of 99mTc. Therapeutic administrations, for example 3.7 GBq of 131I, give rise to a dose rateof about 200 µSv/h at 1 m from the patient, who therefore
will normally be segregated to reduce the exposure of otherpersons. Work involving the preparation and assay of radio­
pharmaceuticals is associated with the highest occupational
exposures in this field and can give annual doses of up to
about 5 mSv. Doses to hands and fingers can range up to theannual limit of 500 mSv. Various shielding devices can beused to reduce extremity doses. However, the majority ofworkers in nuclear medicine departments who are notdirectly handling radiopharmaceuticals receive very low
exposures, typically less than 1 mSv in a year [N10].
99m

611. An evaluation of effective doses received by the staff
involved in tasks with 131I in nuclear medicine has shown thatthe average annual values range from 0.35 mSv to 3.27 mSv;
the maximum dose was around 9 mSv. The evaluation was
performed using measurements of 131I in the thyroid [K18].
612. Positron emission tomography (PET) is considered
one of the most important diagnostic imaging techniques,
having the unique ability to provide functional and quantitative information on the target organ of interest. During recent
years, great efforts have been made to improve the diagnostic accuracy of this imaging modality through the development of new acquisition/processing systems and theintroduction of new β+-emitting radiopharmaceuticals, all ofwhich has increased the interest of clinicians [Z5].
613. Occupational exposure can be higher by a factor of
2–4 for technologists than for physicians involved in PET
procedures. External exposures over a period of one year tofour workers (two physicians and two technologists) working full time at a PET centre are presented in table 81. The
annual doses ranged from 4.6 to 8 mSv for technologists andwere about 2 mSv for physicians. In this centre, an ECATEXACT HR+ state-of-the-art scanner is used, and 18F-labelled
fluorodeoxyglucose (18F FDG) whole-body imaging represents the principal clinical activity (about 96% of the patient
workload) [Z5].
614. Table 82 shows the doses for various tasks and patient
conditions and for different technologists. Each patient wasadministered 555 MBq of 18F FDG. The doses varied frombelow the detection limit for the monitoring technique to
6.8 µSv per procedure. The tasks that result in the largest
exposures of the technologist were patient positioning, injection of the dosage and measurement, in that order [M17].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

615. Assessments of internal dose for workers from somePET centres in Germany have shown that internal exposureis rather low. Of 79 workers, only 13% received measurable
annual doses, which ranged from 0.05 to 1.5 mSv [E7]. Similar values of effective dose were found in an evaluation ofoccupational exposure at a PET/CT installation in Spain.
The doses were 2 mSv for the workers involved in PET/CT;
this represents 100 times more than the doses received byworkers operating conventional nuclear medicine imaging
equipment (0.02 mSv) [C27].
616. The radiation exposure of PET technologists can bequite high and has a large variation. Annual doses have beenreported variously as 3, 10 and 12 mSv [B28, R16, Z1]. The
annual doses for PET technologists are higher than those fortechnologists performing general nuclear medicine studies,with values averaging about 3 mSv and 2 mSv, respectively.
The estimated average dose per PET procedure was 4.1 µSv(11 nSv/MBq) [R24]. An evaluation of occupational doses
received by the technologists in a PET centre has shown thatthe average daily effective dose was about 14.4 µSv. On
average, each technologist administered 831 MBq daily. The
mean whole-body dose per MBq injected was 0.02 µSv/
MBq. The average daily amount of time at close distances(less than 2 m) from a radioactive source was 32 minutes.
The average effective dose per minute of close contact was
0.5 µSv [B20].

305

solutions and the handling of waste, local skin doses to thehands of the personnel due to beta emitters can reach very
high values. For example, the preparation and applicationof liquid-filled balloon catheters for vascular brachytherapy resulted in a measured daily equivalent dose at thefingertips of a nuclear medicine specialist that could considerably exceed the recommended annual limit for skin,which is 500 mSv. In other radiosynoviorthesis procedures, it was also estimated that the annual skin dose limit
was exceeded owing to direct radiation from beta emitters.
In the case of unsealed sources, additional exposures arelikely because of possible skin contamination [R14]. These
therapeutic applications are in fact a ­combination of
nuclear medicine and radiotherapy.

618. Comparison of occupational exposure due to the useof 18F FDG with exposure due to other radiopharmaceuticals
used in conventional nuclear medicine procedures (such as
gallium scan, bone scan and sestamibi cardiac scans) has
shown that PET, high-dose 67Ga and high-dose 201Tl do not
represent a significantly greater occupational radiation hazard than conventional nuclear medicine procedures; thesedata are shown in table 83 [W11, Z5].

620. Radiation synovectomy or radiosynoviorthesis (RSO)
is a new nuclear medicine procedure in rheumatology andorthopaedics that uses beta emitters. With this treatment it ispossible to efficiently treat local chronic inflammatory joint
diseases. In radiosynoviorthesis, radioactive colloidal solutions (169Er, 186Re or 90Y) are injected into inflammatory
joints. Investigations of exposure to medical staff were performed in ten hospitals and doctors’ surgeries. Very highlocal skin doses were measured both for assistants who prepared the syringes and for physicians who injected the radio­
pharmaceutical solutions. The local equivalent doses to theskin were up to 100 mSv per working day for assistants andup to 200 mSv per working day for the physicians due to
direct radiation [B10]. The very high doses for both assistants and physicians resulted largely from holding the upper
end of the cannula between thumb and index finger whileconnecting or separating the cannula and syringe, or while
injecting the radiopharmaceutical solutions into the joint.
The highest exposure occurred using 90Y solutions, owing to
the high beta energy of 90Y. The use of 169Er causes little
direct radiation exposure, but for 186Re, exposure is not negligible. The mean specific dose (local skin dose related toapplied activity) was about 60 µSv/MBq (range 13–233 µSv/
MBq) for 90Y and about 20 µSv/MBq (4–40 µSv/MBq) whenusing 186Re. In some cases, considerable skin contaminationof personnel occurred. Because of the high specific activitiesof the solutions (for example about 500 MBq/mL of 90Y),
very small, invisible contamination spots may cause highlocal skin doses [R14].

619. Within the field of therapeutic applications in
nuclear medicine, new agents with beta emitters are being
increasingly used. In contrast to most gamma emitters, theenergy of beta rays can be totally absorbed in a smalldelimited tissue volume; thus exposure can be limited tothe tissue to be treated. The higher effectiveness of beta
radiation is reflected in the higher values of beta ray dosecoefficients compared with gamma ray dose coefficients.
This leads to dose rates for beta emitters that are two ordersof magnitude greater than those of gamma emitters forequal activities and short distances [R14]. The increasing
number of medical procedures requires proper attention to
extremity doses received by radiopharmacy staff members
involved in nuclear medicine. During the preparation of621. Table 84 shows the maximum daily skin doses forradio­synoviorthesis (RSO) procedures in seven different
institutions. Although RSO is a well-established andapproved method, it can be carried out differently withrespect to details that may considerably influence the radiation exposure of the personnel. In the case of right-handed
persons, usually the index finger, thumb and middle fingerof the left hand were most exposed, because of the mannerof holding the vials or syringes. Doses to the right handwere often lower by an order of magnitude. Lack of know­
ledge on the part of medical staff of the high exposuresfrom beta radiation can lead to inadequate radiation protection measures being applied in the use of beta-emitting
radionuclides during RSO [B9].

617. The assessment of occupational doses to staff working within the imaging section of a PET/cyclotron service inAustralia has shown that PET involves higher radiation
exposure to staff than do other types of imaging. The average dose per patient for a technologist is calculated at
1.25 µSv. Staff attending sick patients also have increased
exposure [C28].

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622. Radioimmunotherapy following the Zevalin™ procedure (90Y ibritumomab tiuxetan) involves administering the
drug for the treatment of patients with relapsed and refractory
non-Hodgkin’s lymphomas. Zevalin™ consists of an antiCD20 monoclonal antibody linked to the radioisotope 90Y;
the monoclonal antibody component allows radioimmunotherapy to be targeted only towards malignant cells expressing CD20 antigen. The beta radiation of 90Y kills the target
cells and other malignant cells in the surrounding area. Immediately before administration, the antibody component mustbe radiolabelled with 90Y on site; activities of about 2 GBq
are handled in this procedure. As a result, medical personnel
receive significant partial-body doses if adequate radiation
protection measures are not followed. In the Zevalin™ protocol, the preparation of the solution is the critical step before
injection. It is essential that a syringe shield and adaptedshielding be used, since otherwise the dose to the hands canbe extremely high. With radiological protection measures
implemented, the maximum local skin dose to the fingertipsof the nuclear medicine specialist amounted to 25 mSv per
treatment during radiolabelling and administration of thetherapeutic dose of Zevalin™ to the patient [A21, R14].
623. At the Academic Hospital of the Free University ofBrussels, hand doses have been monitored for several years
by means of wrist dosimeters and ring dosimeters (TLDs).
Both types are convenient to wear but do not necessarily represent the location on the hand where the highest skin dose isreceived. The number of manipulations, amounts of activity
handled and results from routine monitoring have highlighted the need for more detailed dosimetry for radiopharmacy workers. In this study, two radiopharmacists weremonitored during more than 300 manipulations at 18 different locations on each hand. The results expressed in dose perunit activity handled during a specific manipulation showed
good reproducibility for individual radiopharmacists. Typical values of Hp(0.07) ranged from 50 to 600 µSv/GBq ofhandled activity; the fingertips received the highest dose.
Particular personal habits in handling radiopharmaceuticals
determined the location and the magnitude of skin doses,
especially in manipulation of radiopharmaceuticals withhigh exposure rates such as 18F FDG. The results from thisstudy have shown that annual skin doses would reach about
400 mSv. The principal radiopharmaceuticals that contributed to extremity doses at the hospital were 99mTc (85%) and18
F (10%), with 5% contributed by other radioisotopes (123I,
201
Tl, 51Cr, 67Ga, 131I and others) [B21].
624. National data on occupational exposures arising from
nuclear medicine over the six periods are given in the thirdpart of table A-24. The worldwide level of occupational exposure is evaluated on the basis of the analysis of trends for thecountries. The data are presented in table 85 and figure LII.
They indicate a progressive increase in the number of monitored workers involved in this practice, from 61,000 in 1975–
1979 to 120,000 in 2000–2002. The collective effective dose
increased from 62 man Sv in 1975–1979 to about 90 man Sv
in 1990–1994, and then decreased slightly to 87 man Sv in
2000–2002. The average effective dose decreased from

1.0 mSv in 1975–1979 to 0.7 mSv in 2000–2002. The percentage of measurably exposed workers has been about thesame over the three last periods at around 56%. About 30% of
the workforce received doses higher than 1 mSv; there wereno recorded doses higher than 15 mSv. Following the trends
from the six periods, the predicted number of monitored
workers for 2007 would be 124,000, the collective effective
dose would be 88 man Sv and the average effective dosewould be 0.7 mSv. It is important to point out that the recorded
doses are related to external exposure. The doses related tonuclear medicine can be underestimated, since there is some
contribution from internal exposure, although it is small­compared with external exposure.
625. There is a wide variation in the effective dose and thepercentage of measurably exposed workers. This variation
may be explained by many factors, including the way data
are recorded in the national database, the mixing of doses
related to exposed workers and non-exposed workers in the
database, the mixing of doses from the various procedures
performed by the medical staff and the protective measures
implemented by each country.
626. In order to obtain improved information about occupational exposures of nuclear medicine workers according to
specialty, new questionnaires covering the period 1995–2002
were distributed to Member States requesting information ondose distributions for medical doctors, ward nurses and technicians. Few countries provided information according toworker specialty. A large variation is evident with respect tothe estimated average annual effective dose and collective
dose; the data are presented in table A-26. The doses for technicians can be substantially higher than the doses for the otherstaff (medical doctors and nurses). This is to be expected,
since the technicians are responsible for the preparation of
the injected solutions. However, the doses are dependent on
the number of manipulations, the types of radioisotope andthe amount of activity handled. Greece has provided data
separately for four different job categories: technicians, medical doctors, nurses and others, as illustrated in figure LIII.
This shows that the levels of exposure for technicians andmedical doctors were higher than for nurses and others. However, the values of average effective doses for the measurably
exposed workers were not statistically different.
627. During the preparation of solutions and handling ofwaste, local skin doses to the hands of the personnel due to beta
emitters can reach high values. According to the literature, theequivalent dose at the fingertips of a nuclear medicine specialist can considerably exceed 500 mSv for skin [A21, B9, B10,
I59, R14]. There is a large range in the doses reported in theliterature. Few countries have reported data on extremity doses,
but the values reported are low—around 5 mSv.
(d)

Radiotherapy

628. Therapeutic uses of ionizing radiation are quite different
in purpose from diagnostic radiology procedures. Radiotherapy

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

is an important treatment modality for malignant disease (seeannex A, “Medical radiation exposures”). In radiotherapy, there
are three main treatment categories where occupational exposure may occur: external beam treatment, brachytherapy andtherapy simulation. Brachytherapy, where there is manual loading of the radioactive sources, is usually the most significant
source of personnel exposure [N10]. Exposures may occur during the receipt and preparation of the sources, during loading
and unloading, and during treatment.
629. Personnel are not normally present in the treatment
room when external beam therapy is performed, with thepossible exception of low-energy (50 kVp and less) X-ray
contact therapy units, which are sometimes used for intra­
cavitary treatments. Some exposures can, however, be caused
by 60Co teletherapy units as a result of leakage while thesource is in the off position and by radiation that penetrates
the barrier during use [N10].
(i) Teletherapy
630. The exposures from linear accelerators, betatrons andmicrotrons depend on the type of beam (photon or electron)and the beam energy. Below 10 MeV, exposure results only
from radiation that penetrates the protective barrier. Above
10 MeV, photonuclear reactions can produce neutrons andactivation products. The neutrons can penetrate the protective barrier while the unit is operating. Residual activity can
expose personnel who enter the treatment room immediately
after the treatment has been delivered. The exposures, however, are usually low. Exposures from simulators and otherdiagnostic imaging equipment used to plan treatments arealso usually low [N10].
631. Radiation therapy staff in the treatment rooms of medical accelerators operating with energies of above about
10 MeV are also exposed to radiation due to activated materials. The activation arises primarily from photonuclear reactions and neutron capture. Published estimates of the annual
activation dose received by staff during typical operations
are in the range 0.7–5 mSv. These numbers demonstrate thatthe activation dose is not negligible and suggest that, at leastin conservatively shielded facilities, the therapist receives agreater occupational dose due to activation than to radiationtransmitted through the shielding barriers [A12, R4].
632. Intensity-modulated radiation therapy (IMRT) may
play a dominant role in oncology practice in the near future.
However, IMRT techniques require a substantial increase inaccelerator beam-on time compared with conventional radiation therapy to deliver the same patient dose. This could
lead to an increased dose being received by radiation therapists. The increased beam-on time influences radiation
exposure in two ways. First, the dose outside the treatment
room due to leakage (including neutrons and capture gammarays) transmitted through secondary barriers will increase,though in principle this can be compensated for by increasing the barrier thickness [M13]. Secondly, in situations

307

where IMRT is delivered using high-energy radiation, thedose inside the treatment room due to induced activation is
also expected to increase. The activation dose rates in atreatment room were evaluated for 28Al, 56Mn, 24Na andlong-lived isotopes generated at 18 MeV, for different treatment regimes. The largest contribution to doses came from28
Al and 56Mn [R4]. It is worthy of note that the two principal isotopes, 28Al and 56Mn, were also observed to be thedominant isotopes responsible for activation in the treatment room of an accelerator operating at 16 MeV, and interestingly, in the treatment room of a fast neutron facility[Y4]. The isotope 24Na is commonly found in concrete thathas been activated by thermal ­neutrons [N12].
(ii)

Brachytherapy

633. Brachytherapy involves the placement of radioactive
sources within the body or on its surface so that the radiation
source is close to the tissue to be treated. This enables a highdose of radiation to be delivered to malignant tissue andlower doses to normal tissue.
634. Intracavitary brachytherapy is used for the treatmentof gynaecological cancers. This involves the placement ofradioactive sources into the uterus. Sources can be manuallyplaced into the uterus in a surgical theatre; however, thisapproach results in the theatre staff, porters and ward nursing
staff receiving a high radiation dose. Afterloading was introduced as a means of reducing the radiation dose to staff; inthis technique the radioactive sources are remotely placed
into position by a treatment machine.
635. A technique of permanent implantation of radioactive
seeds into the prostate so that their decay will deliver theprescribed dose to the tumour is in common use. The isotopes used are predominantly 125I and 103Pd. Procedures forseed implantation vary, but there are generally two stages.
The first is the manual preloading of needles, which can be
performed either by composing loose seeds and spacers or
by cutting off strands of seeds and reabsorbable spacers.
This process can be done according to a previously approved
plan or on the basis of cumulative experience concerning thenumber of needles and the loading usually needed. The second stage consists of the implantation of these preloaded
needles in the operating room. This procedure results in low
occupational exposures because of the low activity used perseed and the low energy emitted by 125I. Table 86 shows themean dose rate levels per implant that best exemplify thedose received by different staff involved during the average
40 min phase of insertion of the seeds into the prostate using
the afterloading technique [G4]. The only step in whichseeds are not properly shielded is during their movement
through the delivery tube, but this process is performed soquickly that it is generally accepted that sufficient protection
is provided by stepping back a minimum distance of 50 cm
from the tube during the process. Assuming a maximum
train of five seeds of maximum activity 40 MBq, the dose
rate at 50 cm would be no greater than 0.01 mSv/h [S13].

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636. National data on occupational exposures arising fromradiotherapy over the six periods are given in the fourth partof table A-24. There is a wide variation in the effective doseand percentage of measurably exposed workers. The worldwide level of occupational exposure is evaluated on the basisof the analysis of trends for the countries. The data are presented in table 87 and figure LIV; they indicate a progressive
increase in the number of monitored workers involved in thispractice, from 84,000 in 1975–1979 to 127,000 in 2000–
2002. The collective effective dose decreased from
190 man Sv in 1975–1979 to 60 man Sv in 2000–2002. The
average effective dose decreased from 2.2 mSv in 1975–1979
to 0.5 mSv in 2000–2002. The percentage of measurably
exposed workers has been about the same over the last three
periods, at around 40%. About 10% of the workforce receives
doses higher than 1 mSv; there are no recorded doses higher
than 15 mSv. Following the trends from the six periods, thepredicted number of monitored workers for 2007 would be
about 130,000, the collective effective dose would be about
57 man Sv and the average effective dose would be about
0.5 mSv.
(e)

All other medical uses

637. The category “all other medical uses of radiation” wasintended to cover new and/or expanding uses of radiation
within the medical sector that did not fit into the categories
of diagnostic radiology, dental radiology, nuclear medicine
or radiotherapy. The principal example has been biomedical
research. Educational establishments use radioactive sources,
X-ray equipment and unsealed radioactive sources for a wide
range of activities. Examples of uses include X-ray crystallography, radioactive labelling (for example using 3H, 14C,
32
P, 35S and 125I) and irradiators using 60Co or 137Cs sealed
sources [U22]. The UNSCEAR 1993 Report [U6] noted thatthe lack of consistency in reporting data made it difficult toestimate the level of exposure or to draw useful comparisons
for this category of exposure. On the basis of the reported
data it is possible to conclude that some countries may notrecord the data separately according to the techniques usedin the medical field. In this case, the doses are reported in“all medical uses”.
638. The number of workers potentially exposed in theseother uses may substantially exceed those in the few occupations for which data have been separately presented in thissection. The average exposure levels of workers involved inother uses of radiation are in general low. However, the way
in which the doses are aggregated may disguise somewhat
higher average doses in particular occupations. The only
way to ascertain the existence of occupations, or subgroups
within occupations, that receive doses significantly above
the average is for the data to be inspected periodically.
639. National data for the various categories were aggregated by country to give data on exposures to workers arising from all medical uses of radiation; they are presented intable A-27. There is a wide variation in the effective doseand percentage of measurably exposed workers. The analysis of trends for the countries indicates a drastic decrease in
the number of monitored workers involved in this practicefrom 1990–1994 to 1995–1999, and a slight increase ofabout 20% in the last period. The collective effective dose
follows the same pattern as the number of monitored workers. The average effective dose has tended to be the sameover the last periods. It is difficult to project with any­accuracy the level of exposure for 2007.
(f)

Summary

640. National data on occupational exposures arising fromall medical uses of radiation averaged over five-year periods
are given in table A-27. The Committee has decided to estimate the number of workers on the basis of the UNSCEAR
Global Survey of Medical Radiation Usage and Exposures.
The average effective dose is estimated on the basis of thedata presented in table A-24.
641. The evaluation of trends in occupational exposure for
20 European countries in the medical sector has shown aslight decrease in the average level of exposure. The averagecollective dose also decreased slightly, from 177 to
171 man Sv, while the average effective dose did not change,
having remained at around 1 mSv from 1996 to 2000 [F15].
642. There is a wide variation in the effective dose and percentage of measurably exposed workers. This variation may
be explained by many factors, including, the way data arerecorded in the national database, the mixing of doses related
to exposed workers and non-exposed workers in the database, the mixing of doses from the various procedures performed by the medical staff and the protective measures
implemented by each country.
643. For X-ray diagnostics there is a trend of increasing
numbers of workers, increasing collective effective doses
and relatively constant values for the average effective dose.
The estimated number of workers is around 6.74 million,
which represents about 90% of the total number of monitored workers involved in the medical uses of radiation. The
estimated average collective dose is around 3,370 man Sv,
which represents about 95% of the total collective dose forall medical uses. Following the trends from the six periods,
the predicted level of occupational exposure for X-ray diagnostic radiology for 2007 would show an increase of 10% in
the number of workers and in the average collective dose,and no change in the effective dose. The effective dose has
been relatively constant over the last three periods (from1990–1994 to 2000–2002); this may be due to the influenceof the high doses related to interventional procedures. On thebasis of the reported data that distinguish between doses
from conventional and from interventional procedures, about
0.1% of the monitored workers in diagnostic radiology are
involved in interventional procedures. The average annual
effective dose due to conventional techniques is about
0.5 mSv (range 0.02–1.24 mSv) for the monitored workers

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

and 1.2 mSv (range 0.33–3.14 mSv) for the measurably
exposed workers. The average annual effective dose forworkers involved in interventional procedures is about
1.6 mSv for the monitored workers and 3.1 mSv for themeasurably exposed workers, with a range of 0.4 to 29.5 mSv.
The doses to the hand can exceed 500 mSv. Although thereported values vary considerably, they are relatively low,
and the highest reported value is around 10 mSv.
644. For dental practice there is a trend of decreasing numbers of workers over the six periods (although an increase ofabout 50% has been observed in the last periods), withdecreasing collective effective doses and decreasing average
effective doses. The estimated number of workers is about
0.40 million, which represents about 5% of the total numberof monitored workers involved in the medical uses of radiation. The estimated average collective dose is 24 man Sv,
which represents about 0.7% of the total collective dose forall medical uses. The estimated average effective dose is
0.06 mSv. Following the trends from the six periods, the predicted level of occupational exposure for dental practice for2007 would show a 10% increase in the number of workers,
a decrease of about 5% in the average collective dose and a
3% decrease in the average effective dose.
645. For nuclear medicine there is a trend of increasing
numbers of workers, decreasing collective effective doses
and decreasing average effective doses. The estimatednumber of monitored workers is 0.12 million, which represents about 5% of the total number of monitored workers
involved in the medical uses of radiation. The estimatedaverage collective dose is 87 man Sv, which represents about
10% of the total collective dose for all medical uses. The
estimated average effective dose is 0.7 mSv. Following thetrends from the six periods, the predicted level of occupational exposure for nuclear medicine for 2007 would show a
3% increase in the number of workers, a 1% decrease in the
average collective dose and a 4% decrease in the average
effective dose. These dose projections may be underestimates, since new technologies have been introduced, and theuse of new radiopharmaceuticals and 18F FDG has increased
considerably. During the preparation of solutions and thehandling of waste, local skin doses to the hands of the personnel due to beta emitters can reach very high values,
exceeding 500 mSv.
646. On the basis of the data from countries that havereported the doses for the different job categories separately,
the Committee concludes that the doses for technicians canbe substantially higher than the doses for the other staff(medical doctors and nurses). This is expected, since thetechnicians are responsible for the preparation of the injected
solutions. However, the doses are dependent on the numberof manipulations, the types of radioisotope and the amountof activity handled.
647. For radiotherapy there is a trend of increasing numbers of workers, decreasing collective effective doses anddecreasing average effective doses. The estimated number of309

workers is 0.13 million, which represents about 5% of thetotal number of monitored workers involved in the medical
uses of radiation. The estimated average collective dose is
60 man Sv, which represents about 7% of the total collective
dose for all medical uses. The estimated average effective
dose is 0.5 mSv. Following the trends from the six periods,
the predicted level of occupational exposure for radiotherapy
for 2007 would show an increase of about 3% in the numberof workers, and a decrease of about 5% for both the averagecollective dose and the average effective dose.
648. For the category “all other medical uses of radiation”,
which covers new and/or expanding uses of radiation within
the medical sector that do not fit into the categories of diagnostic radiology, dental radiology, nuclear medicine or radio­
therapy, the analysis is based on data from only one country
that has reported the doses for the five practices within the
medical uses. The analysis of trends for this country indicates a drastic decrease in the number of monitored workers
involved from 1990–1994 to 1995–1999, and an increase ofabout 20% in the last period. The collective effective dose
follows the same pattern as the number of monitored workers. The average effective dose has tended to remain thesame over the last several years. It is difficult to project any
level of exposure for 2007. These workers represent about
30% of the total number of monitored workers in the practices related to medical uses of radiation. The average collective dose represents about 25% of the total collective dose
for all medical uses.
649. The estimated number of workers involved in the
medical uses of radiation is 7.40 million, the collective effective dose is 3,540 man Sv and the average effective dose is
0.5 mSv. The evaluation of the trends in occupational exposure for all medical uses together shows an increasing
number of monitored workers, and decreasing collective
effective dose and average effective dose, as shown intable 88 and figure LV. The largest contribution to the occupational exposure is from diagnostic radiology. The numberof monitored workers has increased over the six periods,
dominated by those involved in diagnostic radiology.
3.

Industrial uses of radiation

650. Radiation sources, including sealed sources, X-ray
machines and particle accelerators, are used in a number ofindustrial applications. Among these are: industrial irradiation; non-destructive testing (particularly industrial radio­
graphy); well logging; luminizing; thickness, moisture,density and level gauging; tracer techniques; and fluoroscopic and crystallographic analysis of materials. Becauseof the many different occupations involved and the ways inwhich exposures are categorized, it is difficult to obtaincomparable statistics in different countries. Most exposuresin industrial uses of radiation are low, a fact that contributes
to the lack of detail in recorded data. In the UNSCEAR
1993 Report [U6], exposures were considered for thosegroups of workers that generally experience higher doses:

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industrial radiographers, luminizers and well loggers. Workers involved in isotope production and workers employed
and monitored at education and research institutes were also
assessed. The following categories were introduced in thesurvey of data for 1995–1999 and 2000–2002: industrial
irradiation, industrial radiography, luminizing, radioisotope
production, well logging, accelerator operation and all otherindustrial uses. For the first three periods, the exposure ofworkers in educational establishments and tertiary education was included within the general category of industrial
uses; since the UNSCEAR 2000 Report [U3], these exposures have been included within a “miscellaneous” category
in section II.C.4.
651. National data on occupational exposures arising fromthe industrial use of radiation for the categories mentioned
above are given in table A-28. National data for the various
categories were aggregated by country to give data on exposures to workers from all industrial uses of radiation; they
are presented in table A-29. The Committee has decided not
to follow the procedures of the previous UNSCEAR reports
to estimate the worldwide level of exposure. The decision
was based on the lack of sufficient information to calculate areliable figure that would reflect the worldwide level of
exposure for the last two periods (1995–1999 and 2000–
2002). On this basis, it was decided to evaluate the trends forrepresentative countries.
(a)Industrial irradiation

652. The most widespread uses of industrial irradiation
are the sterilization of medical and pharmaceutical products, the preservation of foodstuffs, polymer synthesis andmodification, and the eradication of insect infestation. The
product doses required are extremely high, and the source
activities or beam currents are correspondingly high. For
gamma facilities the source would typically be 60Co in thepetabecquerel range; some 137Cs sources are also used. Dose
rates in the irradiation chamber would be of the order of
1 Gy/s, and in some cases there is a need to protect against
radiogenic heating that could cause fires. Gamma and electron irradiation facilities must be constructed such that during normal use any radiation exposure of workers will bevery low.
653. This category of work was first specified in the previous UNSCEAR Global Survey of Occupational Radiation
Exposures [U3]. The available data over the six periods are
given in the first part of table A-28; these data are limited andcover only 13 countries. Of crucial importance is the factthat there are very few data from the large industrializedcountries, where the greatest number of irradiators arelocated. It is difficult to evaluate the trend of exposure forlack of information.
654. For this annex, data from China are analysed to show
trends over the past periods. Figure LVI indicates anincrease in the number of monitored workers, from 100 in

1990–1994 to 1,400 in 2000–2002. The collective effective
dose increased from 0.10 man Sv in 1990–1994 to
1.22 man Sv in 1995–1999, and then dropped to 0.88 man Sv
in 2000–2002. The average effective dose consistently
decreased, from 1.03 mSv in 1990–1994 to 0.63 mSv in
2000–2002. The percentages of measurably exposed workers fell from 90% in 1990–1994 to 63% in 2000–2002(table A-28). According to the dose distribution data, about
29% of the workers received doses higher than 1 mSv, and
1% of them received doses higher than 15 mSv. As seen in
the first part of table A-28, the other countries follow thesame pattern of occupational exposure as that described for
China, with a decrease in the collective dose and average
effective dose. It is difficult to project a level of exposurefor 2007 on the basis of the ­available data.
(b)

Industrial radiography655. Industrial radiography is a non-destructive practicefor examining materials for defects. Gamma radiation from137
Cs and 60Co sources as well as X-rays are used to examine welded metal joints. This technique can be applied inthree basic formats. The oldest format is direct manual
manipulation, either using handling equipment or with thesource as an integral part of a shielded “torch”. This format, which was prevalent in the 1970s but was alreadydeclining in the 1980s, is still used to some extent. Anotherformat has the source in a shielded container; the source
can be rotated or moved to produce a collimated beam.
This format, too, is being used less frequently. By far thelargest amount of gamma radiography is carried out using
remote exposure containers. Typically the source is on theend of a drive cable that can be controlled from about ten
metres away, so that the source is projected down a flexible
tube to the radiography position, where a collimator is normally positioned to reduce the radiation dose to the operators. These devices are portable and are widely used for site
radiography. They are also used in fixed-facility radiography, where they can be integrated into the installed safetysystems, although this is not always done. The X-ray sets inindustrial radiography typically vary in applied voltage
from 60 to 300 kV, although there are some 400 kV units.
In addition, there are a smaller number of linear accelerators, typically in the range 1–8 MV. These are mostly in
fixed facilities with installed safety systems, but there are a
few mobile units.
656. Industrial radiography is performed in two quite different situations. In the first, it is carried out at a single location, usually in a permanent facility that has been designed
and shielded for the purpose; in this case, items to be radiographed are brought to the facility. In the second situation,
the radiography is conducted at multiple locations in thefield, in which case the radiographic equipment is brought tothe location where the radiograph is required, this procedureoften being referred to as “site radiography”. There are usually significant differences in the degree of control that canbe exercised in the two situations.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

657. The available data over the six periods are given in thesecond part of table A-28. The worldwide estimate of thelevel of occupational exposure was based on an analysis of
the trends in all countries that have provided information;
the data are shown in figure LVII. The number of monitored
workers increased from 72,000 in 1975–1979 to 116,000 in
1980–1984 and then remained about the same in the last four
periods, being 113,500 in 2002. The collective effective dose
followed the same pattern as the number of workers, increasing from 190 to 230 man Sv in the first two periods, thendropping to 170 man Sv and remaining constant for the subsequent periods. The average effective dose dropped from
2.6 to 2.0 mSv for the first two periods and remained about
the same for the subsequent periods at 1.5 mSv. The percentages of measurably exposed workers dropped from 50% in
1990–1994 to 44% in the last period (2000–2002). According to the dose distribution data, about 30% of the monitored
workers received doses higher than 1 mSv, and 1% of themreceived doses higher than 15 mSv.
658. Following the trends from the six periods, the predicted level of occupational exposure for industrial radiography for 2007 would show an increase of 4% in the number ofworkers and the average collective dose, and a slight decrease,of 2%, in the effective dose.
659. The different levels of occupational exposure for­multiple-location and single-location industrial radiography
have been demonstrated by the data from the United Statespresented in table 89. About 90% of the workforce is engaged
in multiple-location industrial radiography. The average collective effective dose for workers involved in single-location
work is less than 1% of that for multiple locations. The average effective dose for workers involved in single-location
work is about 7% of that for multiple locations [U29, U30,
U31, U32, U33, U34, U36, U37, U38].
(c)

Luminizing

660. Luminizing is one of the oldest industrial uses of ionizing radiation. In the past, alpha or beta emitters were mixed
with a phosphor, such as zinc sulphide, and then painted ondials, such as watch faces or airplane instrumentation.
Present-day practice includes using luminizing compounds
in gunsights and as low-level light sources for exit signs andmap illuminators.
661. The data for the six periods are given in the third partof table A-28. Only three countries have reported data for theperiods 1995–1999 and 2000–2002. Switzerland has reported
data for most of the periods, allowing the Committee to analyse the trend in occupational exposure over the years. Figure LVIII indicates that the number of monitored workers
varied over the six periods, dropping from 210 in 1975–1979
to 130 in the second period, increasing to 350 in 1995–1999
and dropping to 220 in 2000–2002. The collective effective
dose consistently decreased, from 2.31 man Sv in 1975–1979
to 0.18 man Sv in 2000–2002. This was due to a decreasing

311

average effective dose, which fell considerably, from
11.2 mSv in 1975–1979 to 0.80 mSv in 2000–2002. The percentages of measurably exposed workers were evaluated forthe last two periods, being 98% in 1995–1999 and 93% in
2000–2002. According to the dose distribution data, 22% of
the monitored workers received doses higher than 1 mSv,
2% received doses higher than 10 mSv and 1% received
doses higher than 15 mSv. Historically the doses to workers
involved in luminizing were high, but in recent years there
has been a significant reduction. It is difficult to project thelevel of occupational exposure for 2007 on the basis of only
a few countries.
662. On the basis of all the reported data, the average
effective doses have decreased over time: 7.44 mSv (1975–1979), 5.01 mSv (1980–1984), 2.71 mSv (1985–1989),
0.38 mSv (1990–1994), increasing to 1.93 mSv in 1995–
1999 and dropping to 0.72 mSv in 2000–2002. Except forthe period 1995–1999, the trend is a progressive decrease ofthe ­effective dose.
(d)Radioisotope production

663. Radioisotopes are produced for a great variety ofindustrial and medical purposes. The main source of occupational exposure in radioisotope production and distribution
is external irradiation; internal exposure may be significant
in some cases. In general, however, internal exposures havenot been included in reported statistics for occupational
exposure except in more recent years, and even then their
inclusion is far from universal. Reporting conventions forworkers involved in radioisotope production may also varyfrom country to country (for example with respect to whether
the reported doses include only those arising during the initial production and distribution of radioisotopes or whether
they also include those arising in the subsequent processing,encapsulation, packaging and distribution of radionuclides
that may have been purchased in bulk from elsewhere), andthis may affect the validity of comparisons between reported
doses.
664. Among the radioisotopes produced, 131I is the one
most likely to contribute a significant dose due to internal
exposure. However, 131I has gradually been supplanted byother radionuclides with shorter half-lives. In the past, many
countries did not record internal exposures. Control of
intakes was accomplished mainly through area monitoring,and little emphasis was given to the use of bioassays, mainly
because of the cost and the difficulty of interpreting theresults. A retrospective study of 131I-contaminated workers in
the radiopharmaceutical industry in Brazil has shown that,even with continuously reinforced implementation of safety
principles and good practice in handling iodine, committed
effective doses can reach values around 4 mSv/a, since thevolatility of iodine makes its compounds readily available
for intake by inhalation [G1]. The results of internal exposure monitoring on four individuals who worked for seven toten years in a 131I radiopharmaceutical production laboratory

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UNSCEAR 2008 REPORT: VOLUME I

in the Islamic Republic of Iran showed the maximum andminimum annual intakes to be 536 and 79 kBq, respectively,
although one worker (involved in an incident) had an annual
intake estimated at 3.8 MBq [A10].
665. The number of cyclotrons dedicated to the productionof positron-emitting radionuclides is increasing in medical
institutions/hospitals owing to the well-established role of
PET imaging in clinical practice. The radiation safety issues
in a cyclotron PET facility are much different from those inconventional nuclear medicine facilities because of the­presence in a cyclotron PET facility of penetrating gamma
photons of 511 keV, higher specific gamma ray constant of
positron emitters and the secondary neutrons from the­cyclotron during production. Therefore work practices in a
­cyclotron-PET facility need to be more stringent. The radiation dose to workers in a cyclotron and radiochemistry laboratory measured over 12 months is shown in table 90. The
dose received by workers in a cyclotron facility was less than
5% of the annual dose limit (20 mSv) to the whole body andless than 2% of the annual dose limit (500 mSv) to theextremities. Similarly, the doses received by workers in theradiochemistry laboratory were less than 10% of the annual
limit to the whole body and less than 1% of the annual limit
to the extremities. This was to be expected, as all the operations in this cyclotron and radiochemistry laboratory arecompletely automated and have adequate shielding [P3].
The evaluation of the occupational exposure in a cyclotron
facility in which the total annual activity of 18F and 13N produced was 31 TBq in 2002 (synthesis of 18F FDG represented
90% of the total activity) showed an average annual effective
dose of about 7 mSv and an extremity dose of 36 mSv [P8].
666. The occupational doses were found to be 5–10
times less than the regulatory limits in the cyclotron vault,
8–30 times less than the regulatory limits in the radiochemistry laboratory and 10–200 times less than the regulatory limits outside the cyclotron laboratory during beam
operation. Internal doses were found to be negligible in
the facility [S18].
667. National data on occupational exposures arising fromradioisotope production over the period 1995–2002 are given
in table A-28. The worldwide level of occupational exposure
has been evaluated on the basis of the analysis of individual
country trends. The data are presented in figure LIX. They
indicate a progressive increase in the number of monitored
workers over the first three periods, from 57,000 to 88,000,
followed by a drop to 24,000 in 1990–1994, after which thenumber began to increase by about 4% for each period,
resulting in 34,560 in 2000–2002. The collective effective
dose fell from 130 man Sv in 1975–1979 to 47 man Sv in
1990–1994, and then increased to 62 man Sv in 2000–2002.
The average effective dose decreased from 2.25 mSv in
1975–1979 to 1.12 mSv in 1985–1989, and then increased toabout 2 mSv in the last three periods. The percentage ofmeasurably exposed workers has been around 50–70%.
About 55% of the workforce received doses higher than
1 mSv, and about 2% received doses higher than 15 mSv.

Following the trends from the six periods, the predicted
number of monitored workers for 2007 would be 41,472, thecollective effective dose would be 75 man Sv and the ­average
effective dose would be 1.8 mSv.
(e)

Well logging

668. Well logging is the practice of using radioactive
sources or miniature X-ray machines to measure geologicalcharacteristics (such as porosity, density and elemental composition) in boreholes drilled for mineral, oil or gas exploration. Well logging has been identified in some countries as
an industrial use of radiation that can lead to higher doses toworkers than other industrial uses. This is sometimes attributed to the manual manipulation of sources in small spaces,
for example on oil rigs. Both gamma and neutron sources areused in well logging, but the contributions from each to thereported doses are generally not indicated.
669. The data on well logging are presented in the fifth partof table A-28. In this practice it is difficult to draw a trend in
the level of occupational exposure. Only 12 countries havereported data to UNSCEAR on occupational exposure. Canada is more relevant in terms of the number of monitored
workers, representing about 60% of the total reported workforce, and has reported data throughout all the periods. The
Canadian data are used to illustrate the trend in occupational
exposure for this practice. Figure LX indicates a significant
increase in the number of monitored workers in the second
period, from 450 in 1975–1979 to 1,010 in 1980–1984. The
number was then approximately constant from 1985 to 1999:
1,110 (1985–1989), 950 (1990–1994), 1,060 (1995–1999).
It increased in the last period to 1,430. The collective effective dose increased from 0.52 man Sv to 1.37 man Sv in
1985–1989 and dropped to 0.71 man Sv in 2000–2002. The
average effective dose was about 1.2 mSv in the first three
periods and decreased to 0.50 mSv in 2000–2002. The percentages of measurably exposed workers are between 40%
and 70%. It is difficult to project levels of exposure for theyear 2007, but they would be expected to reflect increasing
numbers of workers and declining average collective doseand average effective dose.
670. On the basis of all the reported data, the average effective doses have decreased over time: 1.32 mSv (1975–1979),
1.17 mSv (1980–1984), 1.07 mSv (1985–1989), 0.36 mSv(1990–1994), increasing to 0.92 mSv in 1995–1999 and to
0.96 mSv in 2000–2002. Except for the period 1990–1994,
the trend is a progressive decrease of the effective dose.
(f)Accelerator operation

671. Consideration is limited here to occupational exposures arising from accelerators used for nuclear physics
research at universities and at national and international laboratories. Accelerators (generally of somewhat smaller size)
are increasingly being used for medical purposes, i.e.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

therapy and radiopharmaceutical purposes; however, theexposures arising from those uses are more appropriately
associated with exposures arising from the medical uses ofradiation. Similarly, accelerators are also found in radiography and commercial radioisotope production, but again theseare dealt with under those work categories. Most exposuresresulting from accelerators arise from induced radioactivity
and occur mainly during the repair, maintenance and modification of equipment. They result mainly from gamma radiation from the activation of solid surrounding materials by
penetrating radiation. The potential for internal exposure in
the normal operation of accelerators is slight, and doses via
this route are negligible in comparison with those due toexternal irradiation.
672. Early high-energy accelerators used internal targets to
produce either radioisotopes or secondary beams of normallyunstable particles. Very high levels of activation products
were produced in the region of the targets, and before 1960,
typical annual collective doses per accelerator were
1–2 man Sv. Such doses still apply for many of the early
cyclotrons that remain in operation. Between 1960 and 1980,
beam extraction techniques were improved, which led toreduced levels of activation products. However, these reductions were largely offset by the continuing increases in beam
power.
673. In the 1980s, two developments had an important
influence on occupational exposures at accelerators. The firstwas the increasing importance of colliding beam techniques
for the production of events of interest to the particle physics
community. Average beam intensities, as measured by the
number of particles accelerated per day, are several orders ofmagnitude lower than those used in fixed-target physics
experiments. Consequently the production of activation
products has been greatly reduced, and this is reflected in the
exposures of maintenance personnel. The second development was a move towards heavy-ion operation, where again
the accelerated beam intensities are several orders of magnitude lower than those with proton acceleration. This has also
led to a decrease in activation products and consequently inthe exposures during maintenance.
674. The available data are shown in the sixth part oftable A-28. In the first three periods, from 1975 to 1989, thereported data were dominated by those of the United States.
Since 1990, Canada has contributed the majority of thenumber of monitored workers. Data from Canada are used toillustrate the trend in occupational exposure for this practice.
Figure LXI indicates a significant increase in the number ofmonitored workers in the first three periods: 580 (1975–1979), 880 (1980–1984) and 1,000 (1985–1989). The
number then decreased slightly to 888 in 2000–2002. The
collective effective dose increased from 0.17 man Sv in
1975–1979 to 1.06 man Sv in 1985–1989, and then decreased
to 0.44 man Sv in 2000–2002. The average effective dose
increased from 0.30 mSv in 1975–1979 to 1.1 mSv in 1985–
1989, and then decreased to 0.5 mSv in 2000–2002. The percentages of measurably exposed workers are between 26%313

and 50%, with the highest number in the period 1985–1989.
The decrease in the collective dose over the last two periods
is influenced by the decrease in average effective dose.
According to the dose distribution data, 12% of the monitored workers received doses higher than 1 mSv and 2% ofthem received doses higher than 5 mSv.
675. On the basis of all the reported data, the average effective doses decreased in the second period and then kept constant over time: 1.62 mSv (1975–1979), 0.76 mSv(1980–1984), 0.62 mSv (1985–1989), 0.75 mSv (1990–
1994), 0.62 mSv (1995–1999) and 0.74 mSv (2000–2002).
676. Following the trends from the six periods, the predicted level of occupational exposure for accelerator operation for 2007 would show an increase of about 3% in thenumber of workers and a decrease of about 10% in the
average collective dose and the average effective dose.
(g)

All other industrial uses

677. There are many other uses of radiation in industry, forexample in soil moisture gauges, thickness gauges and X-ray
diffraction, but occupational exposure data for these are in
general not separately identified or reported. This categoryof practice has been incorporated into the UNSCEAR 2000
Report [U3] to accommodate the data from all the otherpractices not mentioned under the industrial uses of radiation. The number of workers potentially exposed in theseother uses may substantially exceed the number in the few
occupations for which data have been separately presented
in this section. The average exposure levels of workers
involved in “other uses of radiation” are in general low.
However, the way in which the doses are aggregated may
disguise somewhat higher average doses in particular occupations. The only way to ascertain the existence of occupations, or subgroups within occupations, that receive doses
significantly above the average is for the data to be inspectedperiodically.
678. The available data are shown in the last part oftable A-28. Japan, Germany and France represent about 87%of the reported monitored workers. It certainly is the casethat the national systems for collecting such data do notallow the data to be readily separated into the categories usedin this review. Although the Netherlands represents just 2%of the total number of reported monitored workers, it is used
to illustrate the trend in occupational exposure for this practice, since it has reported data for most of the practices
related to industrial uses of radiation. The number of monitored workers has decreased from 2,880 in 1990–1994 to
2,180 in 2000–2002. The collective effective dose decreased
from 0.22 man Sv in 1990–1994 to 0.15 man Sv in 2000–
2002. The average effective dose remained steady at
0.08 mSv in 1990–1994 and 0.07 mSv in 2000–2002. The
percentages of measurably exposed workers were about
20%. According to the dose distribution data, about 1% of
the monitored workers received doses higher than 1 mSv.

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679. For the great majority of the reported data, the average
effective dose was very low, less than 1 mSv. On the basis of
the reported data, the average effective dose was 0.45 mSv in
1990–1994, 0.27 mSv in 1995–1999 and 0.26 mSv in
2000–2002.
680. Following the trends from the six periods, the projected level of occupational exposure for all other industrial
uses for 2007 would show an increase of about 2% in thenumber of workers, and a decrease of about 10% in the­average collective dose and the average effective dose.
(h)Summary681. Table A-29 shows the national data from all industrial
uses of radiation grouped together. The data are more complete than for the separate categories of industrial uses ofradiation, but as with the data for medical uses they sufferfrom incomplete data for the United States, which is important in the estimation of worldwide exposure. The Committee has decided not to follow the procedures of the previous
UNSCEAR reports to estimate the worldwide level of exposure. The decision was based on the lack of sufficient information to calculate a reliable figure that would reflect theworldwide level of exposure. The estimate of the worldwide
level of occupational exposure for all industrial uses was
based on the trends for all countries and all practices.
682. The trends in the worldwide level of exposure over thesix periods are presented in figure LXII. The total number ofmonitored workers involved in the practices related to theindustrial uses of radiation increased by a factor of 1.6, from
530,000 in 1975–1979 to 870,000 in 2000–2002; it is dominated by industrial radiography. The collective effective dose
decreased by a factor of 3, from 870 man Sv in 1975–1979
to 348 man Sv in 2000–2002. The average effective dose
decreased by a factor of 4, from 1.6 in 1975–1979 to 0.4 mSv
in 2000–2002.
683. The evaluation of the trend of occupational exposurein 21 European countries in general industries has shown anslight decrease in the level of exposure. The average effective dose decreased from 2.0 to 1.8 mSv and the averagecollective dose decreased from 76 to 69 man Sv in the period
1996–2000 [F15].
684. There is a wide variation in the effective dose and thepercentage of measurably exposed workers. This variation
may be explained by many factors, including the way data
are recorded in the national database, the mixing of doses
related to exposed workers and non-exposed workers in the
database, and the protective measures implemented by eachcountry.
685. For industrial irradiation there is a trend of increasing
numbers of workers and decreasing collective effective doseand average effective dose. Industrial irradiation represents
about 3% of the total number of monitored workers in thepractices related to industrial uses of radiation. The averagecollective dose represents about 3% of the total collective
dose for all industrial uses. There is not sufficient information to have a statistically significant prediction for 2007.
686. The trend of occupational exposure in industrial radio­
graphy is an increase in the number of workers and the average collective effective dose and a decrease in the average
effective dose. Industrial radiography represents about 20%of the total number of monitored workers in the practices
related to industrial uses of radiation. The average collective
dose represents about 55% of the total collective dose for allindustrial uses. Following the trends from the six periods,
the predicted level of exposure for 2007 would show anincrease of about 4% in the number of workers and the average collective dose and a decrease in the effective dose ofabout 3%.
687. For luminizing there is a trend of an increasing numberof workers and a decrease in the collective effective dose andthe average effective dose. Luminizing represents about
0.3% of the total number of monitored workers in the practices related to industrial uses of radiation. The average collective dose represents about 1% of the total collective dose
for all industrial uses. There is not sufficient information tohave a statistically significant prediction for 2007.
688. For radioisotope production there is a trend of anincreasing number of workers and a decrease in the averagecollective effective dose and the average effective dose.
Radioisotope production represents about 3% of the total
number of monitored workers in the practices related toindustrial uses of radiation. The average collective dose represents about 10% of the total collective dose for all industrial uses. Following the trends from the six periods, thepredicted level of exposure for the year 2007 would show anincrease of about 20% in both the number of workers and theaverage collective dose, and no change in the average­effective dose.
689. For well logging there is a trend of a slightly increasing number of workers, and a decrease in the average collective effective dose and the average effective dose. Well
logging represents about 0.4% of the total number of monitored workers in the practices related to industrial uses ofradiation. The average collective dose represents about 1% of
the total collective dose for all industrial uses. Following thetrends from the six periods, the predicted level of exposurefor 2007 would show an increase of about 4% in the numberof workers, a decrease of about 5% in the average collective
dose and a decrease of 10% in the average ­effective dose.
690. For accelerator operation there is a trend of a slightly
increasing number of workers, and a decrease of the averagecollective effective dose and the average effective dose.
Accelerator operation represents about 0.3% of the total
number of monitored workers in the practices related toindustrial uses of radiation. The average collective dose represents about 1% of the total collective dose for all industrial

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

uses. Following the trends from the six periods, the predictedlevel of exposure for 2007 would show an increase of about
3% in the number of workers and a decrease of about 10% in
the average collective dose and the average effective dose.
691. For all other industrial uses there is a trend of aslightly increasing number of workers and a decrease in the
average collective effective dose and the average effective
dose. All other industrial uses represent about 73% of thetotal number of monitored workers in the practices related
to industrial uses of radiation. The average collective dose
represents about 29% of the total collective dose for allindustrial uses. Following the trends from the six periods,
the predicted level of exposure for 2007 would show anincrease of about 2% in the number of workers and adecrease of about 10% in the average collective dose andthe average effective dose.
692. In summary the number of monitored workers has
increased over the six periods. The average annual effective
doses to monitored workers involved in industrial uses ofradiation have consistently decreased over the six periods.
The greatest contribution to the occupational exposure comesfrom industrial radiography.

(universities, polytechnics and research institutes with a major
educational role). Exposures resulting from research related to
the nuclear fuel cycle and from such activities as the use ofaccelerators should have been included in those more specificoccupational categories.
695. The data reported by countries are given in the firstpart of table A-30. The worldwide level of occupational
exposure is evaluated on the basis of the analysis of trends
for the countries. The data are presented in figure LXIII; they
indicate a progressive increase in the number of monitored
workers, from 140,000 in 1975–1979 to 446,000 in 2000–
2002. The collective effective dose decreased from 74 man Sv
in 1975–1979 to 38 man Sv in 2000–2002. The average
effective dose decreased from 0.55 mSv in 1975–1979 to
0.09 mSv in 2000–2002. The percentage of measurably
exposed workers has been about 10%. About 2% of theworkforce received doses higher than 1 mSv; there are norecords of doses higher than 5 mSv. Following the trends
from the six periods, the predicted number of monitored
workers for 2007 would be 513,360, the collective effective
dose would be 42 man Sv and the average effective dosewould be 0.08 mSv.
(b)

4.

Veterinary medicine

Miscellaneous uses

693. There remain a number of occupations where radiation exposure may be involved that are not covered by othercategories. These include research in educational establishments, radiology in veterinary medicine, the management ofspent radioactive sources, transport of radioactive material
and others. The data reported by countries are given intable A-30. The Committee has decided not to follow theprocedures of the previous UNSCEAR reports to estimate
the worldwide level of exposure. The decision was based on
the lack of sufficient information for the last two periods(1995–1999 and 2000–2002) to calculate a reliable figurethat would reflect the worldwide level of exposure. On thisbasis, it was decided to evaluate the trend for representative
countries.
(a)

315

Educational establishments

694. Research workers in educational establishments use
radioactive sources, X-ray equipment and unsealed radioactive
sources for a wide range of activities. Examples of uses include
X-ray crystallography, radioactive labels (e.g. 3H, 14C, 32P, 35S,
and 125I) and irradiators using 60Co or 137Cs sealed sources. In
the UNSCEAR 1993 Report [U6], it was noted that the lack ofconsistency in reporting data made it difficult to estimate thelevel of exposure and to draw useful comparisons for this category of exposure. Data that should rightfully be attributed tothis category are often attributed to other broad practices ofradiation, such as research related to the nuclear fuel cycle orindustrial uses, and vice versa. The intent here is to include
exposures arising in tertiary educational establishments

696. Diagnostic radiography is the main source of occupational exposure in veterinary practice. In general, effective
doses to individuals should be low, because they arise essentially from scattered radiation. However, poor practice may
result in the unnecessary exposure of extremities if, forexample, assistants hold animals in position while the radiograph is being taken. The data from the UNSCEAR Global
Survey of Occupational Radiation Exposures are given in
the second part of table A-30. The main contributions of datafor 1995–1999 and 2000–2002 came from Canada, Germany
and the United Kingdom, and to a lesser extent from Denmark and the Netherlands. The United States made the largest contribution of data for the first three periods but has notreported since then.
697. National data on occupational exposures arising fromveterinary medicine are given in the second part oftable A-30. The worldwide level of occupational exposure isevaluated on the basis of the analysis of individual country
trends. The data are presented in figure LXIV. They indicate
a progressive increase in the number of monitored workers
from 48,000 in 1975–1979 to 160,000 in 1985–1989, followed by a drop to 45,000 in 1994–1999 and then an increase
to 119,030 in 2000–2002. The collective effective dose
increased from 25 man Sv in 1975–1979 to 52 man Sv in
1985–1989, decreased in 1990–1994 to 8 man Sv andincreased to 18 man Sv in 2000–2002. The average effective
dose decreased consistently, from 0.52 mSv in 1975–1979
to 0.15 mSv in 2000–2002. The percentage of measurably
exposed workers has been around 30%. About 3% of theworkforce received doses higher than 1 mSv; there are norecorded doses higher than 5 mSv.

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698. For the great majority of the reported data, the average
effective dose was very low over the six periods, less than
1 mSv. The effective doses decreased from 0.73 mSv in
1975–1979 to 0.10 mSv in 2000–2002. The dose distribution
has shown that, for the last period, 12% of the 34,540 workers received doses higher than 1 mSv; there are no recorded
doses higher than 5 mSv.
699. Following the trends from the six periods, the predicted level of occupational exposure for veterinary medicine in 2007 would show an increase of about 10% in thenumber of workers; the average collective effective dosewould not change, and the average effective dose would
decrease by about 8%.
(c)703. An important factor in determining worker doses
appears to be the size and weight of the package. Small, lightpackages, such as those handled by couriers, are usually
touched, handled and carried close to the body. Intermediatesized packages, such as those handled by air cargo handlers,
are usually moved by handcart, conveyor belt or truck. Large
packages, such as those handled in a port or railway yard, orby some truckers, are usually handled only by remote-­
controlled equipment. Thus doses are inversely related to
package size and weight.

Spent sources700. Spent radioactive sources result from industrial applications, research and medicine. A survey was performed in
Turkey on the management of such spent sources (60Co,
137
Cs) at the Waste Processing and Storage Facility, where
11 137Cs sources (total activity 851 GBq) and four 60Co
sources (total activity 27.75 GBq) that had been used as levels and density gauges were conditioned. Reinforced metal
drums (200 L in volume) and cement matrix were used for
conditioning of these sources to achieve greater confinement for long-term storage. The maximum dose rates at thesurface of the conditioned waste packages were 1.60 mSv/h
for 137Cs and 1.63 mSv/h for 60Co. Measurements of the final
waste packages were presented to fulfil the requirements(<2 mSv/h) of transport according to the regulations for thesafe ­transport of radioactive material [O23].
(d)courier companies who physically carried radioactive
sources and the drivers/helpers/sorters of radioisotopes for
medical use. The histograms in figure LXV show the fractions of annual doses in three dose ranges: <1 mSv, 1–5 mSv
and >5 mSv.

Transport

701. Essentially all commercially produced radioisotopes
eventually need to be transported by air, land or sea, depending on the source size and the regulatory control provisions.
In the course of transport, some radiation exposure of thecarriers’ staff occurs. According to the IAEA, the annual
doses due to transport of radioactive material are in the range
0.2–7 mSv, with an average of about 1 mSv [I42]. The highest doses from transport of radioactive material are to drivers/handlers carrying radiopharmaceuticals. The annual
doses to those transporting nuclear fuel is generally low.
702. An evaluation of the occupational doses due to transport was conducted in Canada [E2]. It covered the period
1997–2002 and included 17 companies at 25 sites: a couriercompany, eight trucking companies, a provincial highway
department whose workers transported and used moisturegauges containing radioactive material, a manufacturer, ahospital and a university with shipping/receiving workers,
companies involved in internal transport, air cargo terminals,
a railway and a port. Overall, nearly 90% of the annual dosesin the current study were below 1 mSv. The study participants likely to receive higher doses were the employees of704. A survey of the radiological impact of the normaltransport of radioactive material by air in the United­Kingdom has shown that the highest doses are to handlers
carrying radiopharmaceuticals, including 131I and 201Tl, andtechnetium generators. The average annual effective doses
were given as 1–2 mSv and the maximum annual dose was
3.75 mSv. The annual collective dose for the entire handling workforce in the United Kingdom was about
0.1 man Sv. The doses received by the aircrew were very
low. The average annual effective doses for aircrew inshort- and long-haul passenger flights were 0.003 mSv and0.064 mSv, respectively (the respective collective doses are
0.13 man Sv and 3.8 man Sv). The average annual effective
doses for flight crew of short-haul passenger and cargoflights are 0.0003 mSv and 0.001 mSv, respectively (therespective collective doses are 0.0025 man Sv and0.024 man Sv). The average annual doses for flight crew inlong-haul passenger and cargo flights are 0.007 mSv and0.036 mSv, respectively (the respective collective doses are
0.13 man Sv and 0.56 man Sv) [W3].
705. A recent survey of occupational exposure related totransport in the United Kingdom of material containing
NORM has shown that the maximum annual effective doseto a transport worker would be less than 0.2 mSv [H30].
(e)

Other occupational groups

706. The “other occupational groups” category was includedin the UNSCEAR Global Survey of Occupational Radiation
Exposures to ensure that no sizeable group of exposed persons was overlooked. The data cover disparate groups that
often cut across the other categories. This category was incorporated into the UNSCEAR 2000 Report. The data reported
by countries are given in the third part of table A-30. As China(Taiwan) has reported data for all three periods recorded andrepresents about 20% of the reported number of monitored
workers, it was selected as an example to evaluate the trend in
occupational exposure. There has been a considerable
increase in the number of monitored workers, from 1,990 in
1990–1994 to 3,570 in 2000–2002. The collective effective

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

dose decreased from 1.02 man Sv in 1990–1994 to
0.17 man Sv in 2000–2002. The average effective dose
decreased drastically over three of the periods, from 0.51 mSv
in 1990–1994 to 0.05 mSv in 2000–2002. The percentages ofmeasurably monitored workers decreased from 34% in 1990–
1994 to 6% in 2000–2002. The decreasing collective dose iscorrelated with the decreasing effective dose.
707. The use of X-rays for security purposes has beenreported by two countries in the UNSCEAR survey. Detailed
data of occupational exposure for border policeman and customs personnel have shown that the effective doses range
from 0.3 mSv to 2 mSv.
708. For the great majority of the reported data, the average
effective dose is low over the periods: it decreased from1.03 mSv in 1990–1994 to 0.17 mSv in 2000–2002. The
dose distribution has shown that, for the last period, 4% of
the 21,580 workers received doses higher than 1 mSv; there
are no recorded doses higher than 5 mSv.
709. Following the trends from the six periods, the predicted level of occupational exposure for other occupationalgroups for 2007 would show an increase of about 10% in thenumber of workers, a decrease of about 20% in the averagecollective effective dose and a decrease of 30% in the ­average
effective dose.
(f)Summary710. Table A-30 shows the national data from all other categories of workers not included in the categories of naturalradiation exposures, nuclear fuel cycle, and medical andindustrial uses of radiation. Data that should rightfully be
attributed to the miscellaneous uses of radiation include exposures arising in tertiary educational establishments (universities, polytechnics and research institutes with an importanteducational role), veterinary medicine and all other uses ofradiation involving occupational exposures. Exposures resulting from research related to the nuclear fuel cycle and fromsuch activities as the use of accelerators should have been
included in those more specific occupational categories.
711. There is a wide variation in the effective dose and percentage of measurably exposed workers. This variation may
be explained by many factors, including the way data arerecorded in the national database, the mixing of doses related
to exposed workers and non-exposed workers in the database,and the protective measures implemented by each country.
712. For educational establishments there is a trend ofincreasing numbers of workers, increasing collective effective dose and decreasing average effective dose. Educational
establishments represent about 61% of the total number ofmonitored workers in the “miscellaneous” class. The average collective dose represents about 45% of the total collective dose for all categories classified as miscellaneous.
Following the trends from the six periods, the predicted level317

of occupational exposure for educational establishments for2007 would show an increase of about 15% and 10% in thenumber of workers and the average collective effective dose,
respectively, but a decrease of about 15% in the average
effective dose.
713. For veterinary medicine there is a trend of increasing
numbers of workers, no change in the collective effective
dose and a decreasing average effective dose. Veterinary
medicine represents about 17% of the total number of monitored workers in the “miscellaneous” class. The average collective dose represents about 21% of the total collective dose
for all categories classified as miscellaneous. Following thetrends from the six periods, the predicted level of occupational exposure for veterinary medicine for 2007 would show
an increase of about 10% in the number of workers and in
the average collective effective dose, but no change in the
average effective dose.
714. For all other categories of workers there is a trend ofincreasing numbers of workers and decreasing collective
effective dose and average effective dose. These other categories represent about 22% of the total number of monitored
workers in the “miscellaneous” class. The average collective
dose represents about 34% of the total collective dose for all
categories classified as miscellaneous. Following the trends
from the six periods, the predicted level of occupational
exposure for “other occupational groups” for 2007 would
show an increase of about 10% in the number of workers anda decrease of about 20% and 30% in the average collective
effective and the average effective dose, respectively.
D. Man-made sources for military purposes
715. Radiation exposures of workers in military activities
can be grouped into three broad categories: those arisingfrom the production and testing of nuclear weapons andassociated activities; those arising from the use of nuclear
energy as a source of propulsion for naval vessels; and thosearising from the use of ionizing radiation for the same wide
range of purposes for which it is used in civilian spheres (e.g.
research, transport and non-destructive testing). Previous
UNSCEAR reports reviewed the first two of these activities
separately. This approach is no longer continued here, since
the countries have not reported the data separately. It is recognized that there may be a degree of overlap between the
categories of nuclear facilities and also that the limited
number of countries responding to the UNSCEAR Global
Survey of Occupational Radiation Exposures constrains theconclusions that can be drawn. National data on occupational
exposure resulting from all military activities are presented
in table A-31. Data from the United States and the United
Kingdom dominate the reported data on occupational­exposure for this practice.
716. In the United States, the USDOE is responsible forstewardship of the nuclear weapons stockpile and the associated facilities, for restoring the environment at related sites

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UNSCEAR 2008 REPORT: VOLUME I

and for energy research [U23]. The facilities covered included
accelerators, fuel/uranium enrichment, fuel fabrication, fuelprocessing, maintenance and support, reactor operation,
research, waste management, weapons fabrication and testing. Exposures may arise via two main routes: (a) the intakeof these materials into the body by inhalation or ingestion (orabsorption through the skin in the case of tritium); and(b) external irradiation by gamma rays and, to a lesser extent,
neutrons. External irradiation tends to be the dominant sourceof exposure for those involved in the production, testing andsubsequent handling of nuclear weapons.
717. The USDOE notes [U23] that the number of monitored workers may not be indicative of the size of the exposed
workforce, because some establishments provide dosimetry
to individuals for reasons other than radiation protection,e.g. security, administrative convenience and legal liability.
As a result, it may not be valid to compare the size of themonitored workforce over time. Similarly, such a large monitored population can confound comparisons of dose. The
average effective dose decreased from 1.1 mSv in the first
period to 0.1 mSv in the last two periods. It appears to havedecreased by a factor of 3 between the periods 1985–1989
and 1990–1994 and by a factor of 1.5 between the periods
1990–1994 and 1995–1999. The annual collective dose at
USDOE facilities has experienced a dramatic fall since thefirst period (1975–1979), from 101 to 13 man Sv. The change
in operational status of USDOE facilities has had the largest
impact on radiation exposure over the years owing to theshift in mission from weapons production to clean-up activities and the shut down of certain facilities. The USDOE
weapons production sites have continued to contribute themajority of the collective dose over these periods.
718. In the United Kingdom, the Atomic Weapons Establishment is the organization whose stewardship is comparable to that of the USDOE. The number of monitored workers
in the United Kingdom has stayed roughly constant, ataround 12,000. The average annual collective effective dose
decreased by a factor of 10 over the first five periods (1975–
1979 to 1995–1999), from 36 to 3.6 man Sv, and remained
constant between the last two periods. A similar pattern isseen with the average annual effective dose incurred bymonitored workers, which over the six periods fell from 3.0
to 0.24 mSv.
719. The UNSCEAR 1993 Report [U6] reviewed thepotential for extrapolation on the basis of normalized collective dose, with the normalization performed in terms ofunit explosive yield for weapons, and per ship or installed
nuclear capacity for the naval propulsion programme. It
concluded that such extrapolation was not viable. Pending
the acquisition of further data, the UNSCEAR 1993 Report[U6] proposed adopting a very simple approach for estimating worldwide exposures from this source, namely that theworldwide collective dose from military activities is greaterby a factor of 3 than the sum of the collective dose in the
United Kingdom and the United States. Four assumptions
underlay the choice of this factor. First, the levels of

military activities in the former Soviet Union and the UnitedStates were broadly comparable. Secondly, the levels of
exposure in the former Soviet Union were greater than in
the United States by an indeterminate amount that did notexceed a factor of 2 in 1975–1989. Thirdly, the levels of
exposure in France have been comparable to those in the
United Kingdom. Fourthly, the exposures in China were notas large as those in the former Soviet Union or in the UnitedStates. The addition in the most recent five-year period of
the French data does not significantly change matters, and it
is concluded that the above simple approach is still the best
available in the circumstances. On the basis of these assumptions, the estimated worldwide number of monitored workers has been roughly constant, at between 300,000 and
400,000 workers. The collective effective dose from military activities would have been about 400 man Sv in 1975–
1979, falling to about 250 man Sv in 1985–1989, 100 man Sv
in 1990–1994, 58 man Sv in 1995–1999 and 52 man Sv in
2000–2002 (figure LXVI). Given the coarseness of theunderlying assumptions, it is not possible to give a precise
estimate of the collective dose; perhaps all that can be concluded is that the worldwide average annual collective dose
during the period analysed was about 50–150 man Sv. The
average effective dose decreased by a factor of 10, from
1.3 mSv in 1975–1979 to 0.14 mSv in 2000–2002. This estimate is inevitably associated with much uncertainty, which
can only be reduced by relevant data from other countries
involved in weapons production.
720. The above data need to be qualified with regard totheir completeness, in particular concerning whether they
include all significant occupational exposures associatedwith military activities. For example, they do not include
occupational exposures incurred in the mining of uranium
used in either the nuclear weapons or the nuclear naval programmes, nor is it clear to what extent the reported data
include exposures arising during the enrichment of uranium
for the weapons and naval programmes or exposures arisingin the chemical separation and subsequent treatment of plutonium. Such omissions, should they exist, are significant
only with respect to the correct assignment of exposures todifferent practices. Any omission here is likely to be compensated for by an overestimate of exposures in other practices (e.g. exposures in the commercial nuclear fuel cycle).
721. The data presented above for all military activities
include occupational exposures for three countries that havedeveloped and deployed nuclear weapons or that have operated nuclear ships, namely France, the United Kingdom andthe United States. Any estimate of worldwide occupational
exposures from military activities can therefore be madeonly by extrapolating from the available data. The result will
inevitably be only very approximate.
722. The contributions of each category to overall levels of
exposure and trends with time are shown in figure LXVII.
The worldwide average annual collective effective doses toworkers resulting from nuclear fuel cycle operations in the
periods 1995–1999 and 2000–2002 are estimated to be aboutANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

1,000 and 800 man Sv, respectively. The contribution ofpractices involving medical uses is estimated to be about
3,540 man Sv for the two periods, which corresponds toabout 75% of the total collective dose resulting from all thepractices involving the use of man-made sources of radiation(4,960 and 4,730 man Sv for the last two periods). The collective effective dose resulting from occupational exposures
to natural sources (at levels in excess of the average levels ofnatural background radiation) is estimated to be about
37,260 man Sv. The largest component of this, 30,360 man Sv,
is associated with mining: 16,560 man Sv due to coal mining
and 13,800 man Sv due to other mining operations (excluding uranium mining, which is accounted for in the nuclear
fuel cycle). The mineral processing industries were not distinguished from mining operations, since the available datain the literature rarely distinguish the exposure due to mining
operations from that due to mineral processing. The new category called “workplaces other than mines” contributes
6,000 man Sv, and the cosmic ray exposure of aircrew contributes 900 man Sv. However, the estimated collective dose
due to natural sources of radiation is associated with muchgreater uncertainty than is the estimated dose due to manmade sources of radiation. The trends are illustrated in figure LXVIII. Trends in exposure due to man-made sources
are illustrated in figure LXIX for each of the main occupational categories considered in this annex. For exposure tonatural sources of radiation, the evaluation of the level ofoccupational exposure was first introduced in the period
1990–1994. With respect to earlier periods, the few data thatdo exist suggest that exposures during mining operations
and mineral processing were greater than those estimatedhere, and possibly much greater, owing to the fact thatsomewhat less attention was given in the past to the control
and ­reduction of exposures during underground mining.
1.

Other exposed workers

723. There is one other group of exposed workers not considered elsewhere, namely those working in diamond mines,
where X-ray screening for diamond theft is conducted undercertain conditions in some countries. The security measures
are implemented to reduce diamond theft and are explicitly
authorized through national regulations and cover a large
spectrum from access control to the use of special equipment
to prevent the employees having direct contact with the diamonds. The radiation dose is due to X-ray screening ofworkers to detect if they have swallowed or hidden diamondsin their bodies [I6]. There is no reliable estimate of the total
number of workers involved in these diamond mines, of how
often they are exposed and of the received dose.

319

uses (which comprise education, veterinary medicine and allother uses involving occupational exposure). In the previous
UNSCEAR reports, the worldwide level of exposure was
extrapolated on the basis of the reported data applying a different methodology for the practices in the nuclear fuel cycleand for the other practices, such as those in medical, industrial and miscellaneous uses. Inevitably, the data provided in
response to the UNSCEAR Global Survey of Occupational
Radiation Exposures were insufficient for estimating worldwide levels of dose. Procedures were therefore developed by
the Committee to derive estimates of worldwide doses fromthe data available for particular occupational categories. Twoprocedures were developed, one for application to occupational exposures arising at most stages in the commercial
nuclear fuel cycle and the other for general application toother occupational categories. For the occupational groups
involved in practices other than those in the nuclear fuelcycle, the approach used in the UNSCEAR 2000 Report toderive estimates of worldwide doses is no longer used here.
For the last two periods, 1995–1999 and 2000–2002, theworldwide level of exposure was derived on the basis of thetrends for the countries that provided data. The number ofworkers exposed to radiation in the medical field is estimated on the basis of the UNSCEAR Global Survey ofMedical Radiation Usage and Exposures. In general, thereporting of exposures arising in the commercial nuclear
fuel cycle is more complete than that of exposures arisingfrom other uses of radiation. Hence the degree of extrapolation from reported to worldwide doses is less, and thisextrapolation can be carried out more reliably than for otheroccupational categories. Moreover, worldwide statistics are
generally available on the capacity and production in various stages of the commercial nuclear fuel cycle. Such data
provide a convenient and reliable basis for extrapolating toworldwide levels of exposure.
725. It is difficult to compare the doses among the countries,
since there are discrepancies in doses related to the same practice. These can be due to differences in the type of dose that isrecorded in the databases, for instance in how the doses below
the recording level are recorded, and differences in the criteria
for including workers in the individual monitoring programme.
Some countries include in their individual monitoring programme workers not who do not work in restricted areas, resulting in an increase in the workforce and a decrease in the average
effective dose. Also, some countries record only doses fromexternal exposure, while others record doses from both internaland external exposure. The application of different ICRP methodologies for intake and dose ­calculations leads to different
dose results. This can be an important source of variation in thedoses reported by the countries for the period analysed, whenmost of the countries changed from ICRP Publication 26 [I43]to ICRP Publication 60 [I47] recommendations.

E. Summary on occupational exposure724. Occupational radiation exposures have been evaluatedfor six broad categories of work: natural sources of radiation, the nuclear fuel cycle, medical uses of radiation, industrial uses of radiation, military activities and miscellaneous

726. Results for the periods 1995–1999 and 2000–2002 aresummarized in table 91 and, in abbreviated form, for thewhole period of interest (1975–2002) in table 92. The contribution of each category to overall levels of exposure and thetrends with time are illustrated in figure XLVII.

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727. The total number of workers exposed to ionizing
radiation is estimated as approximately 23 million. About
57% are employed in practices that include exposure tonatural sources of radiation (13 million workers) andabout 43% in practices that include exposure to man-madesources of radiation (10 million workers). For exposure tonatural sources of radiation, the greatest number of workers, about 11.50 million, are in mining operations (60% incoal mining and 40% in other mining operations, excluding uranium mining). The estimated number of workers
exposed to radon in workplaces other than mines is about
1.25 million, and the number of aircrew exposed to cosmic radiation is 0.30 million. For exposure to man-madesources of radiation, the greatest contribution comes frommedical uses of radiation (75% of the number of workers). About 7.40 million workers are involved in medical
uses of radiation, 0.66 million in practices related to the
nuclear fuel cycle, 0.87 million in practices related toindustrial uses of radiation and 0.57 million in other occupational groups, while 0.33 million are involved in ­militaryactivities.
728. The worldwide average annual collective effective
dose to workers exposed to radiation is estimated to bearound 42,000 man Sv The worldwide average annual collective effective dose to workers exposed to natural sourcesof radiation (in excess of the average levels of natural background radiation) is estimated to be around 37,260 man Sv,
which represents about 93% of the total collective effective
dose. The largest component of this, 30,360 man Sv, comesfrom mining: 16,560 man Sv due to coal mining and13,800 man Sv to other mining operations (excluding uranium mining, which is dealt with as part of the nuclear fuelcycle). The mineral processing industries were not distinguished from mining operations, since the available data in
the literature rarely distinguish the exposure due to mining
operations from that due to mineral processing. The newcategory called “workplaces other than mines”, which
includes industries (food industries, breweries, laundries,etc.), waterworks, shops, public buildings and offices,
schools, subways, spas, caves and closed mines open tovisitors, underground restaurants and shopping centres,
tunnels (construction and maintenance) and sewerage facilities, contributes 6,000 man Sv. The contribution of aircrew
exposed to cosmic radiation is 900 man Sv. However, theestimated collective dose due to natural sources of radiation is associated with much greater uncertainty than that
due to man-made sources of radiation. The trends are illustrated in figure LXVIII. The worldwide average annual collective effective dose to workers exposed to man-madesources of radiation is 4,730 man Sv. The average annual
collective effective dose to workers in the nuclear fuel cycle
for the period 2000–2002 is estimated to be about
800 man Sv. The contribution of the practices related to
medical uses is estimated to be 3,540 man Sv, and of practices related to industrial uses and miscellaneous uses about
400 man Sv. Medical uses of radiation contribute about
75% of the collective effective dose due to exposure to
man-made sources of radiation.

729. The average annual effective dose to monitored workers varies widely from occupation to occupation and also
from country to country for the same occupation. On thebasis of the reported data, the average annual effective dose
to monitored workers in industry is less than 1 mSv. In particular countries, however, the average annual dose for someof these occupations is several millisieverts or even, exceptionally, in excess of 10 mSv. The average annual effective
doses to workers in the nuclear fuel cycle are in most cases
higher than the doses to those in other occupations. For thefuel cycle overall, the average annual effective dose is about
1.4 and 1.0 mSv for the last two periods (tables 72 and 92).
For the mining of uranium, the average annual effective dose
to monitored workers in countries reporting data fell from
3.9 mSv in 1995–1999 to about 1.9 mSv in 2000–2002. For
uranium milling operations, the average annual effective
dose fell from 1.6 mSv in 1995–1999 to about 1.1 mSv in
2000–2002. For fuel fabrication, the average annual effective dose is about 1.6 mSv. For reactor operation, the averageannual effective dose is 1.5 mSv and 1.0 mSv for the last two
periods. However, there are very wide variations aroundthese average values. The doses for decommissioning arearound 2 mSv. The individual doses for fuel reprocessing areabout 0.9 mSv, whereas those for fuel enrichment are much
lower, less than 0.1 mSv.
730. Trends in exposure over the period 1975–2002. Trends
in exposure resulting from man-made sources are illustrated
in figure LXIX for each of the main occupational categories
considered in this annex. For exposure to natural sources ofradiation, the evaluation of the level of exposure was first
introduced in the period 1990–1994. With respect to earlier
periods, the few data that do exist suggest that exposuresduring mining operations and mineral processing weregreater than those estimated here, and possibly much greater,owing to the fact that somewhat less attention was given in
the past to the control and reduction of exposures during
underground mining.
731. The worldwide average annual number of workers
involved with man-made uses of radiation is estimated tohave increased from about 2.8 million to about 10 million
between the first and sixth periods (table 92). The greatest
increase (from about 1.3 million to about 7.4 million) was in
the number of monitored workers in medicine, which represents about 75% of the workforce. The number of monitored
workers for the nuclear fuel cycle also increased significantlyin the first three periods, from about 0.6 million in the first
period to about 0.9 million in the third period, but it dropped
to 0.8 million for 1990–1994 and to about 0.7 million for
1995–2002. The main reason for this significant decrease is
the decline in the number of workers in mining operations.
732. The annual collective effective dose averaged overthe five years for each of the first three periods (1975–1989)
for all operations in the nuclear fuel cycle varied little
around the average value of 2,500 man Sv, despite a factorof 3–4 increase in electrical energy generated by nuclear
means. The electrical energy generated has continued toANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

increase, but the average annual collective effective dosefell by a factor of about 2, to 1,400 man Sv, in 1990–1994,
and dropped to 1,000 man Sv and 800 man Sv in the last two
periods. A significant part of this decrease came from thedramatic reduction in the uranium mining component, from
1,100 man Sv in 1985–1989 to 310 man Sv in 1990–1994,
85 man Sv in 1995–1999 and 22 man Sv in 2000–2002.
These last figures may be underestimated owing to limited
data, besides which some reported data are related to thedecommissioning phase, so they must be viewed with somecaution. However, other indicators, such as the reduction in
the amount of uranium mined, the closing of many underground mines, a more general move to open-pit mining andthe introduction of modern techniques, support the viewthat a substantial reduction has taken place. In other parts of
the nuclear fuel cycle the situation is more varied. In repro­
cessing, for example, the downward trend in earlier periods—53, 46 and 36 man Sv—has been reversed, with anincrease to about 68 man Sv for 2000–2002, associated withan increase in the number of workers. However, apart frommining, the other important element within the nuclear fuelcycle is reactor operation, in which the average annual­collective effective dose, after increasing from 600 to
1,100 man Sv over the first three periods, dropped to
900 man Sv for 1990–1994, to 800 man Sv for 1995–1999
and to 620 man Sv for 2000–2002.
733. The average annual effective dose to monitored workers in the nuclear fuel cycle has been consistently reduced
over the whole period, from 4.1 mSv to 1.0 mSv. There aresome variations between parts of the nuclear fuel cycle andbetween countries. Of particular note is the fact that, in thefirst three periods, the dose to monitored workers at LWGRs
increased from 6.6 mSv to 13 mSv, and while no specificvalues for the three latest periods were reported, other indicators at least suggest that the high level of exposure wasmaintained.
734. The number of monitored workers increased by a factor of 6 over the six periods, from 1.3 million to 7.4 million.
The largest increase, from 2.3 million to 7.4 million, wasobserved in the last two periods (1995–2002), because inthis case the estimate was based on more complete information from the UNSCEAR Global Survey of Medical Radiation Usage and Exposures. The worldwide average annual
collective effective dose due to all medical uses of radiation,
about 1,000 man Sv, changed little over the first three fiveyear periods. It then dropped significantly, to 760 man Sv, in
1990–1994, but increased to 3,540 man Sv over the last two
periods. A clear downward trend is evident in the worldwide
average effective dose to monitored workers, which decreased
from about 0.8 mSv in the first five-year period to about
0.3 mSv in 1990–1994, then increased in the last two periods, reaching 0.5 mSv in 1990–2002. However, there wasconsiderable variation between countries. For diagnostic
radiology, the average effective dose remained constant overthe last two periods. This may reflect the influence of thehigher doses due to interventional procedures. However, thenumber of workers increased by a factor of 7, from about

321

1 million to 6.7 million, between 1990–1994 and 1995–2002.
Consequently the collective effective dose has increased in
the same proportion as the number of workers.
735. The worldwide average annual collective effective
dose due to all industrial uses of radiation was fairly uniform over the period 1975–1984 at about 900 man Sv. It
decreased, however, by a factor of almost 2 in the second
half of the 1980s (to 510 man Sv) and then fell further, toabout 360 man Sv in 1990–1994, and to about 300 man Sv
in the last two periods. In general, there was a declining
trend in collective dose for the last two periods. However,there was a trend of increasing collective dose for industrial
radiography. The same trend is reflected in estimates ofindividual dose: the average annual effective dose to monitored workers decreased from about 1.6 mSv in 1975–1979
to 1.4 mSv in 1980–1984, 0.9 mSv in 1985–1989, 0.5 mSv
in 1990–1994 and 0.3 mSv in 2000–2002.
736. It should be noted that in UNSCEAR reports prior to
1990–1994, the category “industrial uses” included a component reflecting “educational uses”, which tended to distortthe data. Since then, educational uses have been dealt with ina separate category, and the industrial data for earlier years
have been adjusted to remove the educational component.
For military activities, the average individual and collective
doses both fell by a factor of about 10 over the whole period,
from 1.3 mSv to 0.1 mSv and from 420 man Sv to about
50 man Sv, respectively.
737. The estimates of occupational radiation exposure inthis annex have benefited from a much more extensive andcomplete database than was previously available to the Committee. The efforts by countries to record and improve dosimetric data were reflected in the responses to the UNSCEAR
Global Survey of Occupational Radiation Exposures and the
UNSCEAR Global Survey of Medical Radiation Usage andExposures and have led to improved estimates and understanding of occupational exposures. However, the Committee considers that further guidance on these matters would
help improve the quality of its assessments.
738. The Committee’s current estimate of the worldwide
collective effective dose due to occupational exposure fromman-made sources is 4,430 man Sv (about 800 man Sv toworkers in the nuclear fuel cycle, about 3,540 man Sv to workers in medical uses and about 400 man Sv to workers in industrial uses, military activities and miscellaneous activities).
This estimate is about the same as that made by the Committee for the late 1970s. This is because, for the latest period, theevaluation of occupational exposure in the medical field ismore reliable, and it contributes about 75% of the collective
dose. The figure for occupational exposure from man-madesources has changed greatly since 1970, when occupational
exposure was dominated by the practices in the nuclear fuelcycle. Except for medical uses, all other practices have showna reduction in the level of exposure. A significant part of the
reduction comes from the nuclear fuel cycle, particularly fromuranium mining. However, reductions are seen in all the main

322

UNSCEAR 2008 REPORT: VOLUME I

categories: industrial uses, medical uses, military activities
and miscellaneous uses. This trend is also reflected in theworldwide average annual effective dose due to occupational
exposure, which has fallen from about 1.7 mSv to 0.5 mSv(table 92).
739. No attempt has been made to deduce any trends in theestimates of dose for occupational exposure to natural sourcesof radiation, because the supporting data are somewhat limited. The UNSCEAR 1988 Report [U7] made a crude estimate of about 20,000 man Sv for this source, which was
subsequently revised downward to 8,600 man Sv in the
UNSCEAR 1993 Report [U6]. The UNSCEAR 2000 Report[U3] estimated a value of 11,700 man Sv, of which
6,000 man Sv was due to elevated levels of radon and its
progeny in workplaces other than mines, 5,700 man Sv to
extraction and processing activities, and 800 man Sv to the
exposure of aircrew to cosmic radiation. In the current report,
the estimate for the collective dose has risen to 37,260 man Sv,
the largest contribution coming from mining operations—16,560 man Sv from coal mining and 13,800 man Sv

from other mining. About 6,000 man Sv is due to exposure toradon and its progeny in workplaces other than mines, and
900 man Sv is due to the exposure of aircrew to cosmic radiation. The estimate is still considered to be crude, although thedata on occupational exposure in Chinese coal mines have
reduced some of the uncertainty in this estimate. The main
contributor to the average collective dose is coal mining. On
the basis of the data presented in the literature, the level of
exposure may be declining, since the average effective dose
decreased by a factor of 2 for workers in the Chinese coalmines [T4]. According to the literature, the level of exposure
has decreased in nine European countries, while the averagecollective dose and average effective dose decreased by a­factor of almost 2 from 1996 to 2000 [F15].
740. The doses for each practice for the year 2007 were
projected on the basis of the trends for all countries. In general, the trend is for an increasing number of workers, anddecreasing collective doses and effective doses for all thepractices in the categories of medical and industrial uses ofradiation, and also for the category of miscellaneous uses.

CONCLUSIONS ON PUBLIC AND WORKER EXPOSURE
741. Exposure to natural sources of radiation is an unavoidable fact of the human condition. The estimates of the
global average per caput values of exposure to natural
sources of radiation are essentially the same as in the
UNSCEAR 2000 Report. The estimated value of worldwide
average annual exposure to natural radiation sources remains
at 2.4 mSv. The normal range of exposures to the various
components is presented in table 12. The dose distribution
worldwide is expected to follow approximately a log-normal
distribution, and most annual exposures would be expected
to fall in the range 1–13 mSv.
742. The values for occupational exposure for the periods
1995–1999 and 2000–2002 have changed greatly comparedwith those in the UNSCEAR 2000 Report. The collective
effective dose resulting from exposures to natural sources (inexcess of the average levels of natural background) is estimated to be about 37,260 man Sv, about three times higher
than the value estimated in the UNSCEAR 2000 Report[U3]. The largest component of this, 30,360 man Sv, comesfrom mining (16,560 man Sv due to coal mining and13,800 man Sv due to other mining operations, excluding
uranium mining); 6,000 man Sv is due to “workplaces otherthan mines” and 900 man Sv is due to the exposure of aircrew to cosmic radiation. The large difference with respect
to the UNSCEAR 2000 Report comes from the level of
exposure in coal mines. For the current period, the estimate
is based on an assessment of exposure in Chinese mines,
which represents a very large number of workers. No matter
how the estimates are made, the collective dose due to natural sources of radiation is associated with much greater
uncertainty than that due to man-made sources of radiation.
The trends are presented in table 92.

743. Residues due to conventional mining operations also
give rise to very large quantities of material with enhanced
levels of naturally occurring radionuclides; these represent
a challenge regarding both the disposal of the residues andsite restoration. The large diversity of ores containing low
levels of nuclides from the uranium and thorium families,
which may be concentrated in products, by-products andwastes, complicates the problem, and the detailed picture ofworldwide exposure is far from complete. Although doses
to the public are usually low, of the order of a few microsieverts or less, some exposed groups can receive doses inthe ­millisievert range, which may deserve attention.
744. For all fuel cycle operations (mining and milling,
reactor operation and fuel reprocessing), the local andregional exposures are estimated to be 0.72 man Sv/(GW a).
For the present world nuclear energy generation of 278 GW a,
the collective dose per year of practice is of the order of
200 man Sv. The collective doses due to globally dispersedradionuclides are delivered over very long periods; if thepractice of nuclear power production is continued for 100
years at the present capacity, the maximum annual per caput
effective dose to the global population would be less than
0.2 μSv. This dose rate is low compared with that due tonatural background radiation.
745. The current estimate for the total collective dose toworkers in practices using man-made sources of radiation
has changed the figure of occupational exposure. The collective effective dose has in the past been dominated by practices in the nuclear fuel cycle, but the current estimate has
shown that occupational exposure in the medical field has
become dominant. The collective effective dose in practices

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

using man-made sources of radiation may be around
4,730 man Sv (about 800 man Sv to workers in the nuclear
fuel cycle, about 3,540 man Sv to workers in medical uses,
and about 400 man Sv to workers in industrial uses, militaryactivities and miscellaneous uses). These figures haveincreased compared with the estimates in the UNSCEAR
2000 Report [U3]; the most important decrease was due tothe nuclear fuel cycle. The trends are presented in table 92
and figure LXVII.
746. The main contribution to the global collective dose to
the public due to man-made sources comes from the testingof nuclear weapons in the atmosphere in the period between
1945 and 1980. The estimated global average annual per
caput effective dose reached a peak of 110 μSv in 1963 andhas since decreased to about 5 μSv (mainly due to residuallevels of 14C, 90Sr and 137Cs in the environment). The average323

annual doses are higher than the global average by 10% in
the northern hemisphere (where most of the testing took
place) and are lower in the southern hemisphere.
747. In addition to areas related to atomic bomb production
and testing, military uses of radiation have also left a legacyof numerous small contaminated sites across the planet.
Efforts to decontaminate these sites and return them to public use have been a focus of attention in many countries.
Exposures and collective doses are site-specific; once theareas are defined, exposures can be constrained and clean-up
procedures implemented. In general, site release criteria
consider annual individual doses for a hypothetical criticalgroup of people in the range 0.3–1.0 mSv. Average local andregional annual individual doses will be at least one order ofmagnitude lower, and the contribution to the worldwide­population doses will most probably be negligible.

TablesTables available as MS Excel workbooks on the attached CD-ROM
Public.xls
A-1 Natural radionuclide content of soil
A-2

Activity concentration in building materials

A-3

Activity concentration of naturally occurring radionuclides in drinking water (mBq/L)

A-4

Nuclear power plants operating in the period 1998–2002

A-5

Energy generated by nuclear power plants in the period 1998–2002 (GW a)

A-6

Noble gases released from nuclear power plants in airborne effluents (GBq)

A-7

Tritium released from nuclear power plants in airborne effluents (GBq)A-8

Iodine-131 released from nuclear power plants in airborne effluents (GBq)

A-9

Carbon-14 released from nuclear power plants in airborne effluents (GBq)

A-10 Particulates released from nuclear power plants in airborne effluents (GBq)
A-11 Tritium released from nuclear power plants in liquid effluents (GBq)
A-12 Other radionuclides released from nuclear power plants in liquid effluents (GBq)
A-13 Releases from nuclear fuel cycle reprocessing plants in airborne effluents (GBq)
A-14 Releases from nuclear fuel cycle reprocessing plants in liquid effluents (GBq)
Workers.xls
A-15 Dose monitoring and recording procedures for occupational exposure
A-16 Exposures to workers from natural sources of radiation
A-17 Exposures to workers from uranium mining
A-18 Exposures to workers from uranium milling
A-19 Exposures to workers from uranium enrichment and conversion
A-20 Exposures to workers from fuel fabrication
A-21 Exposures to workers at nuclear power reactors
A-22 Exposures to workers from fuel reprocessing
A-23 Exposures to workers from research related to the nuclear fuel cycle
A-24 Exposures to workers from medical uses of radiation
A-25 Exposures to workers from diagnostic radiology
A-26 Exposures to workers from nuclear medicine according to job category
A-27 Exposures to workers from all medical uses of radiation
A-28 Exposures to workers from various industrial uses of radiation
A-29 Exposures to workers from all industrial uses of radiation
A-30 Exposures to workers from miscellaneous uses of radiation
A-31 Exposures to workers from military activities

325

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Table 1.

327

Dose conversion factors and dose conversion coefficients for natural radionuclides [U3]Radionuclide

Dose conversion factora, DCFsoil((nGy/h)/(Bq/kg))

K

0.041 7

U series

0.462

Th series

0.604

Radionuclide

Dose conversion coefficientb (Sv/Bq)

40

238

232

Inhalation, einh(50)Ingestion, eing(50)U

2.9 × 10–6

4.5 × 10–8

U

3.5 × 10–6

4.9 × 10–8

Th

1.4 × 10–5

2.1 × 10–7

Ra

3.5 × 10–6

2.8 × 10–7

Pb

1.1 × 10–6

6.9 × 10–7

Po

3.3 × 10–6

1.2 × 10–6

Th

2.5 × 10–5

2.3 × 10–7

Ra

2.6 × 10–6

6.9 × 10–7

Th

4.0 × 10–5

7.2 × 10–8

U

3.1 × 10–6

4.7 × 10–8

238

234

230

226

210

210

232

228

228

235

a External dose rates due to radionuclides in soil.
b Effective dose per unit intake due to internal exposure for adults.

Table 2.

Collective effective dose per unit release of radionuclides from nuclear reactors [U3]Type of release

Reactor typea

Pathway

Collective dose per unit release (man Sv/PBq)

PWR, LWGR, FBR, HWR

Immersion

0.11

BWR

Immersion

0.43

GCR

Immersion

0.9

H

All

Ingestion

2.1

C

All

Ingestion

270

External

4.5

Ingestion

250

Inhalation

49

External

1 080

Ingestion

830

Inhalation

33

Radionuclide

Noble gases

3

14

Airborne
I

131

Particulate

Liquid

All

All

3

H

All

Ingestion/
inhalation

0.65

Other

All

Ingestion

330

a PWR: pressurized water reactor; LWGR: light-water-cooled, graphite-moderated reactor; FBR: fast breeder reactor; HWR: heavy-water-cooled and -moderated reactor; BWR:
boiling water reactor; GCR: gas-cooled, graphite moderated reactor.

328

UNSCEAR 2008 REPORT: VOLUME I

Table 3.

Collective effective dose per unit release of radionuclides from fuel reprocessing plants [U3]Type of release

Radionuclide

Collective dose per unit release (man Sv/TBq)

H

0.0021

C

0.27

3

14

Kr

0.000007 4

129

I

44

I

0.3

Cs

7.4

H

0.000001 4

C

1

Sr

0.0047Ru

0.0033

I

0.099

Cs

0.098

85

Airborne

131
137

3

14
90

Liquid

106

129
137

Table 4.

Population distribution of cosmic ray dose rates outdoors at sea level [U3]Latitude (degrees)

Population in latitude band (%)

Effective dose rate (nSv/h)Northern hemisphere

Southern hemisphere

Directly ionizing component

Neutron component

80–90

0

0

32

11

70–80

0

0

32

11

60–70

0.4

0

32

10.9

50–60

13.7

0.5

32

10

40–50

15.5

0.9

32

7.8

30–40

20.4

13.0

32

5.3

20–30

32.7

14.9

30

4

10–20

11.0

16.7

30

3.7

0–10

6.3

54.0

30

3.6

Total

100

100

Northern hemisphere

31.0

5.6

Southern hemisphere

30.3

4.0

World

30.9

5.5

Population-weighted average

Table 5.

Population-weighted average annual effective doses (mSv) due to cosmic radiation [U3]Conditions

Directly ionizing component

Neutron component

Total

Northernhemisphere

Southernhemisphere

World

Northernhemisphere

Southernhemisphere

World

World

Outdoors, at sea level

0.27

0.27

0.27

0.05

0.04

0.05

0.32

Outdoors, adjusted for altitude

0.34

0.33

0.34

0.12

0.09

0.12

0.46

Adjusted for altitude, shielding
and occupancy

0.29

0.28

0.28

0.10

0.07

0.10

0.38

Table 6. Absorbed dose rates in air (nGy/h)Data not referenced are from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
Region/country

Population
(106)
[C17]

Outdoors
Cosmic radiation
Range

Terrestrial radiation
Mean

Range

Total
Mean

Cosmic radiation
Range

Mean

Range

Terrestrial radiation
Mean

Range

Total
Mean

Range

105

80–126AfricaLibyan Arab Jamahiriya

5.9

Mauritius
Tanzania (United Rep. of) [B6]

28

25–31

23

18–24

51

48–54

1.24

98

80–12637.4

104

98–121

North America
Canada33.1

Mexico

107

24

11–44

54

31–75

88.3

23–184105

37–217151

85–191

Central America
Costa Rica [M25, M30]

5.5

36

29.3–80.2

29.9

5.6–66.6

65.9

35–147

Cuba [T6, T7]

11.4

34

32–67

24

4–162

55

38–196

27

26–54

30

10–76

44

37–10375–205

21

16–26

123

87–160144

103–186156

57–319124.1

1.12–174.1

East Asia
Azerbaijan

8

Bangladesh[A4, A5, A6, H24, U40]

37

102

45–160147

120

44–245China [C11, Z4]

1 313

69.9

12.7–1 300

81.5—Taiwan [L11, L12]

23

52

24–68

79

India [N2]1 095

Indonesia

245

27.5

21.1–61.9

23.9–40.1

67.5

Japan [A2, F4]

124.7635.3

30.1–59.4

Kazakhstan

15.2

Korea, Rep. of [K16]48.8

Pakistan [B50]166

Philippines

75.9

27

30–45

25.7–58

140

11.6–523
24

23–52

101

66–189

125

41.5
40

45–102
78.3

60–500

150–28079

59
24

18–200

1.0–9745329

21

52.4–106.5

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Mean

Indoors

Population
(106)
[C17]

Outdoors
Cosmic radiation
Mean

Range

Indoors

Terrestrial radiation
Mean

330

Region/country

Range

Total
Mean

Cosmic radiation

Terrestrial radiation

Total

Range

Mean

Range

Mean

Range

Mean

Range

69–187.6

16.8

13.2–31.3

115

70–165

131.8

83.2–196.3

West Asia
Armenia

2.98

Islamic Republic of Iran

68.7

40.5

Kuwait2.4

35

Turkey [K2]

70.4

15.7

33.0–57.6

71

36–13052
8.4–35.6

48.8

111.5
87

15–80

65

29

90

119

32–94

North Europe
5.5

31

35

25–70

66

56–10131

54

19–25985

50–290Finland [A19, A20]

5.2

32

71

45–139103

77–17132

73

24–181

105

56–213

31

40

4–83

71

25

23

14–32

48

Iceland [E4]
Lithuania [L3]

3.45

33

32–35

62

46–82

95

79–115

26

26–28

81

34–224107

53–250Swedena,b[M27, M29, S15]

9

33

32–50

64

10–58097

40–630

36

17–75

98

10–1 250120

20–1 300

45–120

26

20–36

60

32–18086

55–20080

20–70062

10–14094

43–168

West Europe
Belgium [G10]

10.4

33a

Germany

82.4

32

57

Ireland[C24, M9, M15, M16]

3.84

33

32

2–110

65

35–14326

74

11–209112

57–243

31

26–43

105

0–690

136

29–71773.1

40–12495.2

60–151100

55–215

Italy [B29, C4]

38

32–36

32–54

43

13–80

76
89

Liechtenstein

0.03

38

34.6

72.6

Luxembourg

0.22

32

49

14–73

81

Spain
[Q1, Q2, Q6, S42, S43]

40.4

34.6

30.8–41.0

50.4

19–88

85

50–12922.1

19.7–26.2

Switzerland [L22]

7.5

39

34–68

42

14–11881

53–15531

27–54

UNSCEAR 2008 REPORT: VOLUME I

Denmark[A1, N14, S36, U42]

Region/country

Population
(106)
[C17]

Outdoors
Cosmic radiation
Mean

Range

Indoors

Terrestrial radiation
Mean

Range

Total
Mean

Cosmic radiation

Terrestrial radiation

Total

Range

Mean

Range

Mean

Range

Mean

Range

22–35

75

57–93

100

80–13085

42–2 000d

119

74–2 000

93.8

54.7–193.8

110

56–20055

32–109

East Europe
7.4

30

27–42

70

48–96

100

75–14025

Czech Republic [M18]

10.2

34

32–62

66

6–245

100

40–28534c

Poland [B24, B25]

38.15

33.5

31.3–60.8

47.4

18.8–86.0

80.9

51.0–126.2

Romania[B1, B2, C20, C21, C22, I1]

21.3

33

32–38

59

20–12592

52–163

Slovenia [A18]

2.0

32

30–47

56

4–147

27

26–30

83

30–17075

40–25036

20–10123

0–77

South Europe
Albania [I4]

3.6

94

77.2–103

Croatia4.5

115

70–140

Greece[C18, C19, P14, S5]

10.7

31

17–88

Montenegro

0.63

63

28–150Oceania

New Zealand

4.1

32a

a Ionizing component.
b Average values refer to population-weighted mean.
c Assumed same as outdoor.
d Excluding area in Jachymov contaminated with naturally occurring radioactive material.

44

2–90

76

34–12232c

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Bulgaria

331

Region/country

Population
(106)

332

Table 7. Distribution of population with respect to the outdoor absorbed dose rate in air due to terrestrial gamma radiation
Data not referenced are from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
Population (103) residing in areas with various levels of outdoor absorbed dose rate in air (nGy/h)
<20

20–29

30–39

40–49

50–59

60–69

70–79

80–89

90–99

100–199

200–299>300

AfricaLibyan Arab Jamahiriya

1.50

1 500
Central America

Costa Rica

5.5

605

918.5

605

2 926

412.5

Cuba

11.20

6 000

5 000

200

2

<1

27.5

East Asia
57.08

China

1 282.35

Indonesia

213.68Japan [U3]124.76

Korea, Rep. of [U3]44.61

Malaysia [U3]19.64

Philippines

75.90

94 88448 2031 760

26 775

1 200

1 510

12 001

4 780

10 9702 921

16 8106 890

232 426364 745279 327

122 493

75 40436 07475 011

1 984

620

3 096

10 252832

31 97535 61617 301

3 929

2 968

1 431

9 619

26 46320 561

23 38239 546

5 193

3 096

9 605

4 097

2 220

1 724

4 421

4 421

2 211

11 053

984

213

1 214

2 498

8 487

6 248

1 377

688.5

275.4

45.90

12 050.64

24 356.32

2 550.4318.8

3 060.48

27 92214 535

4 284

UNSCEAR 2008 REPORT: VOLUME I

BangladeshWest Asia
Azerbaijan

8.00

Islamic Republicof Iran

63.76

6 000

1 500

500
3 188

1 402.72

16 832.64
North Europe

Denmark [A17]5.20

Estonia [U3]1.47

Iceland [E4, T3]

0.30

Finland[A19, A20, C5]

5.20

Lithuania [G14]3.45

Sweden[S15, S45]

8.88

6

214

250

2 100

2 200

600

50

5

25

149

314

367

592

9

150

150
922

1 143

2 633

174

816

606

1 386

188

455

1 135

1 370

1 289

1 094

1 024

396

798

328

840

564

2

0.2

Region/country

Population(106)

Population (103) residing in areas with various levels of outdoor absorbed dose rate in air (nGy/h)
<20

20–29

30–39

40–49

50–59

60–69

70–79

80–89

90–99

100–199

200–299

>300

3 125

0

West Europe
10.22

300

2 200

2 400

2 600

2 500

200

20

Germany [U3]

81.10

700

8 600

10 00020 90028 000

9 600

1 500

800

700

300

Ireland [C6, M9]

3.84

298

787

1 148

992

251

7

41

0

0

2

Italy
[B29, B31, C4]

57.30

125

50

5 600

28 050

8 100

1 950

250

3 550

6 500

Luxembourg

0.45

31.9

14.3

57.2

250.2

90.2

4.4

Netherlands [U3]

15.58

3 459

5 484

2 353

2 976

1 262

47

Switzerland

6.71

60

620

1 100

3 900

570

70

160

70

60

100

United Kingdom
[U3]

54.00

6 000

12 00030 0006 000

East Europe
Bulgaria

9.41

170

357

4 756

1 130

214

1 472

Czech Rep. [M7]

10.30

3

89

262

605

1 898

4 342

1 846

829

252

177

Hungary [U3]

10.14

163

479

836

1 017

1 316

3 488

1 163

765

367

530

Poland

38.12

426

3 219

13 09715 528

4 208

1 193

419

30

Romania [I1]

21.83

293.1

1 309.1

4 149.2

6 404.2

5 096.3

3 878.6

721.6

562.8

45.1

Russian Fed. [U3]148.10460

7 150

22 80084 4705 730

17 800

5 330

3 910

Slovakia [U3]

5.29

192

721

1 364

1 292

868

498

243

85

450
22

17

South Europe
3.50

50

Greece10.36

Montenegro

0.60

31

Portugal [U3]

9.43

333

Spain

40.84

50

100

100

500

2 000

300

200

100

50

50

1 160

5 605

1 067

1 250

572

147

225

231

50

50

45

155

117

173

444

1 814

606

1 325

653

313

582

417

2 352

594

1 198

5 644

5 181

10 403

2 871

912

2 424

8 477

80

333

Albania [U3]

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Belgium [G10]Population
(106)

334

Region/country

Population (103) residing in areas with various levels of outdoor absorbed dose rate in air (nGy/h)
<20

20–29

30–39

40–49

50–59

60–69

70–79

80–89

90–99

100–199

200–299>300

Oceania
New Zealand

3.80

1 570

1 390

600

200

40

<10

Total
Total

136 641

148 346

193 504

373 330

526 162488 173

193 160

122 170

66 983145 808

13 527

117

Fraction of total

0.056 7

0.061 60.080 40.155 00.218 50.202 70.080 20.050 70.027 80.060 60.005 60.000 1

Cumulative fraction

0.056 70.118 4

0.198 70.353 80.572 30.775 00.855 20.906 00.933 80.994 30.999 91.000 0UNSCEAR 2008 REPORT: VOLUME I

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Table 8.

335

Reference annual intake of air, food and water [U3]IntakeInfants (1 year)

Children (10 years)

Adults

5 600

7 300

Breathing rate (m3/a)
Air

1 900
Food consumption rate (kg/a)

Milk products

120

110

105

Meat products

15

35

50

Grain products

45

90

140

Leafy vegetables

20

40

60

Roots and fruits

60

110

170

Fish products

5

10

15

Water and beverages

150

350

500

Table 9. Reference values for concentration of radionuclides of the uranium and thorium series in human tissues(mBq/kg) [U3]Radionuclide

Lung

Liver

KidneyMuscle and other tissues

Bone

U

20

3

30

5

100

Th

20

9

5

1

20–70

Ra

4.1

4.1

4.1

4.1

260

Pb

200

400

200

100

3 000

Po

200

600

600

100

2 400

Th

20

3

3

1

6–24

Ra

20

3

2

2

100

238

230

226

210

210

232

228

Region/country

Area

Reference

336

Table 10. Examples of areas of high natural radiation background
Data not referenced are from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
Soil concentration (Bq/kg)
K

40

U

238

Ra

226

Exposure rate in air (nGy/h)Th

Outdoors

232

IndoorsRn (Bq/m3)

Rn (Bq/L)

222

222

Outdoors

Indoors

69

82

Water

High cosmic radiation
China

Ganzua

[Z1]

73

China

Qinghaia[S16, Z1]

95 (65–127)

China

Sichuana

[Z2]

82

China

Tibeta[S16, Z1]

121 (80–140)

United States

Denver, Colorado

[S26]

196

Brazil

Araxá

[V18]

BrazilCaetité[B27]

Brazil

R.G. Norte[M3, M4]

BrazilPhosphate area, PE

[A15,
M4]

United States

Reading Prong, New Jersey

[S27]

2 800

941

50
38–30069

108 (54–253)

4–140

29–207170

Uranium and thorium areas (volcanic intrusive)
BrazilPocos de Caldas, MG, urban areas[S2]

145 (93–244)

BrazilPocos de Caldas, MG, rural areas[V18]

280 (130–1 500)

200 (130–340)

130 (56–280)

204 (50–1 046)

Czech Republic

Central Bohemia, Pluton middle area

[M7]

90–170119

2–25

442 (10–20 870)

Italy

Lazio

[B29,
B32]

175 (120–270)

250 (105–440)

119 (26–1 036)

Italy

Campania

[B29,
B32]

198 (141–243)

310 (115–720)

95 (13–172)

Italy

Orvieto

[U3]

560

Italy

Southern Tuscany[B30]

150–300

190 (40–350)

200 (30–1 240)

Niue Island

Pacific[S27]

Max. 1 100

Romania

Crucea and Grinties[B38]

988–1 599

486

68–22057

76–27574–15931

UNSCEAR 2008 REPORT: VOLUME IUranium areas

Region/country

Area

Reference

Soil concentration (Bq/kg)
K

40

U

Exposure rate in air (nGy/h)

Ra

238

Th

226

232

Outdoors

IndoorsRn (Bq/m3)

222

Outdoors

IndoorsRn (Bq/L)

222

Water

Monazite sand coastal areas
Guarapari and Meaipe, ES[S2]

84 (26–300)b

China

Yangjiang, Quangdong

[S27]

370

Egypt

Roseta coastal area

[S27]

20–400India

Kerala and Madras[G3, N1]

1 500(845–5 270)
Thermal waters

Austria

Bad Gastein

[S27]

1 480

China

Sichuan, Jiangzha[X1]

Hungary

Mount Gellért[S27]

Up to 7.15

India

Tuwa

[S27]

4–40

Indonesia

West Java[S27]

48–252

2.4–422

0.5–66

97 (33–224)

Islamic Republicof IranRamsar[M32,
S27]

300–945

80–50 000

15–47

765
(80–100 000)

Islamic Republicof IranMahallat

[S26,
S27]

364–873

500–7 300

15–41

300–3 800

Japan

Misasa[S27]

437

Slovenia

Podcetrtek

[S27]

1–63

Slovenia

Spas

[V13]

256–9 140

(22–22 000

68 000–340 000

45–83
1 153
(100–105 000)

60–15465(0–500)

2 745(55–31 000)64
(1–160)30(6–200)600(55–1 000)

710
(145–2 730)

15–279OthersAzerbaijan

800–1 000

100–7 000

500–2 500

100–1 000

877–8 770

China

Cave dwellings, Ganzu[S16, Y1]

21–3 660

China

Cave dwellings, Yanan[W12]

32–278Indonesia

Bangka Island

330 (90–540)

Indonesia

Karimu Island

310 (200–410)

Philippines

San Vicente

300 (75–1 558)

Russian Federation

Yssyk-Kul (Kyrgyzstan)

100–150

10–160Up to 300

167 (max. 416)

162–352

337[Z3]

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

BrazilArea

Reference

Soil concentration (Bq/kg)
K

40

U

238

810–1 240

Exposure rate in air (nGy/h)

Ra

226

60–250Rn (Bq/m3)

Rn (Bq/L)

222

222

232

Th

Outdoors

Indoors

Outdoors

Indoors

42–71

136–260

197–377

10–210

150–1 400

Spain

Galicia South, Arribes del Duero,
Sierra de Guadarrama,
Campo de Arañuelo

[M11,
Q6, Q8]

Switzerland

Tessin, Alps, Jura

[S27]

United Kingdom

Kerrier district, south-westpeninsular[W1]<2–17 000

United Kingdom

South Wales caves[F16]max. 3 094

338

Region/country

Water

100–200

World average for natural background radiation
UNSCEAR 2000 Report

[U3]

420

33

32

45

59

84

39

Table 11. High-background areas: distribution of population with respect to total effective dose
Data from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
Region/country

Area

Distribution of population (103) residing in high-background areas with various levels of total effective dose (mSv/a)
3.0–3.49

3.50–3.99

4.0–4.49

4.5–4.99

6.0–6.99

7.0–7.99

8.0–8.99

9.0–9.99>10

Population (103)

20

10

23

31

20

200

799

5.0–5.99

West Asia
Islamic Republicof Iran [S28]Ramsar

110

125

120

140
East Europe

Czech Republic[M7]

Central Bohemian Pluton

41

39

32

31

49

36

28

20

15

54

345

Central Bohemian Pluton

2

3

3

3

6

6

5

5

4

28

65

Central Moldanubian Pluton

36

29

21

18

25

16

11

7

5

12

180

Trebic Massif

22

21

18

17

27

20

15

11

8

29

188

Krkonose-Jizera Pluton

59

44

30

24

31

18

12

7

5

11

241

Carlsbad Pluton

48

37

25

21

27

17

11

7

4

11

208

20

1 860

West Europe
Spain

Galicia South, Arribes del Duero,
Sierra de Guadarrama, Campo de
Arañuelo

40

800

1 000

UNSCEAR 2008 REPORT: VOLUME I

a External exposure not including neutrons.
b Kerma rates of up to 5 460 nGy/h can be measured at localized spots [S2].

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

339

Table 12. Public exposure to natural radiation
Source of exposure

Annual effective dose (mSv)
Average

Cosmic radiation

External terrestrial radiation

Inhalation

Directly ionizing and photon component

0.28

Neutron component

0.10

Cosmogenic radionuclides

0.01

Total cosmic and cosmogenic

0.39

Outdoors

0.07

Indoors

0.41

Total external terrestrial radiation

0.48

Uranium and thorium series

0.006

Radon (222Rn)

1.15

Thoron ( Rn)

0.1

Total inhalation exposure

1.26

40

K

0.17

Uranium and thorium series

0.12

Total ingestion exposure

0.29

0.2–1.0d

2.4

1.0–13220

Ingestion

Typical range

Total

0.3–1.0a

0.3–1.0b

0.2–10ca Range from sea level to high ground elevation.
b Depending on radionuclide composition of soil and building material.
c Depending on indoor accumulation of radon gas.
d Depending on radionuclide composition of foods and drinking water.

Table 13. Doses to members of the public due to the industrial release of NORM in the United Kingdom [W6]
Industry

Coal-fired power station

Discharge route

Pathway

Annual dose (µSv)
Critical group

General public
0.1

Atmospheric releases via stack

All

1.5

Building material made from ash

Radon inhalation

600

External

900

Oil and gas extraction

Authorized discharges to sea, and scales

Ingestion of seafood andexternal exposure due tofishing gear<30

Gas-fired power station

Atmospheric releases via stack

All

0.75

0.032

Steel production

Atmospheric releases via stack

All

<100

<2

Building material made from slag

Radon inhalation

550

External

800

Inhalation<1

Zircon sands

Atmospheric releases via stack

<1

340

UNSCEAR 2008 REPORT: VOLUME I

Table 14.

Worldwide uranium production [O16, O17, O21, W8]Country

Annual production (t)Cumulative production (t)

1998

1999

2000

2001

2002

2003

Total to 2003

Argentina

7

4

0

0

0

0

2 631

Australia

4 894

5 984

7 579

7 720

6 854

7 573

113 304

Belgium

15a

0

0

0

0

0

680a

Brazil

0

0

80

56

272

230

1 645

Bulgaria

0

0

0

0

0

0

16 735

Canada10 9228 214

10 68312 52211 607

10 455374 548

China

590b

700b

700b

700b

730b

730b

27 689b,cCongo, D.R.

0

0

0

0

0

0

25 600Czech Republic

610

612

507

456

465

452

108 649d

Finland

0

0

0

0

0

0

30

France

452

416

296

184

18e

9

75 965

Gabon

725

0

0

0

0

0

25 403Germany

30

29

28

27e

221e

150e

219 239

Hungary

10

10

10

10

10

4

21 080

India

207f

207f

207f

230f

230f

230f

7 963f

0

0

0

0

0

0

84

Kazakhstang

Japan

1 270

1 560

1 870

2 114

2 822

3 327

24 639Madagascar

0

0

0

0

0

0

785

Mexico

0

0

0

0

0

0

49

Mongolia

0

0

0

0

0

0

535

Namibia

2 780

2 690

2 715

2 239

2 333

2 037

78 794

Niger

3 714

2 907

2 911

2 919

3 080

3 157

91 186

Pakistan

23f

23f

23f

46f

38f

40f

931f

Poland

0

0

0

0

0

0

660

Portugal

19

10

14

4

0

0

3 680

Romania

132

89

86

85e

90e

90e

17 989eRussian Federationg

2 530

2 610f

2 760f

3 090f

2 850f

3 073f

32 136South Africa

965h

927h

838h

878h

828h

747h

157 618h

Spain

255

255

255

30e

37e

0

6 156

Sweden

0

0

0

0

0

0

91

Ukraineg

1 000

1 000

1 000

750f

800f

800f

9 900f,g

1 015

902

769

356 485

United States

1 810

1 773

1 522fUzbekistang

1 926

2 159

2 028

1 945

1 859

1 603

23 682

Zambia

0

0

0

0

0

0

102

World total

34 88632 17936 11237 02036 04235 4922 204 656i

a Produced from imported phosphates.
b Estimate for continental China.
c Production in China since 1990.
d Total production since 1946.
e Production resulting from decommissioning.
f Provisional data.
g Production since 1992.
h Uranium is by-product of gold mining.
i Includes 377 613 t of uranium produced in the former Soviet Union from 1945 to 1991, and 380 t produced in the former Yugoslavia before 1991.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

341

Table 15. Worldwide installed capacity for fuel cycle installations [I35]Country
Argentina

Conversion to UF6(t U/a)Enrichment(103 SWU/a)a

Fuel fabrication(t/a)b

62c

20c

150

Belgium

Reprocessing(t/a)b435

Brazil40

280

Canada12 500

2 700

China

1 500

1 000

400

France14 35010 800

1 585

1 800

650

Germany
India

1 700

594

Japan

1 050

1 689

Korea, Rep. of

120

800

Netherlands

2 500

Pakistan

5

20

Romania

110

Russian Federation

30 000

15 000

2 600

Spain

400

Sweden600

United Kingdom

6 000

2 300

1 680

United States

14 000

11 300

3 450

Total

78 45245 77518 143

400

2 700

4 920

a SWU: separative work unit. The SWU is a complex unit that is a function of the amount of uranium processed, the degree to which it is enriched and the level of depletion of
the remainder. It is indicative of the energy used in enrichment when feed and product quantities are expressed in kilograms.
b Tonnes of heavy metal.
c Design capacity.

Table 16. Electrical energy generated (GW)
Reactor type

1998

1999

2000

2001

2002

Average 1998–2002

AGR

6.52

6.39

6.14

6.26

0.33

5.13

BWR

64.06

66.86

67.32

69.09

64.77

66.42

FBR

0.28

0.43

0.41

0.44

0.43

0.40

GCRa

n.a.

n.a.

n.a.

n.a.

n.a.

n.a.

LWGR

7.29

8.14

8.30

7.87

8.06

7.93

HWR

11.35

12.17

12.56

13.71

13.66

12.69

PWR

154.27158.19164.59167.09170.91163.10

WWER

20.46

20.62

22.52

23.58

25.62

22.56

All

264.23272.80281.84288.05283.79278.23a n.a. = not available.

342

UNSCEAR 2008 REPORT: VOLUME I

Table 17.

Historical values for normalized releases of radionuclides from nuclear reactors (TBq/(GW a))

Period

PWRa

BWR

GCRb

HWR

LWGR

FBR

Totalc

4 800

5 000d150d

13 000460

5 000d

150d

3 300
1 200

Noble gases
1970–1974
1975–1979

530
430

44 000
8 800

580
3 200

1980–1984

220

2 200

2 300

210

5 500

150d

1985–1989

81

290

2 100

190

2 000

820

330

1990–1994

27

354

2 050

2 100

1 700

380

330

1995–1997

13

180

252

250

460

210

130

1998–2002

11

44

28

80

3 156

36

112

Tritium
1970–1974

5.4

1.8

9.9

680

26d

96d

448

1975–1979

7.8

3.4

7.6

540

26d

96d

38

1980–1984

5.9

3.4

5.4

670

26d

96d

44

690

26d

44

40

650

26d

49

36
16

1985–1989
1990–1994

2.7
2.3

2.1
0.94

8.1
4.7

1995–1997

2.4

0.86

3.9

330

26

49d

1998–2002

2.1

1.6

3.3

874

26d

49d

43

Iodine-131
1970–1974

0.003 30.15

0.001 4d0.001 40.08d

0.003 3d0.047

1975–1979

0.005

0.41

0.001 4d0.003 10.08d

0.005d0.12

1980–1984

0.001 80.093

0.001 40.000 20.08

0.001 8d0.03

1985–1989

0.000 90.001 80.001 40.000 20.014

0.000 9d0.002
0.000 7

1990–1994

0.000 33

0.000 80.001 40.000 40.007

0.000 3d

1995–1997

0.000 20.000 30.000 40.000 10.007

0.000 20.000 40.000 10.009 90.000 2d0.000 6

1998–2002

0.000 30.000 60.000 07
Carbon-14

1970–1974

0.22d

0.52d

0.22d

6.3d

1.3d

0.12d

0.71

6.3d

1.3d

0.12d

0.70

1975–1979

0.22

0.52

0.22d

1980–1984

0.35

0.33

0.35d

6.3

1.3d

0.12d

0.74

1985–1989

0.12

0.45

0.54

4.8

1.3

0.12d

0.53

1.6

1.3d

0.12d

0.44

1990–1994

0.22

0.51

1.4

1995–1997

—

—

—

—

—

—

—

1998–2002

0.22

0.53

1.3

1.2

1.3d

0.12d

0.39

Particulates
1970–1974

0.018

0.04

0.001d0.000 04d

0.015d0.000 2d0.019

1975–1979

0.002 2

0.053

0.001

0.000 04

0.015d0.000 2d

0.017

0.016

0.000 2d0.014

1980–1984

0.004 50.043

0.001 40.000 04

1985–1989

0.002

0.009 10.000 70.000 20.012

0.000 20.004

1990–1994

0.000 20.18

0.000 30.000 05

0.014

0.012

0.04

1995–1997

0.000 10.35

0.000 20.000 05

0.008

0.001

0.085

1998–2002

0.000 03

0.049

0.000 2d0.000 03

0.002 70.000 1

0.012

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Period

PWRa

GCRb

BWR

343

HWR

LWGR

FBR

Totalc

11d
11d

2.9d
2.9d

19

11d
11d

2.9d

38

0.4

41

1.8

48
59

Tritium (liquid)
1970–1974

11

3.9

9.9

180

1975–1979

38

1.4

25

350

1980–1984

27

2.1

96

290

42

1985–1989

25

0.78

120

380

1990–1994

22

0.94

220

490

1995–1997

19

0.87

280

340

11d
11d

1998–2002

20

1.8

402

817

0.78

1.7
1.7d

0.2d
0.18d
0.13d

0.38

0.004

0.095

38

Other (liquid)
1970–1974

0.2d

2

5.5

0.6

1975–1979

0.18

0.29

4.8

0.47

1980–1984

0.13

0.12

4.5

0.026

0.2d
0.18d
0.13d

1985–1989

0.056

0.036

1.2

0.03

0.045d

2.1
0.7

1990–1994

0.019

0.043

0.51

0.13

0.005

0.049

0.047

1995–1997

0.008

0.011

0.044

0.006

1998–2002

0.011

0.008

0.7
0.7d

0.260

0.002

0.023
0.023d0.03

0.04

a Includes all PWRs and WWERs.
b Includes GCRs and AGRs.
c Weighted by the fraction of energy generated by the reactor types.
d Estimated values.

Table 18. Estimated average annual collective doses due to effluents from nuclear power plants for the period 1998–2002
NuclidesQuantity

PWRa

BWR

GCRb

HWR

LWGR

FBR

Atmospheric releases
Noble gases

Tritium

I

131

ParticulatesC

14

Total release (PBq)

2.0 × 100

2.92 × 100

148 × 10–1

1.0 × 100

2.5 × 101

1.4 × 10–2

Collective dose (man Sv)

2.2 × 10–1

1.25 × 100

1.3 × 10–1

1.1 × 10–1

2.8 × 100

1.6 × 10–3

Total release (PBq)

3.9 × 10–1

1.1 × 10–1

1.7 × 10–2

1.1 × 101

2.0 × 10–1

2.0 × 10–2

Collective dose (man Sv)

8.2 × 10

2.2 × 10

3.6 × 10

2.3 × 10–1

4.3 × 10

4.1 × 10–2

Total release (PBq)

5.6 × 10–5

2.1 × 10–5

3.6 × 10–7

1.3 × 10–7

7.9 × 10–5

8.0 × 10–8

Collective dose (man Sv)

2.5 × 10

1.8 × 10

1.6 × 10

5.7 × 10–4

3.5 × 10

3.6 × 10–7

–1

–4

–1

–4

–2

–6

1

–6

Total release (PBq)

5.6 × 10–6

3.3 × 10–3

1.0 × 10–6

3.8 × 10–7

2.1 × 10–5

4.0 × 10–8

Collective dose (man Sv)

4.8 × 10–3

2.8 × 100

8.9 × 10–4

3.3 × 10–4

1.9 × 10–2

3.5 × 10–5

Total release (PBq)

4.1 × 10–2

3.5 × 10–2

7.0 × 10–3

1.5 × 10–2

1.0 × 10–2

4.8 × 10–5

Collective dose (man Sv)

1.0 × 100

9.5 × 100

1.8 × 10

4.1 × 100

2.8 × 10

1.3 × 10–2

10

Liquid releases
Tritium

OthersTotal release (PBq)

3.7 × 100

1.2 × 10–1

2.1 × 100

1.0 × 101

6.2 × 10–3

6.8 × 10–4

Collective dose (man Sv)

2.4 × 100

7.8 × 10–2

1.3 × 100

6.7 × 100

4.0 × 10–2

4.4 × 10–4

Total release (PBq)

2.0 × 10–3

5.3 × 10–4

3.6 × 10–6

3.3 × 10–3

9.2 × 10–6

9.2 × 10–6

Collective dose (man Sv)

6.7 × 10

1.8 × 10

1.2 × 10

1.1 × 10

5.2 × 10

3.0 × 10–3

–1

–1

–3

0

–3

Summaryc
Total collective dose (man Sv)

75

Total normalized collective dose due to airborne effluents (man Sv/(GW a))

0.22

Total normalized collective dose due to liquid effluents (man Sv/(GW a))

0.05

Total normalized collective dose due to releases from nuclear power plants (man Sv/(GW a))

0.27

a Includes all PWRs and WWERs.
b Includes GCRs and AGRs.
c Weighted by the fraction of energy generated by the reactor types.

344

UNSCEAR 2008 REPORT: VOLUME I

Table 19.

Releases from reactors no longer in commercial operation

Reactor

Shut down

Atmospheric releases in 2002 (GBq)
Noble gases

Tritium

Liquid releases in 2002 (GBq)

Iodine-131

Particulates

Tritium

Other nuclides

BWR
Big Rock Point-1

1997

0

9.5

0

0.001 6

0.15

0.12

Lacrosse-1

1987

0

1.1

0

0.000 53.1

0.60

Humboldt Bay

1976

0

0

0

0

0.15

0.008 5

Millstone-1

1998

0

33

0

0.000 500Browns Ferry-1

1985

35 855

1 463

3.4

0.075

0

0

Dresden-1

1978

2 184

1 346

0.047

0.22

1 253

0.34

79

0.67

PWR
Haddam Neck-1

1996

0

57

0

0.001 7Maine Yankee-1

1977

0

47

0

0.001 8

7.2

0.097

Rancho Seco-1

1989

0

52

0

0.000 4

427

0.050

San Onofre-1

1992

0

53

0

0.000 07

214

0.42

Three Mile Island-2

1979

0

34

0

0

0.020

0.000 4

Trojan-1

1992

852

278

0

0

3

0.084

Yankee Rowe-1

1991

0

0

0

0

0

0

Zion-1

1998

0

0

0

0.006 600.000 4

Zion-2

1998

0

0

0

0.006 600.000 4

Indian Point-1

1974

31 99716 2620.020

1.6

19 703

10

Table 20.

Collective doses due to fuel reprocessing
Airborne effluents
H

14

C

85

Kr

129

I

131

Total releases for the five-year period
1998–2002 (TBq)

2 001

44.16

2 160 300

0.14

0Collective dose conversion factor (man Sv/TBq)

0.002 10.27

0.000 007 4

44

0.3

7.4

1998–2002 collective dose (man Sv)

4.20

11.9

16.0

6.17

0

0.024

Quantity

3

Collective dose from all nuclides (man Sv)

38.31

Average annual collective dose (man Sv)

7.66

Normalized annual collective effective dose(man Sv/(GW a))

0.028

I

Cs

137

0.003 2Liquid effluents
Quantity
Total release for the five-year period
1998–2002 (TBq)
Collective dose conversion factor (man Sv/TBq)
1998–2002 collective dose (man Sv)

H

14

C

90

106

Ru

129

84 473

105.5

133.6

131.9

12.18

39.44

0.000 001 4

10.004 70.003 3

0.099

0.098

0.118

105.5

0.63

0.44

1.21

3.87

3

Sr

Collective dose from all nuclides (man Sv)

111.8

Average annual collective dose (man Sv)

22.35

Normalized annual collective effective dose(man Sv/(GW a))

0.081 4

I

Cs

137

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

345

Table 21. Spent fuel and arisings of low- and intermediate-level radioactive waste from nuclear power plants
TypeCountry

BWR

PHWR

PWR

WWER

Spent fuela

Reference plant

Spain

Cofrentes, S.M.Garoña

Switzerland

Leibstadt, Muehleberg

Argentina

AtuchaCanadaGentily-2, Point Lepreau

Korea, Rep.

Wolsong

Switzerland

Beznau, Goesgen

Korea, Rep.

Kori, Ulchin, Yongwang

Spain

Almaraz, Vandellós

Hungary

Paks

Low- and intermediate-level radioactive wastet/MW(e)m3/MW(e)Bq/MW(e)

0.02

0.10

1.32 × 1010

0.18

0.07

1.02 × 1010

0.02

0.04

5.17 × 109

0.04

0.26

n.a.b

a Tonnes of heavy metal.
b Not available.

Table 22. Normalized collective effective doses (man Sv/(GW a)) to local and regional population groups due to radionuclides
released in effluents of the nuclear fuel cycle
Source1970–1979

1980–1984

1985–1989

1990–1994

1995–1997

1998–2002

Mining

0.19

0.19

0.19

0.19

0.19

0.19

Milling

0.008

0.008

0.008

0.008

0.008

0.008

Mine and mill tailings(releases over five years)

0.04

0.04

0.04

0.04

0.04

0.04

Fuel fabrication

0.003

0.003

0.003

0.003

0.003

0.003

Airborne effluents

2.8

0.7

0.4

0.4

0.4

0.22

Liquid effluents

0.4

0.2

0.06

0.05

0.04

0.05

0.3

0.1

0.06

0.03

0.04

0.028

Reactor operation

Reprocessing
Airborne effluents

8.2

1.8

0.11

0.10

0.09

0.081

Transportation

Liquid effluents

<0.1

<0.1

<0.1

<0.1

<0.1

<0.1

Total (rounded)

12

3.1

0.97

0.92

0.91

0.72

Table 23. Annual collective doses (man Sv) due to transport for three types of spent fuel management in Germany [B7]
Population group

Comments

Nuclear reprocessing
centrea

Integrated back-end
conceptb

Alternative back-end
conceptc

Railway personnel handling radioactive
material in shunting yards

Average: 0.52 man Sv pertransport

4

10.8

3.9

Train drivers

No structure shielding
considered

0.026

0.068

0.026

Railway passengers

124 passengers per train

0.2

0.51

0.2

0.054

0.110

0.052

4.3

12

4.2

Inhabitants in the proximity ofshunting yards
Total

All population groupsa Transfer directly to nuclear reprocessing centres.
b Transfer using separate sites for interim storage of spent fuel, reprocessing and waste disposal.
c Direct disposal of spent fuel at the repository with no reprocessing.

346

Table 24.

UNSCEAR 2008 REPORT: VOLUME I

Number of packages and total activity of Suez Canal shipments from 1986 to 1992 [S1]Material

Package typea

Number of packages

Cumulative transport indexb

Total activity
(Bq)

Co
Cs
UF6
U3O8
UO2
252
Cf
85
Kr
3
H

B
A
B
A
B
A
A
A

36
30
920
1 852
145
1
17
19

122.4
59.1
1 265.5
9 651.4
397.6
0.5
13.0
27.0

5.06 × 1016
5.68 × 1014
9.70 × 1013
5.85 × 1014
2.32 × 1013
8.89 × 109
1.85 × 1013
1.02 × 1014

60

137

a Packages type refers to those described in reference [I33].
b The cumulative transport index refers to the sum of the transport index for all cargoes in the period 1986–1992.

Table 25.

Number of packages containing radioactive material carried by aircraft in the United Kingdom in 2001 [W3]Short-haula
Situation

Long-haulb

Passenger

Cargo

Passenger

Cargo

Total

Unitc

Loosed

Unitc

Loosed

Unitc

Unitc

Into United Kingdom

245

981

4 111

4

2 920

1 383

9 644

From United Kingdom

2 533

3 104

14 5402 746

13 19726 01562 135

Within United Kingdom

0

379

0

39

—

—

418

In transit

0

310

600

595

23

23

1 551

Number of consignments

25

1 042

2 530

337

2 840

90

6 864

a Short-haul: flights that take up to 4 h.
b Long-haul: flights that take over 4 h.
c Wide-body aircraft loaded with unit load devices (UDLs) to carry cargo.
d Narrow-bodied aircraft unable to take UDLs; cargo is loaded loose in the aircraft.

Table 26.

Collective doses due to transport of radioactive material by air in the United Kingdom in 2001 [W3]Exposed group

SubgroupType of flightCollective dose(man Sv/a)

Handlers

—

All

0.1

Aircrew

Cabin crew

Short-haul passengers
Long-haul passengers
Short-haul passengers
Short-haul cargo
Long-haul passengers
Long-haul cargo

0.13
3.8
0.002 5
0.024
0.13
0.56

Short-haul passengers
Long-haul passengers

0.43
2.8

Flight crew

Passengers——

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

347

Table 27. Estimated collective and individual doses to the public due to the normal transport of radioactive and nuclear
material [I5, W8]Country

PeriodFormer GDR

Product

1975–1984

Mode of transport

Collective dose(man Sv/a)

Individual dose(mSv/a)Spent fuel

Road

0.15

0.01

Road

0.01

Italy

1981

Fuel elements, PWR

United Kingdom

1981

Spent fuelRoad and rail

0.001

0.002

India

1982

Radioactive materials for medicine and industry

Road and air

0.1

b

Italy

1982

Medical use

Road

0.006

United Kingdom

1982

All

Mainly road

0.004

Germany, Fed. Rep.

1983

Fresh and spent fuel,UF6, ores, wastes

Rail

0.019

Austria

1984

192

Air, road, rail

0.23

Finland1982–1985

Spent fuelRoad and rail

(0.6–1.4) × 10–3

Turkey1984

192

Road and air

0.429

United States

1985

All from fuel cycle

All

19

0.02

All

All

100

0.02

Ir, 99Mo, 60Co, 123I, 131I, 133XeIr, 60Co, 131I, 99mTc, 137Cs, 241Am

United Kingdom

2001

Medical and industrial sources

Air

3.23a

United Kingdom

2003

Medical and industrial sources

Road

0.24

United Kingdom

2003

Spent fuelRoad and rail

0.003

0.04

a Doses to the public. Total annual collective dose, including passengers, crew and other workers, is 8 man Sv.
b Dose rate to the public: 1–55 µGy/h.

Table 28. Maximum annual doses (µSv) to members of the public due to the transport of various fuel cycle materials and byvarious modes [W14]Material

Road

Rail

Sea

Non-irradiated material

<4

<1

<20

Spent fuel

<4

<6

<1

Waste (low- and intermediate-level)<4<4High-level waste

20<1

MOX/plutonium<1

Table 29. Doses to the public from consumer products and miscellaneous items
Conservative estimates [W6]ItemRadioluminous wristwatch containing 147Pm

Estimated annual individual effective dose (µSv)
0.3Radioluminous wristwatch containing H

10

Smoke alarms

0.07

3

Uranium glazed wall tiles<1

Geological specimens

100

Photographic lensesa

200–300

Po in tobacco [C23, N9]210

a No longer in use.

10

348

UNSCEAR 2008 REPORT: VOLUME I

Table 30. Annual doses from by-products and radioactive materials in the United States [U35]Effective dose (mSv)

By-product
Automobile lock illuminators
Precision balances<0.01

Automobile shift quadrants
Marine compasses and navigational instruments
Thermostat dials and pointers
Self-luminous products
Timepieces, hands and dials

0.01–<0.1

Electron tubes
Gas and aerosol detectors

0.1–<1.0

Ionizing radiation measurement instruments
Spark gap irradiators

Effective dose (mSv)

Source material
Vacuum tubes
Electric lamps for illuminating purposes
Germicidal lamps, sunlamps and lamps for outdoor or industrial lighting<0.01

Personnel neutron dosimeters
Piezoelectric ceramic
Photographic film, negatives and prints
Uranium in fire detection units

0.01–<0.1

Glassware
Uranium shielding in shipping containers
Glazed ceramic tableware
Finished tungsten–thorium or magnesium–thorium alloy products or parts

0.1–<1.0

Uranium in counterweights
Thorium in finished optical lenses
Aircraft engine parts containing nickel–thorium alloy
Unrefined and unprocessed ore

1.0–<10

Incandescent gas mantles
Welding rods

≥10

Chemical mixtures, compounds, solutions or alloys
Rare earth metals and compounds, mixtures and products

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

349

Table 31. Summary of annual per caput doses due to peaceful uses of atomic energy (µSv)
Local component

Nuclear fuel cycle and energy generation

Other uses

Mining and milling

25

Fuel fabrication

0.2

Reactor operation

0.1

Reprocessing

2

Transport of radioactive waste<0.1

By-products

0.2
Regional component

Nuclear fuel cycle and energy generation

Fuel fabrication<0.01

Reactor operation<0.01

Reprocessing

0.02

Solid waste disposal and global component
Nuclear fuel cycle and energy generation

Globally dispersed radionuclides

0.2

Other uses

Disposal of radioactive waste<0.01

Table 32. Atmospheric nuclear tests at each test site [adapted from reference U3]
Test site

Number of tests

Yield (Mt)
FissionPartitioned fission yield (Mt)

FusionTotal

Local and regional

Troposphere

Stratosphere

20.72

0.15

0.66

11.4

China
Lop Nor

22

12.2

8.5
FranceAlgeria

4

0.073

0

0.073

0.036

0.035

0.001

Fangataufa

4

1.97

1.77

3.74

0.06

0.13

1.78

Mururoa

37

4.13

2.25

6.38

0.13

0.41

3.59

Total

45

6.17

4.02

10.19

0.23

0.58

5.37

0.05

0.049

0.000 7

United Kingdom
Monte Bello Island

3

0.1

0

0.1

Emu

2

0.018

0

0.018

0.009

0.009

0

Maralinga

7

0.062

0

0.062

0.023

0.038

0

Malden Island

3

0.69

0.53

1.22

0

0.56

0.13

Christmas Island

6

3.35

3.3

6.65

0

1.09

2.26

Total

21

4.22

3.83

8.05

0.08

1.75

2.39

United States
New Mexico

1

0.021

0

0.021

0.011

0.01

0

Nevada

86

1.05

0

1.05

0.28

0.77

0.004

Bikini

23

42.2

34.6

76.8

20.3

1.07

20.8

Enewetak

42

15.5

16.1

31.7

7.63

2.02

5.85

Pacific

4

0.102

0

0.102

0.025

0.027

0.05

Atlantic

30.004 500.004 5

0

0

0.005

Johnston Island

12

10.5

10.3

20.8

0

0.71

9.76

Christmas Island

24

12.1

11.2

23.3

0

3.62

8.45

Total

195

81.5

72.2

153.8

28.2

8.23

44.9

350

UNSCEAR 2008 REPORT: VOLUME I

Test site

Number of tests

Yield (Mt)
FissionFusionPartitioned fission yield (Mt)
Total

Local and regional

Troposphere

Stratosphere

Former Soviet Union
Semipalatinsk

116

3.74

2.85

6.59

0.097

1.23

2.41

Novaya Zemlya

91

80.8

158.8

239.6

0.036

2.93

77.8

Totsk, Aralsk

2

0.04

0

0.04

0.000 15

0.037

0.003

Kapustin Yar

10

0.68

0.3

0.98

0

0.078

0.61

Total

219

85.3

162

247.2

0.13

4.28

80.8

29

15

145

All countries
Total

502

189

251

440

Table 33. Number and total yield of weapons tests and other events by country [B16, U3]
Type

UnitedStatesFormerSoviet Union

UnitedKingdom

France

China

India

Pakistan

Democratic
People’sRepublic of
Korea

Total

Atmospheric nuclear tests
Airburst

13

8

9

0

0

0

0

0

30

Air drop52

174

Balloon

24

0

1

3

16

0

0

0

246

1

34

0

0

0

0

59

Tower/surface

66

34

10

4

6

0

0

0

120

Barge
Underwater

35

0

0

4

0

0

0

0

39

5

3

0

0

0

0

0

0

8

Total

195

219

21

45

22

0

0

0

502

Fission (Mt)

82

85

4

6

12

0

0

0

189

Fusion (Mt)

72

162

4

4

9

0

0

0

251

6

6

1

1 877

Underground nuclear tests
Number of tests

908

750

24

160

22

Yield (Mt)

46

38

2

3

1

90

Total atmospheric and underground tests
Number of tests

1 103

969

45

205

44

Yield (Mt)

200

285

10

13

22

6

6

1

2 379
530

Other events with nuclear weapons
Safety trials

22

0

12

5

0

0

0

0

39

Operational

2

0

0

0

0

0

0

0

2

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

351

Table 34. Radionuclides produced and globally dispersed in atmospheric nuclear tests [U3]Radionuclide

Half-life

Global release (PBq)

H

12.33 a186 000

14

C

5 730 a213

Mn

312.3 d3 980

Fe2.73 a1 530

Sr50.53 d117 000

Sr

28.78 a622

Y

58.51 d120 000

Zr

64.02 d148 000

Ru

39.26 d247 000

Ru

373.6 d12 200Sb

2.76 a741

I

8.02 d675 000

3

54

55

89
90

91
95

103
106
125

131

Ba

12.75 d759 000

Ce

32.5 d263 000

Ce

284.9 d30 700Cs

30.07 a948

Pu

24 110 a

6.52

Pu

6 563 a4.35

Pu

14.35 a

142

140
141
144

137

239
240
241

Table 35. Latitudinal distribution of radionuclides from atmospheric nuclear tests based on 90Sr measurements [U3]Latitude band(º)

Population distribution(%)

Integrated deposition of90
Sr (PBq)

Deposition in band(%)

Deposition density perunit deposition((Bq/m2)/PBq)

Latitudinal value relative
to hemispheric value

Northern hemisphere
80–90

0

1

0.2

0.56

0.12

70–80

0

7.9

1.7

1.48

0.32

60–70

0.4

32.9

7.1

3.78

0.81

50–60

13.7

73.9

16.1

6.27

1.35

40–50

15.5

101.6

22.1

7.01

1.51

30–40

20.4

85.3

18.5

5.09

1.09

20–30

32.7

71.2

15.5

3.85

0.83

10–20

11

50.9

11.1

2.58

0.56

0–10

6.3

35.7

7.8

1.76

0.38

0.53

0.14

Southern hemisphere
80–90

0

0.3

0.2

70–80

0

2.5

1.7

1.5

0.4

60–70

0

6.7

4.6

2.46

0.66

50–60

0.5

12.1

8.4

3.28

0.88

40–50

0.9

28.1

19.5

6.19

1.65

30–40

13

27.6

19.1

5.26

1.4

20–30

14.9

28.1

19.5

4.85

1.29

10–20

16.7

17.8

12.3

2.89

0.77

0–10

54

21

14.6

3.3

0.88

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UNSCEAR 2008 REPORT: VOLUME I

Table 36. Estimated average effective doses (μSv) due to global fallout received by the world population [B46, U3, U6]
Radionuclide

Received before 2000

To be received
2000–2100

To be received
beyond 2100
All pathways

External irradiation

Inhalation

Ingestion

All pathways

All pathways

H

—

—

24

24

0.1

14

C

—

—

144

144

120

Mn

19

0.1

19

—

Fe

—

0.01

6.6

6.6

—

Sr

—

2.6

1.9

4.5

—

Sr

—

9.2

97

106

8.6

Y

—

4.1

—

4.1

—

Zr

81

2.9

—

84

—

Ru

12

0.9

—

13

—

3

54

55

89
90

91
95

103

Ru

25

35

—

60

—

Sb

12

0.1

—

12

0.003

I

1.6

2.6

64

68

—

Ba

27

0.4

0.5

28

—

Ce

1.1

0.8

—

1.9

—

Ce

7.9

52

—

60

—

Cs

166

0.3

154

320

124a

Pu

—

20

—

20

—

106
125

131
140
141
144

137

239

Pu

—

13

—

13

—

241

Pu

—

5

—

5

—

Total

353

149

492

994

253

240

2 230

0.02

13

2 243

a 114 μSv from external irradiation and 10 μSv from internal irradiation.

Table 37. Estimated effective doses for several regions of Maralinga and Emu [H7]Zone

Annual effective dose (mSv)

Principal pathway

Principal nuclide

Taranaki North Plume, 241Am A contour
10-year-old childPu

5

Inhalation239

10-year-old child4

Inhalation239

10-year-old child23

Inhalation; external gamma

234/238

10-year-old child3

Inhalation; external gamma

234/238

3-month-old infant

2

Soil ingestion; inhalation239

10-year-old child5

Inhalation239

10-year-old child9

Inhalation

239

Taranaki Northwest, 241Am A contour
Pu

Kuli 238U D contour
U

Kuli 238U A contour

Northeast One Tree: 2 kBq/m of
2

U

Cs

137

Pu

TM100, 241Am A contour

Emu—Totem II,

Pu

Am A contour

241

Pu

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Zone

Annual effective dose (mSv)

Principal pathway

353

Principal nuclide

Emu—Totem I, 137Cs A contour
Pu

10-year-old child

0.5

Inhalation; ingestion239

10-year-old child

470

Inhalation239

10-year-old child

31

Inhalation

239

Inner Taranaki
Pu

Emu—centre of Totem II
Pu

Table 38. Median (and mean) activitiesa of 137Cs, 90Sr, 239+240Pu and 241Am per unit dry weight of soil on Bikini Island (Bq/g) [I9]Soil depth (cm)

Cs

Sr

137

90

Pu

Am

239+240241

Interior of island
0–5

2.3 (3.0)

1.7 (2.1)

0.32 (0.42)

0.26 (0.30)

0–40

0.70 (0.91)

1.1 (1.5)

0.17 (0.21)

0.11 (0.14)

Village area
0–5

1.2 (2.0)

1.0 (2.0)

0.20 (0.40)

0.11 (0.22)

0–40

0.67 (1.1)

1.6 (1.5)

0.24 (0.29)

0.13 (0.17)

a Decay corrected to 1999. The numbers in parentheses are the arithmetic means.

Table 39. Radionuclide and pathway contributions to hypothetical doses on Bikini Island assuming a local diet [I9]Exposure pathway

Annual dose (mSv)

External gamma
Ingestion

0.4
Cs

14.6

137

Sr

0.15

90

239+240Pu

0.001 9

Am

0.001

241

Inhalation

239+2400.000 74Am

Pu

0.000 49

241

Total (rounded)

15

Table 40. Absorbed dose rates in air in settlements outside and inside the Semipalatinsk nuclear test site [I10]
Location

Dose rate at 1 m above ground (µGy/h)
Outside the nuclear test site

Entire perimeter (over 500 measurements)
Dolok

0.06–0.17
0.07

Sarzhal and surrounding pasture

0.08–0.09

Kainar

0.08–0.11

Akzhar and surrounding pasture
Dolon
Other settlements

0.08
0.09
0.07–0.14

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UNSCEAR 2008 REPORT: VOLUME I

Location

Dose rate at 1 m above ground (µGy/h)
Inside the nuclear test site

Lake Balapan

0.1–33Ground Zero
1 km from centre

0.1

Within 1 km

0.1–17

South-eastern plume0.09

Polygon farm

0.1

Beriozka State Farm

0.2

Sary-Uzen

0.5

Lake Tel’kem-2

0.2–1.0

Table 41. Estimated annual effective doses to persons living around the Semipalatinsk test site, to visitors to Lake Balapan
and Ground Zero, and to potential future permanent inhabitants
Pathway

Annual dose (mSv)
Outside test area

External gamma
Inhalation

Other settlements

Frequent visitors

Future permanent inhabitants

0.01

0.01

10

90

0.05

1.2

0.2

3.5

0.02

0.4

0.007

Pu

0.04

Am

0.004

239+240
241

Ingestion

Dolon

Pu

238

Inside test area

0.01

Cs

0.03

0.03

3

30

Sr

0.02

0.02

0.06

10

Pu

0.004

0.07

0.6

Pu

0.02

0.2

2

241

Am

0.002

0.02

0.2

Total (rounded)

0.14

14

140

137
90

238

239+240

0.001

0.06

Table 42. Summary of residual radionuclide inventory on the Nevada Test Site as of January 1996 [U25]Source of radioactivity

Type of area

Environmental media

Major known isotopes or wastes

Depth

Activity (Bq)Atmospheric andtower tests

Above-ground nuclear weapons
proving area

Surficial soil and test
structures

Am, Cs, Co, Pu, Eu, Sr

At land surface~7.4 × 1010

Safety trials

Above-ground experimental area

Surficial soil

Am, Cs, Co, Pu, Sr<0.9 m~1.3 × 1012

Nuclear rocket
development area

Nuclear rocket, motor, reactor andfurnace testing area

Surficial soil

Cs, Sr<3 m~3.7 × 1010

Shallow borehole tests

Underground nuclear testing area

Soils and alluvium

Am, Cs, Co, Eu, Pu, Sr<61 m~7.4 × 1013

Shallow land disposal

Waste disposal landfill

Soils and alluvium

Dry packaged low-level and mixed
wastes<9 m~1.85 × 1015

Crater disposal

Test-induced subsidence craterwith sidewalls, cover and drainage

Soils and alluvium

Bulk contaminated soil andequipment<30 m~4.6 × 1013

Greater confinement
disposalMonitored underground waste
disposal borehole

Soils and alluvium

Am, tritium

37 m~3.4 × 1017(~300 × m3)

Deep undergroundtests

Underground nuclear testing area

Soils, alluvium andconsolidated rockTritium, fission and activation
products

Typically less than
640 m but may be
deeper

>1.1 × 1019

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

355

Table 43. Properties of uranium isotopes 238U, 235U and 234U and their relative abundance in natural and depleted uranium [I7]
IsotopeAverage energy per transformation(MeV/Bq)

Half-life

Natural uranium

Depleted uranium

Specific
activity

Relative isotopic abundance(%)

Specific
activity

Relative isotopic abundance
(%)

Alpha

Beta

Gamma

(a)

(Bq/mg U)

By mass

By activity(Bq/mg U)

By mass

By activity

U

4.26

0.01

0.001

4.51 × 109

12.44

99.28

48.2

12.44

99.8

87.5

U

4.47

0.04

0.154

7.1 × 108

0.6

0.72

2.2

0.16

0.2

1.1

U

4.84

0.001 30.002

2.47 × 105

12.44

0.005 5

49.5

1.61

0.000 7

11.4

238
235
234

Table 44. Estimated amount of depleted uranium used in armed conflict
Conflict

Total DU (t)Gulf War I (1991)

286

Bosnia and Herzegovina (1994–1995)

3

Kosovo (1999)

10

Serbia and Montenegro (1999)

0.7

Table 45. Lands contaminated with radionuclides at enterprises of Minatom of Russia
As of 1 January 2000 [L2]Enterprise

Area (km2)

Area (km2) with exposure rates of greater than 2 µGy/hPriargun Mining and Chemical Association

8.53

—

Mining and Metallurgical Plant (Lermontov)

1.34

1.03

Machine-building Plant (Elektrostal)

0.26

0.261

Novosibirsk Plant of Chemical Concentrates

0.15

0.14

Moscow Plant of Polymetals

0.016

0.001

Chepetsk Mechanical Plant (Glazov)

1.35

0.062

Zabaikalski Mining and Enrichment Combine

0.04

Mayak Production

452.16

65.7

Mining and Chemical Complex (Zheleznogorsk)

4.7

0.203

Siberian Chemical Complex (Seversk)

10.39

4.191

Kirovo-Chepetsk Chemical Complex

0.7

All-Russian Research Institute of Technical Physics (Snezhinsk)

0.13

Research Institute of Atomic Reactors (Dimitrovgrad)

0.39

Institute of Physics and Power Engineering (Obninsk)
Total

0.01
0.081
0.001

480.32

71.68

Table 46. Number of particles retrieved and their average activity close to the Dounreay site as of May 2007 [D5]Particle location

Number of particles foundAverage particle activity (Bq)

Marine sediment930

1.4 × 106

Dounreay offshore

248

5.5 × 106

Sandside Beach94

7.3 × 104

Dunnet Beach

1

8.9 × 103

Murkle Beach

1

Dounreay site (estimate)86

1.3 × 104
n.a.aa Not available.

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Table 47.

UNSCEAR 2008 REPORT: VOLUME I

Maximum total annual individual doses estimated for selected population groups close to the Kara Sea [I11]
Scenarios

Annual doses (µSv)
Seafood consumers – Groups (a) and (c)Military personnel – Group (b)Best estimate scenario

<0.1

700

Plausible worst scenario

<1

4 000

Climate change scenario

0.3

3 000

Groups:(a) Living in Ob and Yenisei estuaries and on Taimyr and Yamal peninsulas; habits typical of subsistence fishing communities in Arctic.
(b) Hypothetical group of military personnel patrolling, for 100 hours in a year, foreshores of fjords containing dumped radioactive material.
(c) Seafood consumers representative of northern Russian population situated on Kola Peninsula.

Table 48.

Practices for which UNSCEAR evaluates occupational exposure
Category of practice

Practice

Exposure to natural sources of radiation

Civilian aviation
Coal mining
Other mineral mining
Oil and natural gas industries
Workplace exposure to radon other than in mines

Nuclear fuel cycle

Uranium mining
Uranium milling
Uranium enrichment and conversion
Fuel fabrication
Reactor operation
Decommissioning
Fuel reprocessing
Research in the nuclear fuel cycle
Waste management

Medical uses

Diagnostic radiology
Dental radiology
Nuclear medicine
Radiotherapy
All other medical uses

Industrial uses

Industrial irradiation
Industrial radiography
Luminizing
Radioisotope production
Well logging
Accelerator operation
All other industrial uses

Miscellaneous

Educational establishments
Veterinary medicine
Other occupations

Military activities

All military activities

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

357

Table 49. Occupational exposure of aircrew
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures and the literature [S38]Country

Number of workers

Collective dose(man Sv)

Average effective dose
(mSv)

100

0.6

1.36

CanadaMaximum effective dose(mSv)

Czech Republic

1 195

1.5

1.28

3.5

Denmark

3 990

6.8

1.7

—

Finland

2 520

4.2

1.7

—

Germany

31 000

60.0

2.0

6.5

Lithuania

160

0.2

1.2

Netherlands

12 500

17.0

1.3

<6

United Kingdom

40 000

80.0

2.0

—

United States

150 000a0.2–5.0ba Data from reference [U27].
b Data from references [W2, W16].

Table 50. Estimated effective doses for specific flight routes leaving Frankfurt, Germany
Destination

Range of the dosea (µSv)

Gran Canaria

10–18

Johannesburg

18–30

New York

32–75

Rio de Janeiro

17–28

Rome

3–6

San Francisco

45–110Singapore

28–50

a A range of values is given because of differences in flying altitude and variations in the intensity of the cosmic ray flux due to varying solar activity.

Table 51.

Dose equivalent rate and mission dose equivalent in crewed space missions [R7, R8]

Mission

Inclination(grad)

Altitude(km)

Mission duration(h)

Dose equivalent rate
(mSv/d)Mission dose equivalent
(mSv)

SL1

57

250

247.5

0.46

4.7

D-1

57

324

168

0.48

3.3

IML-1

57

348

194

0.38

3.0

MIR92

51.5

400

190

0.64

5.1

D-2

28.5

296

240

0.19

1.9

IM-L2

28.5

296

353

0.26

3.8

Euromir ’94

51.5

400

756

~0.86

~27

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Table 52. Annual doses to underground coal miners in China [C12]Type of coal mine

Average annual effective dose
(mSv)

Collective dose(man Sv)

Large-sized

0.28

280

Medium-sized

0.55

550

Small-sized

3.3

13 200Bone-coal

10.9

545

Average

2.4

14 600

Table 53. Occupational exposure in underground gold mines in South Africa [W17]Year

Average annual dose(mSv)

Number of workers

Number of workers receiving
doses of >20 mSv

1997

6.3

258 080

12 9041998

4.9

232 500

2 325

1999

5.4

175 333

5 260

2000

7

123 333

3 700

Table 54. Estimated external doses for workers in Abu-Tartor phosphate mine tunnels
Estimates based on individual and workplace monitoring using TLDs [K11]
Type of worker

Mean

SEa

SDb

Minimum

Maximum

Numberc

Effective dose rate estimated using individual monitoring (mSv/a)Mine workers

15.55

2.73

12.20

6.78

53.52

20

Mine maintenance workers

10.25

0.97

3.64

5.90

18.23

14

Ore crushing and transport workers

11.34

1.03

1.78

9.83

13.31

3

Beneficiation factory workers

10.95

0.35

0.79

10.09

12.11

5

Ore drying and storage workers

10.21

0.15

0.26

9.97

10.49

3

Average

11.66

—

—

—

—

—

Effective dose rate estimated using workplace monitoring (mSv/a)
Mine

8.51

0.60

3.36

2.19

17.09

31

Ore crushing

10.06

0.31

0.70

8.94

10.81

5

Processing facility

8.35

0.52

1.08

6.82

9.07

4

Average

8.97

—

—

—

—

—

a Standard error of the mean.
b Standard deviation.
c Number of measurements.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

359

Table 55. Doses received by workers in zircon milling plants [I41]
Location

Annual effective dose (mSv)
Gamma radiation

Dust inhalationa

Total— New autogenous mill, dust extraction, enclosed bagging, goodindustrial hygiene

0.4

0.27

0.67

— Old roller mill, no special dust extraction during bagging

0.3

0.73

1.03

— Old ball mill, no special dust extraction

0.1

0.56

0.66

— Old ball mill, semi-automatic bagging, no special dust
extraction

0.4

0.56

0.96

Australia

Netherlands

0.8South Africa— Mill operators (mill areas)

0.102

0.163

0.265

— Mill operators and maintenance personnel (warehouse)

0.238

0.046

0.284

— Maintenance personnel (mill areas)

0.067

0.042

0.109

— General workers (mill areas)

0.096

0.06

0.156

— General workers (warehouse)

0.210

0.04

0.250

— Mill attendant

0.275

0.165

0.44

— Shift supervisor

0.18

0.094

0.274

— Cleaner

0.2

0.134

0.33

0.16

0

0.16

South Africac

South Africa— Wet mill operator
United Kingdom

0.5

United States: bagger operator (respiratory protection mandatory)— Without respiratory protectionb

0.2

1.9

2.1

— With respiratory protection<1

a Except where otherwise stated, values are based on the assumption that no respiratory protection was used.
b Doses calculated from values of gamma exposure, airborne dust activity concentration and occupancy period using the inhalation dose coefficients for an AMAD of 5 µm.
c Maximum values measured after implementation of the following dose reduction measures: reduction of dust generation through revised engineering practices, reduction ofstockpile quantities and thus of gamma exposures, and reduction of surface contamination by continuous cleaning practices [I41].

Table 56. Occupational exposure in Germany due to radon inhalation in workplaces other than mines
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
Workplace

Spas

Waterworks
Tourist caves andvisitor mines

PeriodMonitored workers(103)

Measurably exposed
workers(103)

Annual collective
effective dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

Measurably exposed
workers

1995–1999

0.002

0.002

0.01

4.77

4.77

2000–2002

0.004

0.002

0.01

4.09

4.47

1995–1999

0.128

0.075

0.24

1.85

3.12

2000–2002

0.081

0.047

0.11

1.39

2.50

1995–1999

0.135

0.101

0.31

2.26

3.01

2000–2002

0.131

0.087

0.23

1.76

2.63

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Table 57. Estimated worldwide levels of annual exposure due to natural sources of radiation for the period 1995–2002
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures and the literature
Workplace

Monitored workers(103)

Annual collective effective dose(man Sv)

Average annual effective dose
(mSv)

6 900

16 5602.4

Coal mining
Other mining (excluding uranium mining)

4 600

13 800

3.0

Workplaces other than mines

1 250

6 000

4.8

Aircrew

300

900

3.0

Total

13 05037 260

2.9

Table 58. Estimated worldwide levels of annual exposure in uranium mining
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodAnnual amount ofore extracted(kt U)

Collective dose
per unit mass(man Sv/kt)

Monitored workers(103)

Annual collective dose(man Sv)

Annual effective dose(mSv)

1975–1979

52

26

240

1 300

5.5

1980–1984

64

23

310

1 600

5.1

1985–1989

59

20

260

1 100

4.4

1990–1994

39

8

69

310

4.5

1995–1999

34

2

22

85

3.9

2000–2002

34

1

12

22

1.9

Table 59. Exposure to workers from underground and above-ground uranium mining
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
Country/type ofuranium mine

PeriodMonitored workers
(103)
Total

Measurably exposed

Annual collective
effective dose(man Sv)

Average annual effective dose
(mSv)
Total

Measurably exposed

CanadaAbove-ground

Underground

1995–1999

1.30

0.64

0.61

0.44

0.97

2000–2002

1.06

0.50

0.48

0.45

0.99

1995–1999

1.03

0.77

3.05

3.13

4.13

2000–2002

0.65

0.48

1.46

0.95

2.00

Germany
Above-ground

Undergrounda

1995–1999

0.73

0.73

0.96

1.32

1.32

2000–2002

0.53

0.53

0.15

0.28

0.28

1995–1999

1.04

0.90

2.10

2.01

2.46

2000–2002

0.88

0.88

0.94

1.07

1.07

a Extracted only in connection with decommissioning of mining facilities.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

361

Table 60. Contribution of internal and external exposure to the effective dose due to uranium mining
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
Type of mine

Percentage contribution of internal and external exposure to the effective dose
Dose less than 1 mSv
Radon progeny

Ore dust

Dose more than 1 mSv
External exposure

Radon progeny

Ore dust

External exposure

CanadaUnderground

82%

—

—

62%

—

—

Above-ground

74%

—

—

41%

—

—

72%

20%

8%

Czech Republic
Underground

45%

39%

16%
Germany

Underground

36%

32%

32%

37%

53%

10%

Above-ground

23%

34%

43%

51%

40%

9%

Table 61. Effective dose to workers in above-ground and underground uranium mines
Reports on Canadian occupational radiation exposure [H9, H10, H11, H12, H13, H14]
Year

Personnel
Number ofworkers

Averageeffective
dose (mSv)

Maintenance
Per centradon
progeny

Number ofworkers

Average effective
dose
(mSv)

Miner
Per centradon
progeny

Number ofworkers

Average effective
dose
(mSv)

Per centradon
progeny

77

154

1.24

59

Above-ground uranium mines
1995

50

0.59

80

211

1.31

1996

61

0.75

71

247

1.33

73

214

1.30

51

1997

102

0.32

62

202

0.55

66

244

0.94

23

1998

126

0.37

79

176

0.36

93

96

0.80

50

1999

177

0.35

62

219

0.40

73

74

0.83

11

2000

186

0.64

74

194

0.64

45

89

1.35

52

2001

208

0.58

68

189

0.57

51

47

2.15

55

66

386

10.90

63

Underground uranium mines
1995

368

0.98

64

109

5.37

1996

387

0.82

60

101

3.85

62

469

9.62

58

1997

476

0.69

39

103

1.48

64

354

5.53

39

1998

346

0.55

75

139

0.93

87

362

1.97

78

1999

155

1.01

71

204

0.90

79

341

2.60

63

2000

111

0.88

53

194

0.71

70

284

2.57

40

2001

73

0.48

54

115

0.46

64

161

2.29

32

362

UNSCEAR 2008 REPORT: VOLUME I

Table 62. Estimated worldwide levels of exposure due to uranium milling
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodAnnual amount of ore
refined (kt U)

Equivalent amount ofenergy (GW a)

Monitored workers(103)Collective effective dose(man Sv)

Average annual effective
dose (mSv)

1975–1979

53

240

12

124

10.1

1980–1984

64

290

23

117

5.1

1985–1989

58

260

18

116

6.3

1990–1994

39

180

6

20

3.3

1995–1999

34

155

3

4

1.6

2000–2002

34

155

3

3

1.1

Table 63. Contribution of internal and external exposure to the effective dose due to uranium milling
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
Uranium milling

Percentage contribution of internal and external exposure to the effective dose
Dose less than 5 mSv
Radon progeny

Canada70

Germany35

Ore dust

Dose more than 5 mSv
External exposure

Radon progeny

Ore dust

External exposure

72

9

50
35

30

19

Table 64. Estimated worldwide levels of exposure due to uranium enrichment
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)Collective effective dose(man Sv)

Average effective dose(mSv)

1975–1979

11.0

5.30

0.46

1980–1984

4.3

0.78

0.18

1985–1989

5.0

0.43

0.08

1990–1994

12.6

1.28

0.10

1995–1999

17.2

1.34

0.08

2000–2002

18.2

1.70

0.09

Table 65. Estimated worldwide levels of exposure due to fuel fabrication
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Annual collective effective dose(man Sv)

Average annual effective dose(mSv)

1975–1979

20

36

1.8

1980–1984

21

21

1.0

1985–1989

28

22

0.8

1990–1994

21

22

1.0

1995–1999

22

30

1.4

2000–2002

20

31

1.6

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

363

Table 66. Summary of worldwide exposures due to reactor operations
Period

PWR

BWR

HWR

GCR

LWGR

All

Average number of monitored workers (10 )31975–1979

63

59

7

13

5

147

1980–1984

140

102

14

25

10

291

13

1985–1989

230

139

18

31

1990–1994

310

160

20

30

431
530

1995–1999

265

144

18

21

448

2000–2002

283

113

23

18

437

1975–1979

3.5

4.7

4.8

2.8

1980–1984

3.1

4.5

3.2

1.4

6.4

3.6

1985–1989

2.2

2.4

3.4

0.8

13.2

2.5

1990–1994

1.3

1.6

1.7

0.5

1.4

1995–1999

1.9

1.7

1.6

0.3

1.5

2000–2002

1.7

1.4

1.6

0.2

1.0

Average annual effective dose to monitored workers (mSv)
6.6

4.1

Average annual collective effective dose (man Sv)
1975–1979

220

279

32

36

36

603

1980–1984

450

454

46

34

62

1 046

1985–1989

500

331

60

24

173

1 088

1990–1994

415

240

35

16

190

896

1995–1999

506

237

29

7

779

2000–2002

415

160

38

4

617

1975–1979

8.1

18.3

Normalized collective effective dose per unit electrical energy (man Sv/(GW a))
11.0

6.6

8.2

10.9

1980–1984

8.0

18.0

8.0

5.8

8.3

10.4

1985–1989

4.3

7.9

6.2

3.2

16.7

5.7

1990–1994

2.8

4.8

3.0

2.0

20.3

3.9

1995–1999

3.0

3.8

2.4

0.7

2.5

2000–2002

2.2

2.4

2.9

2.6

2.5

1975–1979

2.8

5.5

2.6

0.9

3.0

3.1

1980–1984

3.3

7.0

2.4

0.8

3.8

3.7

1985–1989

2.3

4.0

2.3

0.5

8.7

2.8

1990–1994

1.7

2.7

1.1

0.4

9.4

1995–1999

2.0

2.6

1.2

0.2

1.5

2000–2002

1.6

1.8

1.0

0.2

1.1

Normalized collective effective dose per reactor (man Sv per reactor)

2.1

364

Table 67. Annual occupational doses for reactor operation by job category
From ISOE and Canadian National Dose Registry
Refuellinga

Maintenanceb

Inspectionc

Servicingd

Other

Annual collective effective dose

Annual collective effective dose

Annual collective effective dose

Annual collective effective dose

Annual collective effective doseCountry

Averagenumber ofreactors
over theperiodTotal(man mSv)

Averageper reactor(man mSv)

Averagenumber ofreactors
over theperiodTotal(man mSv)

Averageper reactor(man mSv)

Averagenumber ofreactors
over theperiodTotal(man mSv)

Averageper reactor(man mSv)

Averagenumber ofreactors
over theperiodTotal(man mSv)

Averageper reactor(man mSv)

Averagenumber ofreactors
over theperiodTotal(man mSv)

Averageper reactor(man mSv)

1995–1999
BWR
4.4

284.5

64.7

5.8

1 851.7

319.3

4.6

382.3

83.1

5.4

1 187.1

219.8

3.4

269.3

79.2

Mexico

1.2

339.1

282.6

1.2

2 492.9

2 077.5

1.0

107.4

107.4

1.2

466.9

389.1

0.8

131.9

164.9

Spain

1.8

162.3

90.1

2.0

1 266.8

633.4

2.0

227.1

113.6

2.0

971.5

485.8

2.0

1 058.9

529.5

Sweden

8.6

337.7

39.3

9.0

8 111.6

901.3

8.8

1 013.5

115.2

9.0

4 502.6

500.3

6.6

2 416.0

366.1

Switzerland

2.0

150.6

75.3

2.0

386.0

193.0

1.6

101.2

63.2

2.0

282.1

141.1

1.8

739.1

410.6

6 146

87.8

14

16 986243

14

21 143

302

125.2

78.3

1.6

1 210.9

756.83

1.6

3 635.3

2 272.1

0.2

6.4

32.2

0.3

43.9

175.4

0.2

16.4

81.8
91.1

HWR
Canada

14

2 008

28.6

14

30 834

440

14
LWGR

Lithuania

2.0

451.9

226.0

1.6

1 267.9

792.5

1.6
PWR

Armenia
Belgium

6.2

417.8

67.4

6.2

1 809.7

291.9

5.8

288.1

49.7

6.0

763.1

127.2

4.6

419.1

China

2.0

136.0

68.0

2.0

444.2

222.1

1.0

15.8

15.8

2.0

269.8

134.9

1.4

134.3

95.9

Finland

2.0

69.5

34.7

2.0

441.8

220.9

2.0

359.8

179.9

0.8

87.7

109.6

France46.8

5 295.1

113.1

46.8

25 087.7

536.1

46.0

3 257.6

70.8

46.6

13 398.6

287.5

46.0

5 118.3

111.3

Germany

9.2

579.2

63.0

10.6

4 872.9

459.7

6.2

810.2

130.7

8.8

2 552.6

290.1

4.4

688.2

156.4

Hungary

4.0

234.0

58.5

4.0

1 139.9

285.0

4.0

704.2

176.0

4.0

354.5

88.6

Netherlands

1.0

95.9

95.9

1.0

252.9

252.9

1.0

135.9

135.9

1.0

247.7

247.7

1.0

45.0

45.0

South Africa

1.6

111.2

69.5

1.6

522.9

326.8

1.4

57.1

40.8

1.6

248.4

155.3

0.8

19.7

24.6

Slovenia

1.0

179.5

179.5

1.0

685.3

685.3

0.8

21.1

26.4

1.0

205.7

205.7

0.8

11.1

13.9

Spain

5.6

723.9

129.3

5.6

2 179.6

389.2

5.6

326.6

58.3

5.6

1 353.1

241.6

0.8

128.3

160.4

Sweden

3.0

201.5

67.2

3.0

980.7

326.9

2.2

37.2

16.9

3.0

290.2

96.7

0.4

0.6

1.6

UNSCEAR 2008 REPORT: VOLUME I

GermanySwitzerland

2.6

258.2

99.3

2.6

469.7

180.7

1.6

61.6

38.5

2.6

282.0

108.5

1.8

200.8

111.6

UnitedKingdom

0.6

49.9

83.2

1.0

119.5

119.5

0.6

16.7

27.8

1.0

53.8

53.8

1.0

68.1

68.1

2.5

348.6

139.5

2000–2003

Finland

1.5

40.8

27.2

1.8

290.5

166.0

1.8

121.1

69.2

1.5

189.8

126.6

Germany

5.5

273.1

49.7

6.0

1 224.1

204.0

4.5

324.1

72.0

5.3

760.0

144.8

Mexico

1.3

282.5

226.0

1.3

2 100.8

1 680.6

1.3

48.5

38.8

1.3

410.8

328.6

Spain

1.3

144.2

115.4

1.3

1 085.1

868.1

1.3

188.6

150.9

1.3

435.5

348.4

1.3

449.8

359.8

Sweden

7.3

286.3

39.5

7.3

3 254.3

448.9

7.3

311.5

43.0

8.0

2 223.2

277.9

6.5

1 622.0

249.5

Switzerland

2.0

103.3

51.6

2.0

110.2

55.1

1.5

126.9

84.6

1.5

354.9

236.6

0.3

8.6

34.5

4 550

81.3

14

14 195254

14

28 126

502

149.7

1.0

178.2

178.2

HWR
Canada

14

1 069

19.0

14

23 483

419

14
PWR

Armenia

0.8

57.0

75.9

1.0

235.0

235.0

0.8

49.9

66.5

1.0

149.7

Belgium

6.3

Brazil

1.5

182.1

29.1

6.3

921.9

147.5

6.3

192.7

30.8

6.3

471.2

75.4

5.0

215.6

43.1

227.0

151.3

1.5

486.8

324.5

1.0

50.8

50.8

1.5

288.6

192.4

1.5

183.1

122.0

China
Czech
Republic

2.3

78.4

34.8

2.5

702.8

281.1

2.0

42.1

21.1

1.0

225.8

225.8

0.8

20.4

27.2

2.0

2.2

1.1

2.0

236.3

118.2

2.0

474.4

237.2

1.5

72.3

48.2

Finland

2.0

83.4

41.7

2.0

169.5

84.7

4.0

686.1

171.5

4.0

949.7

237.4

France47.3

2 708.1

57.3

47.3

13 891.9

294.0

47.3

3 821.4

80.9

47.3

9 712.0

205.6

46.0

296.6

6.5

Germany

11.0

726.5

66.1

11.8

2 828.5

240.7

8.3

1 034.8

125.4

10.5

1 886.4

179.7

3.5

57.2

16.4

Hungary

4.0

224.7

56.2

4.0

1 170.1

292.5

Netherlands

1.0

14.3

14.3

1.0

76.7

76.7

0.8

4.1

5.5

1.0

64.0

64.0

0.3

16.3

65.0

Slovenia

1.0

138.2

138.2

1.0

306.5

306.5

1.0

24.9

24.9

1.0

86.4

86.4

0.3

63.9

255.5

South Africa

1.5

59.2

39.4

1.5

538.8

359.2

1.3

114.5

91.6

1.5

221.2

147.5

1.0

83.3

83.3

Spain

5.5

513.6

93.4

5.5

944.9

171.8

5.5

193.5

35.2

5.5

631.2

114.8

5.3

419.4

79.9

Sweden

3.0

178.6

59.5

3.0

845.7

281.9

3.0

81.1

27.0

3.0

237.6

79.2

Switzerland

6.0

446.9

74.5

2.8

392.0

142.6

0.8

19.0

25.3

2.0

410.4

205.2

0.5

10.0

20.0

UnitedKingdom

1.0

49.4

49.4

1.0

116.8

116.8

0.8

5.6

7.5

1.0

63.3

63.3

1.0

96.0

96.0

365

a Refuelling: all activities related to refuelling, including the cleaning of the refuelling pool.
b Maintenance: work on reactor vessel or internals, steam generators, residual or shutdown heat removal system and safety injection system, chemical and volume control system, pressurizer, reactor water clean-up system, reactor coolant pumps,
primary circuit, valves, steam system, recirculation system and coolant pump seal water system, control rod drives.
c Servicing: general work, scaffolding and insulation.
d Other: other work not listed above.

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

BWR

366

UNSCEAR 2008 REPORT: VOLUME I

Table 68. Occupational exposure due to the decommissioning of 13 nuclear power plants in the United States
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMeasurably exposed workers(103)

Annual collective dose(man Sv)

Average annual effective dose(mSv)

1995–1999

1.90

3.83

2.01

2000–2002

2.17

4.15

1.91

Table 69. Estimated worldwide levels of exposure due to fuel reprocessing
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Annual collective effective dose(man Sv)

Average annual effective dose(mSv)

1975–1979

8

53

7.1

1980–1984

9

46

4.9

1985–1989

17

36

2.5

1990–1994

45

67

1.5

1995–1999

59

61

1.1

2000–2002

76

68

0.9

Table 70. Estimated worldwide levels of exposure due to nuclear fuel cycle research
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Annual collective effective dose(man Sv)

Average annual effective dose(mSv)

1975–1979

120

170

1.4

1980–1984

130

150

1.1

1985–1989

130

100

0.8

1990–1994

120

90

0.8

1995–1999

96

37

0.4

2000–2002

90

36

0.4

Table 71.

Occupational exposure due to radioactive waste management in the nuclear fuel cycle

Country

ChinaaUnited Kingdomb

United Statesc

PeriodMonitored
workers(103)

Measurably
exposed workers(103)

Annualcollective dose

Average annual effective dose
(mSv)

Total
(man Sv)

Monitored
workers

Measurably
exposed workers

Distribution ratio(number of workers)
NR20

NR10

1990–1994

1.51

5.33

3.53

0.09

1995–2000

1.30

4.03

3.10

0.08

1995–1999

0.29

0.09

0.30

2000–2002

0.35

0.07

0.20

NR5

NR1

1995–1999

8.05

1.90

1.22

0.15

0.65

0.00

0.00

0.00

0.04

2000–2002

5.88

1.76

1.07

0.18

0.61

0.00

0.00

0.00

0.05

a Data from reference [T4].
b Data from reference [I38].
c Data from the UNSCEAR Global Survey of Occupational Radiation Exposures.

Table 72. Worldwide average annual exposures due to the commercial nuclear fuel cyclea

Practice

Monitored
workersb(103)

Average annual collective
effective dose(man Sv)

Average annual effective dose
to monitored workers(mSv)

Distribution ratioc
NR15d

SR15

1975–1979
Mininge,f

240

1 300

5.7

5.5

0.37

0.69

Millinge,f

12

124

0.5

10

0.41

0.76

Enrichmente

11

5

0.02

0.5

0.00

0.00

Fuel fabrication

20

36

0.6

1.8

0.012

0.38g

Reactor operation

150

600

11.0

4.1

0.078h

0.60i

Reprocessing j

78

53

0.7

7.1

0.16

0.29

Research

120

170

1.0

1.4

0.035

0.42

Total

560

2 300

20

4.4

0.20

0.63

1980–1984
Mininge,f

310

1 600

5.5

5.1

0.30

0.61

Millinge,f

23

117

0.4

5.1

0.30

0.64

Enrichmente

4

1

0.02

0.2

0.00

0.00

Fuel fabrication

21

21

0.2

1.0

0.002

0.11g

Reactor operation

290

1 000

10.0

3.6

0.069h0.52i

Reprocessing j

9

46

0.8

4.9

0.10

0.39

Research

130

150

1.0

1.1

0.021

Total

800

3 000

18

3.7

0.16

1985–1989
260

1 100

4.3

4.4

0.25

0.52

Millinge,f

18

116

0.4

6.3

0.18

0.43

Enrichmente

5.0

0.4

0.02

0.1

0.00

0.00

Fuel fabrication

28

22

0.1

0.8

0.002

0.019gReactor operation

430

1 100

5.7

2.5

0.033h0.34i

Reprocessing j

12

36

0.7

3.0

0.064

0.12 jResearch

130

100

1.0

0.8

0.011

0.30

Total

888

2 500

12

2.6

0.10

0.42

367

Mininge,f

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Average annual collective
effective dose per unit
energy generated(man Sv/GW a)

Monitored
workersb(103)

Average annual collective
effective dose(man Sv)

Average annual collective
effective dose per unit
energy generated(man Sv/GW a)

Average annual effective dose
to monitored workers
(mSv)

368

Practice

Distribution ratioc
NR15d

SR15

1990–1994
Mininge,f69 (62)310

1.7

4.5 (5.0)

0.10

0.32

Millinge,f

6

20

0.1

3.3

0.00

0.01

13

1

0.02

0.1

0.00

0.00

Fuel fabrication

21 (11)22

0.1

1.0 (2.0)

0.01

0.11

Reactor operation

530 (300)

900

3.9

1.4 (2.7)

0.00h

0.08

Reprocessing j,k

45 (24)67

3.0

1.5 (2.8)

0.00

0.13

Research

120 (36)90

1.0

0.8 (2.5)

0.00

0.22

Total

800 (450)

1 400

9.8

1.8 (3.1)

0.01

0.11

0.04

0.14

1995–1999
Mininge,f

22

85

0.5

3.9

Millinge,f

3

4

0.03

1.6

Enrichmente

17

1

0.02

0.1

0.00

0.00

Fuel fabrication

22

30

0.1

1.4

0.00

0.01

2.5

1.5

0.00

0.03

Reactor operation

448

779

Reprocessing j,k59

61

Research

96

37

1

0.4

0.01

0.22

Total

700

1 000

1

1.4

0.01

0.07

0.05

0.14

1.1

2000–2002
Mininge,f

12

22

0.1

1.9

Millinge,f

3

3

0.02

1.1

Enrichmente

18

2

0.02

0.1

0.00

0.00

Fuel fabrication

20

31

0.1

1.6

0.01

0.01

2.5

1.0

0.02

0.13

Reactor operation

437

617

Reprocessing j,k76

68

Research

90

36

1

0.4

0.01

0.02

Total

660

800

1

1.0

0.02

0.07

0.9

a Data are annual values averaged over the indicated periods.
b Data in parentheses are for measurably exposed workers.
c The values of the distribution ratios should be considered as only indicative of worldwide levels, as they are in general based on data from far fewer countries than the data for the number of workers and collective doses.

UNSCEAR 2008 REPORT: VOLUME I

Enrichmente

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

d Ratio applies to monitored workers.
e Also includes uranium obtained or processed for purposes other than the commercial nuclear fuel cycle.
f For 1985–1989 the data for mining and milling (except for NR and SR) have been modified from those reported by using a conversion factor of 5.6 mSv/WLM for exposure to radon daughters (10 mSv/WLM used in the reported data). The ratios
NR15 and SR15 are averages of reported data in which, in general, the previously used conversion factor has been applied. The tabulated ratios are thus strictly for a value of E somewhat less than 15 mSv. The relationship between the reported
and the revised data is not linear, because exposure occurs from other sources besides the inhalation of radon progeny. For 1990–1994 a conversion factor of 5.0 mSv/WLM for exposure to radon daughters has been used.
g Ratio applies to LWR and HWR fuels only, as data for other fuels are not available; the ratio would be smaller if all fuel types were included.
h Does not include data for LWGRs, FBRs and HTGRs.
i Does not include data for GCRs, LWGRs, FBRs and HTGRs.
j Also includes the reprocessing of some fuel from the defence nuclear fuel cycle.
k In the absence of sufficient data on equivalent electrical energy generated by reporting countries for 1990–1994, the Committee has taken the normalized average annual collective effective dose per unit energy generated to be the same as thatfor the previous period.

369

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UNSCEAR 2008 REPORT: VOLUME I

Table 73. Exposure of physicians during lung biopsy (mGy per minute of fluoroscopy) [N15]
Fluoroscopy conditions set

AB

Use of needle holder

No

Yes

PhysicianOperatingAssistingOperatingAssistingCorner of the eye, right

93 ± 4415.0 ± 10.4

15.3 ± 2.9

2.54 ± 0.93

Corner of the eye, left

23 ± 14.8

4.9 ± 3.6

3.1 ± 0.48

1.23 ± 0.78

Neck (thyroid)

86 ± 46

15.2 ± 10.1

8.0 ± 1.02

2.3 ± 1.54

Upper arm, right

125 ± 68

21 ± 24

18.1 ± 6.2

2.4 ± 0.83

Upper arm, left

19.8 ± 11.9

7.8 ± 5.9

2.8 ± 2.5

1.27 ± 1.13

10 900 ± 11 600

8.4 ± 7.5240 ± 125

1.76 ± 0.99

Back of the hand, left

150 ± 117

6.6 ± 5.5140 ± 0.08

1.10 ± 0.90

Fingers, right

2 600 ± 2 100

45 ± 8084 ± 101

2.45 ± 1.02

Fingers, left

590 ± 400

4.2 ± 3.2

6.9 ± 0.53

1.66 ± 1.43

Back of the head

38 ± 20

5.2 ± 5.5

1.27 ± 1.05

0.79 ± 0.65

Back

4.2 ± 2.7

0.98 ± 1.05

2.4 ± 0.47

0.70 ± 0.20

Inside of femur

3.3 ± 3.0

2.3 ± 3.0

0.68 ± 0.40

0.85 ± 0.90

Chest, inside protector

7.6 ± 5.4

2.3 ± 2.9

0.94 ± 0.46

1.35 ± 1.50

Abdomen, insider protector

6.9 ± 4.6

1.23 ± 1.35

0.54 ± 0.54

0.76 ± 0.69

Chest, outside protector

66 ± 35

9.8 ± 8.5

9.0 ± 1.83

1.48 ± 0.97

Abdomen, outside protector

68 ± 397.9 ± 8.8

13.1 ± 2.2

1.11 ± 0.59

Back of the hand, right

Table 74.

Occupational doses incurred by primary medical doctors during interventional procedures [S21]

Procedure type (number ofmeasurements)

Coronariography(n = 62)

Angioplasty(n = 30)

Arteriography(n = 4)

TLD position/
dosimetric quantity

Range of dose(mSv)

Average dose
(mSv)

Chest outside the apron/Hp(10)0–2.35

0.29

Chest inside the apron/Hp(10)0–0.270.06

Right hand/Hp(0.07)

0–2.54

0.26

Left hand/Hp(0.07)

0–3.880.38

Knee/Hp(0.07)

0–1.61

0.21

Neck without collar/Hp(10)0.18–2.88

Neck with collar/Hp(10)0–0.270.13

Forehead/Hp(3)0–0.82

0.14

Chest outside the apron/Hp(10)0–0.180.04

Chest inside the apron/Hp(10)0–0.180.02

Right hand/Hp(0.07)

0–0.30

0.05

Left hand/Hp(0.07)

0–0.400.13

Knee/Hp(0.07)

0–0.720.10

Neck without collar/Hp(10)0.14–0.27

Neck with collar/Hp(10)0–0.140.07

Forehead/Hp(3)0–0.330.05

Chest outside the apron/Hp(10)0–0.18

0.09

Chest inside the apron/Hp(10)0–0.130.07

Right hand/Hp(0.07)

0–0.180.05

Left hand/Hp(0.07)

0–0.650.16

Knee/Hp(0.07)

0–0.180.09

Neck without collar/Hp(10)0.18–0.18

Neck with collar/Hp(10)

0–0

0.05

Forehead/Hp(3)

0–0

0.00

Average time offluoroscopy (min)Average numberof frames

7.8

991

13

762

4

390

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Procedure type (number ofmeasurements)

Valvuloplasty(n = 5)

TLD position/
dosimetric quantity

Range of dose(mSv)

Average dose
(mSv)

Chest outside the apron/Hp(10)0–0.180.04

Chest inside the apron/Hp(10)0–0.270.08

Right hand/Hp(0.07)

0–0

0.00

Left hand/Hp(0.07)

0–0.370.07

Knee/Hp(0.07)

0–0.34

0.12

Neck without collar/Hp(10)0–0Neck with collar/Hp(10)

0–0

0.00

Forehead/Hp(3)

0–0

0.00

371

Average time offluoroscopy (min)Average numberof frames

16

225

Table 75. Mean and maximum doses for the staff of the interventional cardiology (IC) and interventional radiology (IR)services of a university hospital [V9]Year

Professionals

1999

2000

2001

1999

2000

2001

Table 76.

Maximum dose (mSv/month)

Whole body(under the apron)

Shoulder(on the apron)

Hand (wrist)

ShoulderIC physicians

0.1

1.0

1.1

13.7

IC fellows

0.3

2.3

IC nurses

0.2

0.9

IC physicians

0.2

0.7

1.3

18.8

IC fellows

0.2

2.5

IC nurses

0.1

0.9

IC physicians

0.1

2.0

9.3

10.5

IC fellows

0.3

2.6

IC nurses

0.1

1.0

IR physicians

0.1

2.7

2.9

5.8

IR technicians

0.1

0.4

IR physicians

0.1

2.1

3.2

7.1

IR technicians

0.1

0.2

IR physicians

0.1

2.1

3.6

3.5

IR technicians

0.1

0.4

Dose per procedure in vascular interventional radiology, measured by TLD [V7]

TLD location
Left shoulder
Right eyeLeft eye
ForeheadNeck
Right hand
Left hand
Left forearm
Arm

Mean dose (mSv/a)

Sample size

Average dose (µSv)

Median dose (µSv)

Range (µSv)

21
18
19
19
19
23
23
22
29

283
296
284
222
325
260
396
326
365

182
122
95
159
138
120
184
225
243

45–1 214
45–2 103
40–1 683
19–1 013
48–2 104
47–97440–2 150
40–1 886
50–1 068

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UNSCEAR 2008 REPORT: VOLUME I

Table 77.

Dose per procedure in interventional cardiology, measured by TLD [V7]

TLD location

Sample size

Average dose(µSv)

Median dose
(µSv)

Range
(µSv)

185
140
193
178
214
144
256
445
414

30–1 031
39–742
53–1 005
40–934
43–81645–921
60–1 500
88–2 890
70–1 919

145
140
148
150
160
128
195
350
237

30–25052–252
53–46040–415
43–39845–46660–740
88–2 890
70–727

308
138
425
330
389
149
372
801
1 027

125–1 031
39–742
158–1 005
103–934
60–81645–921
65–1 500
168–2 006
108–1 919

Average of doses with and without lead screen
Left shoulder
Right eyeLeft eye
ForeheadNeck
Right hand
Left hand
Left forearm
Arm

55
53
54
53
54
54
58
54
54

252
167
294
236
269
191
364
646
618
With lead screen

Left shoulder
Right eyeLeft eye
ForeheadNeck
Right hand
Left hand
Left forearm
Arm

29
29
29
29
29
28
31
29
30

136
136
170
145
163
147
235
440
265
Without lead screen

Left shoulder
Right eyeLeft eye
ForeheadNeck
Right hand
Left hand
Left forearm
Arm

Table 78.

26
24
25
26
27
25
25
25
24

382
205
439
344
392
242
514
885
1 061

Occupational doses associated with specific interventional procedures

Dose per procedure(mSv)

Procedure type

X-ray system

Relevant exposure
parameters

Protection tools used

Reference

0.215–0.370(at thyroid level)

Coronary angiography

Philips Polydiagnost C2

2.8–3.4 min fluoroscopy,
637–1 058 frames

Ceiling-mounted screen,
1 mm lead[S37]

0.008–0.113(forehead)

Cardiac catheterization

Three centres, five X-ray
units

2–3 min fluoroscopy,
500–2 000 frames,
4 400 cGy/cm2

Ceiling-mounted screen,protective eye shields(reduction in exposure rateby a factor of about 20)[P12]

0.05–0.14(neck)

Vascular and liver

Philips Integris 3000 GE L-U

5 400–6 700 cGy/cm2

Ceiling-mounted screen in
only one room

[W15]

0.28 (left eye), 0.20
(thyroid)

Cardiac catheter ablation

Siemens Angioskop D

44 min fluoroscopy

Ceiling-mounted screen

[C1]

0.05
(collar level)

Coronariography and PTCA

Philips Integris 3000 DC

6 600 cGy/cm2

Movable shield[Z6]

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Dose per procedure(mSv)

Procedure type

X-ray system

Relevant exposure
parameters

Protection tools used

373

Reference

0.43(forehead, eye)Coronary angiography andPTCA

14 laboratories

Cine 53 s, fluoroscopy
6.8 minProtective eyeglasses

[K3]

0.014
(eye)

Cerebral angiography,
arterial embolization

CGR DG 300

4 850 cGy/cm2,
12 220 cGy/cm2

Waist-height lead shield

[M10]

1–2
(eye)

Interventional radiology

General Electric Phasix 80

10 images, 10 min
fluoroscopy

No ceiling-mounted
screen; lead apron andgloves[V8]

Table 79. Worldwide levels of occupational exposure due to diagnostic radiology
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
Period1975–1979

Monitored workers(103)

Measurably exposed
workers(103)

630

Annual collective
effective dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

600

0.9

1980–1984

1 060

720

0.7

1985–1989

1 350

760

0.6

1990–1994

950

470

0.5

1995–1999

6 670

350

3 300

0.5

2000–2002

6 670

3 300

0.5

Measurably exposed
workers

1.3

Table 80. Worldwide levels of occupational exposure due to dental practiceData from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Measurably exposed
workers(103)

Annual collective effective
dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

1975–1979

370

120

0.32

1980–1984

500

93

0.20

1985–1989480

1990–1994

265

17

25

0.05

16

0.06

1995–1999

404

24

0.06

2000–2002

404

24

0.06

Measurably exposed
workers

0.89

Table 81. Annual doses incurred by PET workers [Z5]WorkerTechnologist 1
Technologist 2
Physician 1
Physician 2

Effective dose
(mSv)

Equivalent dose to hands(mSv)

Film monthly dose range(mSv)

Ring monthly dose range
(mSv)

8.0
4.6
2.2
1.9

90.0
63.5
5.2
6.0

0.4–1.1
0.3–0.7
0.1–0.7
0.1–0.7

4.1–9.9
0.7–9.1
0.2–1.2
0.3–4.2

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UNSCEAR 2008 REPORT: VOLUME I

Table 82. Occupational doses (μSv) for various tasks, different patients and different technologists
Each patient was administered 555 MBq of 18F FDG [M17]
Patient 1
Technologist A

Patient 2
Technologist A

Patient 3
Technologist B

Patient 4
Technologist B

PC Cc

PC Aa

PC Aa

PC Bb

Measure dosage
Carry dosage
InjectEscort to waiting area
InterviewEscort to imaging lab
Position
Reposition after 1 h
Exit

1.4
0.1
1.8
0.5
1.8
1.4
2.0
0.0
0.2

0.2
0.0
0.7d
0.8
0.0
0.3
3.4
0.0
0.0

4.0
0.1
1.9
0.0
0.0
0.5
1.2
0.0
0.0

2.4
0.9
3.9
1.8
0.7
0.5
6.8
1.1
0.8

Total

9.2

5.4

7.7

18.9

a PC A: patient fully ambulatory and able to follow instructions.
b PC B: patient haltingly ambulatory, occasionally in a wheelchair.
c PC C: patient confined to a wheelchair, able to follow instructions with physical assistance.
d Physician performed injection.

Table 83.

Occupational exposures due to the use of different types of scan [W11]Scan

IsotopeF
Tc-MDP
Tc-RBC
201
Tl
Tc-DTPA
67
Ga
Sestamibi

PET
Bone
LVEFa
Thallium
Renal
Gallium
Cardiac18

Number ofpatients

Median administeredactivity(MBq)

Median timepost-injection(h)Median exposure (µSv/h) for distance
2m

1m

0.5 m

0m

41
57
23
28
33
38
31

57.4
760.2
900.0
250.0
389.6
400.0
900.0

1.1
3.5
0.3
1.0
2.0
120.0
1.0

2.0
1.8
3.0
1.1
1.0
0.9
4.1

5.2
3.4
8.0
1.9
2.1
1.6
9.4

14.1
9.0
21.0
5.2
4.6
4.2
19.3

71.0
43.0
144.0
28.0
23.0
19.1
110.3

a Left ventricular ejection fraction.

Table 84.

Maximum daily dose to skin of the hands during radiosynoviorthesis [B9]

Measurement cycle

A-1a
A-2
A-3
A-4
A-5
A-6
B-1
C-1
F-1
F-2
G-1
G-2
H-1
I-1
I-2
I-3
I-4

Maximum daily dose Hp(0.07) (mSv)
Preparation

Application

Assistance

82
101
16
18
Not measured
Not measured
108
14
7
8
15
4
55
8
Not measured
6
Not measured

43
132
16
33
1b
1b
27
41
62
1b
5
11
207
31 (Doctor 1)
84 (Doctor 2)
1b (Doctor 1)
1b (Doctor 2)

5
102
Not measured
Not measured
Not measured
Not measured
Not measured
9
36
1
2
None
None
None
None
None

a Seven different institutions.
b Use of forceps during application of radionuclides.

Y administered(MBq)

Specific dose(µSv/MBq)

805
1 675
620
1 480
555
1 110
2 035
555
460
2 005
180
360
888
1 332
2 442
1 554
1 332

53
79
26
22
1.8
0.9
13
74
135
0.5
28
31
233
23
34
0.6
0.8

90

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

375

Table 85. Worldwide levels of occupational exposure in nuclear medicine
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Measurably exposed
workers(103)

Annual collective
effective dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

1975–1979

61

62

1.0

1980–1984

81

85

1.0

1985–1989

90

85

1.0

1990–1994

115

90

0.8

1995–1999

117

65

89

0.8

2000–2002

120

87

0.7

Measurably exposed
workers

1.4

Table 86. Mean dose per application of 125I in prostate using afterloading technique, and annual effective dose incurred bystaff at various distances from the source [G4]Staff

µSv per application (annual) for distancePhysicians
Physicist
NursesAssistants

0.5 m

1m

2m

3m

15 (1 200)

4.3 (347)
4.3 (347)
4.3 (347)

0.2 (13)
1.3 (104)0.2 (14)0.06 (4.8)

15 (1 200)

0.06 (4.8)
0.06 (4.8)

Table 87. Worldwide levels of occupational exposure in radiotherapy
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Measurably exposed
workers(103)

Annual collective
effective dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

1975–1979

84

190

2.2

1980–1984

110

180

1.6

1985–1989

110

100

0.9

1990–1994

120

65

0.6

1995–1999

264

132

0.5

2000–2002

264

132

0.5

48

Measurably exposed
workers

1.3

Table 88. Trends of occupational exposure in all medical uses
Data from the UNSCEAR Global Survey of Occupational Radiation Exposures
PeriodMonitored workers(103)

Measurably exposed
workers(103)

Annual collective
effective dose(man Sv)

Average annual effective dose (mSv)
Monitored workers

Measurably exposed
workers

1975–1979

1 280

650

993

0.8

1.5

1980–1984

1 890

520

1 140

0.6

1.7

1985–1989

2 220

590

1 030

0.5

1.7

1990–1994

2 320

550

760

0.3

1.4

1995–1999

7 440

3 540

0.5

2000–2002

7 440

3 540

0.5

Year

Multiple-location
Monitored
workers

Measurably
exposed workers

Annual collective
effective dose(man Sv)

376

Table 89. Occupational exposure in industrial radiography in the United States [U29, U30, U31, U32, U33, U34, U36, U37]
Single-location
Average annual
effective dose (mSv)
Monitored
workers

Measurably
exposed workers

Monitored
workers

Measurably
exposed workers

Annual collective
effective dose(man Sv)

Average annual
effective dose (mSv)
Monitored
workers

Measurably
exposed workers

3 245

2 404

13.32

4.10

5.54

285

61

0.06

0.21

0.99

1996

3 340

2 477

13.75

4.12

5.55

291

60

0.10

0.35

1.67

1997

3 140

2 370

12.81

4.08

5.40

296

84

0.10

0.34

1.19

1998

4 571

3 355

18.51

4.05

5.52

369

84

0.08

0.22

0.95

1999

3 571

2 777

15.44

4.32

5.56

266

50

0.07

0.26

1.41

2000

3 029

2 399

14.80

4.89

6.17

258

78

0.08

0.31

1.01

2001

3 522

3 082

21.05

5.98

6.83

256

79

0.06

0.23

0.75

2002

3 292

2 773

17.19

5.22

6.20

112

55

0.04

0.40

0.81

1995–1999

3 573

2 677

14.77

4.13

5.51

301

68

0.08

0.28

1.24

2000–2002

3 281

2 751

17.68

5.36

6.40

209

71

0.06

0.31

0.86

UNSCEAR 2008 REPORT: VOLUME I

1995

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

377

Table 90. Radiation dose to workers in a cyclotron and radiochemistry laboratory
As measured by national personnel monitoring service using CaSO4:Dy TLD badges [P3]Occupational exposure in cyclotron/
radiochemistry laboratory
Cyclotron

Whole-body dose(mSv)

Extremity (wrist) dose
(mSv)

0.35

7.95

Cyclotron

0.85

0.45

Radiochemistry

0.60

4.45

Radiochemistry

1.80

3.4

Table 91. Global occupational exposures due to the nuclear fuel cycle and natural sources of radiation
Practice

Monitored workers(103)

Average annual collective
effective dose(man Sv)

Average annual effective dose
to monitored workers(mSv)

Nuclear fuel cycle
1995–1999
Mining

22

85

3.9

Milling34

1.6

Enrichment

17

1

0.1

Fuel fabrication

22

30

1.4

Reactor operation

448

779

1.5

Reprocessing

59

61

1.1

Research

96

37

0.4

Total

670

1 000

1.4

2000–2002
Mining

12

22

1.9

Milling33

1.1

Enrichment

18

2

0.1

Fuel fabrication

20

28

1.6

Reactor operation

437

600

1.0

Reprocessing

76

68

0.9

Research

90

36

0.4

Total

660

800

1.0

6 900

16 560

2.4

Natural sources of radiation
1995–1999 and 2000–2002
Coal mining
Other mining

4 600

13 800

3.0

Workplaces other than mines

1 250

6 000

4.8

Aircrew

300

900

3.0

Total

13 05037 260

2.9

378

UNSCEAR 2008 REPORT: VOLUME I

Table 92.

Global occupational exposures associated with man-made and natural sources of radiation

Source of exposure

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

6 500

13 05013 050

Number of monitored workers (103)Natural radiation
Nuclear fuel cycle

560

800

888

800

670

660

Medical uses

1 280

1 890

2 220

2 320

7 440

7 440

Industrial uses

530

690

560

700

790

869

Military activities

310

350

400

420

378

331

Miscellaneous

140

180

160

360

476

565

Total (man-made)

2 820

3 910

4 228

4 600

9 754

9 865

Total

2 820

3 910

4 228

11 10022 80422 915

11 70037 26037 260

1 400

1 000

800

Annual collective effective dose (man Sv)
Natural radiation
Nuclear fuel cycle

2 300

3 000

2 500

Medical uses

1 000

1 140

1 030

760

3 540

3 540

Industrial uses

870

940

510

360

315

289

Military activities

420

250

250

100

52

45

Miscellaneous

70

40

20

40

53

56

Total (man-made)

4 660

5 370

4 310

2 660

4 960

4 730

Total

4 660

5 370

4 310

14 36042 22041 990

1.8

2.9

2.9

Average annual effective dose (mSv)
Natural radiation
Nuclear fuel cycle

4.4

3.7

2.6

1.8

1.4

1.0

Medical uses

0.8

0.6

0.5

0.3

0.5

0.5

Industrial uses

1.6

1.4

0.9

0.5

0.4

0.3

Military activities

1.3

0.7

0.7

0.2

0.1

0.1

Miscellaneous

0.5

0.3

0.2

0.1

0.1

0.1

Total (man-made)

1.7

1.3

1.0

0.6

0.5

0.4

Total

1.7

1.3

1.0

1.3

1.8

1.8

figures
Figure I.

Variation in solar activity in terms of the historical monthly average sunspot numbers during solar cycles [N4]300200MONTHLY AVERAGE SUNSPOT NUMBER

100
0
1750

1760

1770

1780

1790

1800
YEAR

1810

1820

1830

1840

1850

1860

1870

1880

1890

1900
YEAR

1910

1920

1930

1940

1950

300
200
100
0
1850
300
200
100
0
1950

1960

1970

1980
YEAR

1990

2000

160

160

140

140

120

120

100

100

80

80

60

60

40

40

20

20

0
1980

1985

1990

1995

2000

YEAR
Sunspot index

Dose (FRA-NYC-FRA)
379

2005

0

DOSE �µSv)

SUNSPOT INDEX

Figure II. Example of the influence of variation in solar activity on cosmic ray dose received during a return transcontinental
flight between Frankfurt and New York City [S38]380

UNSCEAR 2008 REPORT: VOLUME ILATITUDE

Figure III. Vertical cut-off rigidity, Rc (in GV), at 20 km altitude [S30]

LONGITUDE

Components of the dose equivalent rate due to cosmic rays in the atmosphere [U3]80
CONTRIBUTION TO DOSE EQUIVALENT RATE �%�

Figure IV.

Muons
Electrons/photons
Protons
Charged pions
Neutrons

70
60
50
40
30
20
10
0
0

5 000

10 000
ALTITUDE �m�

15 000

20 000ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure V. Reported concentrations of 238U in soil
Data from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
AVERAGE VALUES (Bq/kg)

MAXIMUM VALUES (Bq/kg)

Legend>1 000300-1 000
100-300
30-10010-30<10NData source: United Nations Scientific Committee
on the Effects of Atomic Radiation (UNSCEAR)
3 000

0

3 000

Kilometres

6 000

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

381

382

UNSCEAR 2008 REPORT: VOLUME I

Figure VI. Reported concentrations of 232Th in soil
Data from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources
AVERAGE VALUES (Bq/kg)

MAXIMUM VALUES (Bq/kg)

Legend300-1 000
100-300
30-10010-30<10NData source: United Nations Scientific Committee
on the Effects of Atomic Radiation (UNSCEAR)
3 000

0

3 000

Kilometres

6 000

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

383

Figure VII. Reported concentrations of 40K in soil
Data from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources.
AVERAGE VALUES (Bq/kg)

MAXIMUM VALUES (Bq/kg)

LegendN>3 000

Data source: United Nations Scientific Committee
on the Effects of Atomic Radiation (UNSCEAR)

1 000-3 000
300-1 000
<300
No data2 000

02 000 4 000 6 000 8 000

Kilometres

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

384

UNSCEAR 2008 REPORT: VOLUME I

Figure VIII. Distribution of population with respect to ranges of absorbed dose rate in air(Left: number of persons in each range; right: per cent cumulative distribution fraction for each range)
120
CUMULATIVE FREQUENCY �%�

600

POPULATION �106�

500
400
300
200
100

100
80
60
40
20

0
<20
20–
29
30–
39
40–
49
50–
59
60–
69
70–
79
80–
89
90–
99
100
–19
9
200
–29
9
>30
0

<20
20–
29
30–
39
40–
49
50–
59
60–
69
70–
79
80–
89
90–
99
100
–19
9
200
–29
9
>30
0

0

DOSE RATE �nGy/h�DOSE RATE �nGy/h�

Figure IX. Reported ratios of 238U/232Th concentrations in soil
Data from the UNSCEAR Global Survey on Exposures to Natural Radiation Sources

N
LegendRatio of U/Th in soil
Ratio = >1
Ratio = 1
Ratio = <1

2 000

0

2 000

Kilometres

4 000

Data source: United Nations Scientific Committee
on the Effects of Atomic Radiation (UNSCEAR)
The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

80 000100 000

8 000

10 000

800

1 000

80

100

8

10

0.8

1

0.08

0.1

0.008

0.01

0.000 8

Arg
en
tin
a
Bra
zil
C
Cz hinaechRe
Fin p.
lan
Fra d
Ge ncerman
Gre y
Hu ece
ng
ary
Ind
ia
It
Mo alyro
Ro cco
ma
nia
S
p
Sw
ain
i
Un tzerlaited S ndtat
es

238

1 000 000

Cumulative frequency distribution of 238U concentration in drinking water
99.8
99
CUMULATIVE FREQUENCY �%�

Figure XI.

95
90
70
50
30
10
5
1
0.2
10-1

100

101

102

AVERAGE 238U CONCENTRATION �mBq/L)

103

NATURAL URANIUM CONCENTRATION �μg/L�

U CONCENTRATION �mBq/L�

Figure X. Ranges of 238U concentration in drinking water

385

386

UNSCEAR 2008 REPORT: VOLUME I

Cumulative frequency distribution of the concentrations in bone of radionuclides of the uranium and thorium series

CONCENTRATION (mBq/kg)

Figure XII.
[U3]

PERCENTAGE

Figure XIII. Distribution of 210Po in human body organs at
steady state after chronic ingestion of radionuclides of theuranium and thorium series [C23, F8]

Figure XIV. Distribution of average radon concentrations in
the indoor air of houses
Data from the UNSCEAR Global Survey on Exposures to Natural
Radiation Sources and reference [D14]Lung 0.3(0.16) Liver 1.4 (0.4)
Kidney 0.2 (0.33)

99.8
99

Muscle 3.4 (0.06)
Red bonemarrow
0.23 (0.01)
Blood 0.17
(0.02)CUMULATIVE FREQUENCY �%�

Spleen 0.04 (0.11)

95
90
70
50
30
10
5
1

Skeleton 16(0.38)TOTAL ORGAN VALUE Bq (EQUIVALENT DOSE mSv/a)

0.2
100

101

102

INDOOR RADON CONCENTRATION �Bq/m3�

103

Figure XV. Total worldwide production of uranium (t) to 2003

Canada374 548

Finland30

PolandGermany
660
Belgium 219 239
UkraineCzech Republic
680
9 900
108 649
Hungary Romania
France21 080 17 989
75 965
Bulgaria
16 735
Spain
Portugal 6 156
3 680

United States
356 485

Kazakhstan
24 639Uzbekistan

Niger
91 186

United States
356 485

Brazil
1 645

India
7Kazakhstan
963
Uzbekistan
23 682
Pakistan
931

Niger
91 186

ZambiaDR
102 Gabon Congo,
Madagascar
25 403 25 600
785
Namibia
Zambia78 794
Madagascar
South Africa Namibia 102785
157 61878 794

Brazil
1 645

Argentina
2 631

China

Russian Federation
32 27
136 689

24 639Gabon Congo, DR
25 403 25 600

Mexico49

Mongolia
535

Finland 23 682
30
SwedenPakistan91

931

Canada374 548

Mexico49

Russian Federation
32 136

Sweden91

South Africa
157 618

Argentina
2 631

Japan
84

Mongolia
535
China
27 689

Japan
84

India
7 963

Australia
113 304
Australia
113 304

N

N

2 000

2 000

0

2 000
Kilometres

0

4 000

2 000

6 000
Kilometres

4 000

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

PolandGermany
660
219
239
Belgium
UkraineCzech
Republic
680
9 900
108 649
Hungary Romania
France21 080 17 989
75 965
Bulgaria
16 735
Spain
Portugal 6 156
3 680

6 000

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations
387

388

UNSCEAR 2008 REPORT: VOLUME I

Figure XVI.

Annual uranium production from 1990 to 2003 [O16, O17, O21]
55 000URANIUM PRODUCTION �t�

50 00045 00040 00035 00030 000
1990

1995

2000
YEAR

2005

Figure XVII. Tailings from uranium mining and milling (106 t)

Finland0.04
SwedenEstonia
1.5
4PolandGermany

Russian Federation
56.85

660
Belgium 219 239
UkraineCzech Republic
680
9 900
108 649
Hungary Romania
France21 080 17 989
75 965
Bulgaria
16 735
Spain
Portugal 6 156
3 680

Canada202.13

United States
235

Kazakhstan
165
Kyrgyzstan
32.3
Uzbekistan Finland
60 Sweden 30
91

China

Pakistan

Canada374 548

MexicoMongolia

United States
356 485

India
8

Niger
17.2

Kazakhstan
24 639Uzbekistan
23 682
Pakistan
931

Mexico
49

Gabon
6.5

Niger
91 186

Congo, DR

Brazil
2.45

ZambiaMongolia
535
China
27 689

Japan
84

India
7 963

Gabon Congo, DR
25 403 25 600

BrazilNamibia1 645
350South Africa
700

Argentina
0.7

Japan
Russian Federation
32 136 0.054

Zambia102
Namibia
78 794South Africa
157 618

Argentina
2 631

Madagascar
785

Australia
79

Australia
113 304

N
2 000

0

2 000
Kilometres

4 000

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

PolandGermany
0.25
Belgium 174.45
UkraineCzech Republic
89.5
89
Hungary
FranceSlovenia
Romania
20.4
29.32
0.7
6.8
Bulgaria
16
Spain
Portugal 1.41
4

6 000

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

389

390

Figure XVIII. Countries with facilities for production of nuclear fuel for power reactors

SwedenUnited Kingdom
CanadaBelgium
FranceUnited States

Russian Federation

Netherlands
Germany
Romania

Rep. of Korea

Spain

India
Finland30

Russian Federation
32 136

Sweden91

Canada374 548

Kazakhstan
24 639Uzbekistan
23 682

United States
356 485Brazil

Pakistan
931

Mexico49

Niger
91 186

Argentina

Japan

China

Mongolia
535
China
27 689

Japan
84

India
7 963

Gabon Congo, DR
25 403 25 600
Brazil
1 645

Zambia102
Namibia
78 794South Africa
157 618

Argentina
2 631

Madagascar
785

Australia
113 304N
2 000

Conversion to UF6

Fuel fabrication

0

2 000

Enrichment
Kilometres

4 000

6 000

Fuel reprocessing

Symbols do not indicate exact locations of facilities within countries.

The boundaries and names on these maps do not imply any
official endorsement or acceptance by the United Nations

UNSCEAR 2008 REPORT: VOLUME I

Pakistan

PolandGermany
660
Belgium 219 239
UkraineCzech Republic
680
9 900
108 649
Hungary Romania
France21 080 17 989
75 965
Bulgaria
16 735
Spain
Portugal 6 156
3 680

Figure XIX. Nuclear power reactors in the world, 1998–2002

Canada17

Finland4
Sweden12

Russian Federation
30

UnitedGermany
KingdomPoland
660
239
Belgium 219
35CzechUkraineRepublic
680
9 900
108 649
Hungary Romania
France21 080 17 989
75 965Spain
Bulgaria
9
16 735
Spain
Portugal 6 156
3 680

United States
104

Kazakhstan
1
Armenia1
FinlandSweden91

30

Canada374 548

Mexico2

Pakistan
2

Russian Federation
32 136

India
14
Kazakhstan
24 639Uzbekistan
23 682

United States
356 485

Pakistan
931

Mexico49

Niger
91 186

Brazil2

Mongolia
535
China
27 689Rep. of Korea
18
Japan
53
China,
Taiwan Province
14
Japan
84

India
7 963

Gabon Congo, DR
25 403 25 600
Brazil
1 645

Argentina
2South Africa
2

Argentina
2 631

Zambia102
Namibia
78 794South Africa
157 618

Madagascar
785

Australia
113 304

N

N

2 000

3 000

0

3 000

0

2 000

6 000

Kilometres

4 000

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Lithuania
2
Netherlands
1 Germany
19
Belgium Czech Republic
Ukraine7
6 Slovakia
14
6
Switzerland
Hungary Romania
Slovenia 4
5France1158Bulgaria
6

6 000

The boundaries and names on these maps do not imply any

The boundaries and names on these maps do not imply any
official endorsementby the United Nations
official endorsement
or acceptance or
byacceptancethe United Nations

Kilometres

391

392

Figure XX.

UNSCEAR 2008 REPORT: VOLUME I

Trends in nuclear energy generation(a) Total installed electrical energy capacity worldwide

INSTALLED CAPACITY �GW�e��

400

300

200

100

0
1960

1965

1970

1975

1980
1985
YEAR

1990

1995

2000

2005

1980
1985
YEAR

1990

1995

2000

2005

1980
1985
YEAR

1990

1995

2000

2005

(b) Total number of nuclear power reactors worldwide

NUMBER OF REACTORS

500
400
300
200
100
0
1960

1965

1970

1975

INSTALLED CAPACITY PER REACTOR UNIT
�GW(e))(c) Average electrical energy capacity per reactor unit
1
0.8
0.6
0.4
0.2
0
1960

1965

1970

1975

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXI.

393

Historical trends of energy generation by nuclear power reactors

ENERGY GENERATED �GW a�

1 000

PWR

100

BWR
GCR
HWR

10

LWGR
FBR
1

Total

0.1
1970

1975

1980

1985

1990

1995

2000

2005

YEAR

Figure XXII.
1998–2002

Contribution of each type of reactor to the total nuclear energy generated in the periods 1970–1997 and
1970–19975%5%

1998–2002

5%

5%

2% 3%

PWR
BWR

24%

61%

PWR
BWR

24%

HWR

HWR

GCR

GCR

LWGR

LWGR
66%

394

UNSCEAR 2008 REPORT: VOLUME I

Figure XXIII. Normalized noble gas releases for different periods and types of reactor

100 000

1 000

100

10

1

974
979
984
5–1
989
7
0–1
9
8
1
994
5–1
9
1
997
0–1
198
9
002
5–1
9
9
1
9
8–2
1
9
9
1
PERIOD

0–1

197

FBR
LWGR
HWR
GCR
BWR
PWR

REACTOR TYPENORMALIZED RELEASES(TBq per GW a)

10 000Figure XXIV. Number of nuclear fuel transports in Germany, including irradiated and non-irradiated fuel and waste, by road,rail, sea and air [B48]

NUMBER OF NUCLEAR FUEL TRANSPORTS

1 200
1 000
800
International

600

National
400
200
0
1994

1995

1996

1997

1998
YEAR

1999

2000

2001

2002

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXV. Number of research reactors worldwide
Operational status (upper figure) and power range (lower figure)250

NUMBER OF REACTORS

200

150

100

50

0
Operational

Shut down

Decommissioned

Under construction

OPERATIONAL STATUS

NUMBER OF REACTORS

200

150

100

50

0
<0.1

0.1 – <1.0

1.0 – <10

10 – <100

100 –<1 000RANGE OF POWER �kW�1 000 –
<10 000

10 000 –<100 000≥100 000

395

N
A o
Fa lge r
ng ri
M
at a
on
a
te Mu ufa
Be ru
llo ro
Isl a
an
d
M E
M ar mu
Ch alde alin
ris n gatm Isla an
Ne s Is d
w lan
M d
ex
Ne ico
va
d
Bi a
k
En in
ew i
et
Pa ak
Jo
c
hn At ificl
st an
Se on ticm Is
No ip lanva ala dya tinTo Zem sktsl
Ka k, A ya
pu ra
st lsk
in
Ya
r

Lo
p

FISSION YIELD �Mt)NA oFa lge r
ng ri
M
at a
on
au
M
te
f
u
Be ru allo ro
Isl a
an
d
M Em
M ar uCh alde alin
ris n gatm Isla an
Ne s Is d
w lan
M d
ex
Ne icovad
B a
En ikin
ew i
et
Pa ak
Jo
c
hn At ific
st lanSe on ticm Is
No ip lanva ala dya tinTo Zem sktsl
Ka k, A ya
pu ra
st lsk
in
Ya
r

Lo
p

NUMBER OF TESTS

396
UNSCEAR 2008 REPORT: VOLUME I

Figure XXVI.
Number of tests and fission yields for different atmospheric layers for each nuclear test site(a) Number of tests at each test site
140

120

100
Former Soviet Union

80
United States

United Kingdom

60
France

40
China

20

0

(b) Fission yield at each test site

100

10

1
Stratosphere

0.1

0.01
Troposphere

Local/regional

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXVII.

397

Worldwide average per caput effective doses from nuclear weapons tests

0.12

ANNUAL EFFECTIVE DOSE �mSv�

0.1
0.08
External
Ingestion
Inhalation
Total

0.06

0.04

0.02

0
1945

1950

1955

1960

1965

1970

1975
YEAR

1980

1985

1990

1995

2000

2005

398

Figure XXVIII. Sites of nuclear weapons tests
Clean map from [C17]
Lithuania
Lithuania
22Netherlands
Netherlands
1 Germany
1 Germany
Novaya Zemlya
19
19
Belgium Czech Republic
Belgium Czech Republic
UkraineUkraine7
6 Slovakia
6 7Slovakia
14
14
6
6
Switzerland
Switzerland
Hungary Romania
Hungary Romania
Slovenia 4Totsk
Slovenia 4 5
5
Semipalatinsk
FranceFinland1
11
1 France
58
58
4
Kapustin Yar Bulgaria
SwedenBulgaria
Lop
Nor
6
12
6

Amchitka

Pacific

NevadaNew Mexico

Canada17Reganne

Canada17

Pakistan
United Kingdom
35

In Ekker Germany

Christmas Island

Mexico2

Mururoa
Fangataufa

Finland30
Sweden91

Mexico2

Canada374 548

Spain
9

Spain
9

Uzbekistan
23 682
Pakistan
931

Pakistan
2

Niger
Brazil91 186
2

Emu
Mongolia
535 Maralinga
China
27 689

Kazakhstan
1
Rep. of Korea
18
Japan
53

EnewetakArmenia
1BikiniArmenia
1

Russian Federation
32 136

United States
356 485 Atlantic

Brazil2

Kazakhstan
1Monte Bello Island
Kazakhstan
24 639Mexico49

Russian Federation Russian Federation
30
30

Japan
84

India
7 963

India
14

Pakistan
2

UNSCEAR 2008 REPORT: VOLUME I

Malden Island

United Kingdom
35

India

Poland660
Belgium 219 239
UkraineCzech Republic
680
9 900
108
649
UnitedStatesUnited States
Hungary Romania
France104
21 080 17 989
75104
965
Bulgaria
16 735
Spain
Portugal 6 156
3 680

Johnston Island

Finland4
Sweden
DPRK
12

India
14

China,
Taiwan Provin
14

Gabon Congo, DR
25 403 25 600
Brazil
1 645

Argentina
2

Argentina
2
Argentina

China
France
India

N

DPRK
2 631
Pakistan
Former Soviet Union

NN

2 000

000
33000

0

2 000

South Africa
2

Madagascar
785

South Africa
2Australia
113 304

United Kingdom
United Kingdom and United States
United States

4 000

0 3 000 Kilometres3 0000000
Kilometres
Kilometres

Zambia102
Namibia
78 794South Africa
157 618

6 000

000
63000
6 000
Kilometreson these maps doboundaries
not imply any names on these maps d
6 The
000 boundaries and names
The boundaries andThenames on theseandmaps do not imply any
official endorsement orofficial
acceptancebytheUnitedNationsby theofficial
endorsement
orUnited
acceptanceby the U
endorsement or acceptance
Nations

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXIX.

Number of nuclear tests performed by each country
1 200

NUMBER OF TESTS

1 000
800
Safety trials
600

Underground tests
Atmospheric tests

400
200

DP
RK

an
ist
Pa
k

ia
In
d

na
Ch
i

ce

ite

d

Ki

Fr
an

m
ng
do

io
Un
et
Un

vi
So
er

Fo
r

m

Un
ite
d

St
at
es

n

0

Figure XXX. Sites where radioactive waste has been dumped at sea [I11]North-East PacificNorth-West Atlantic

North-East AtlanticWest Pacific0.55 PBq

2.94 PBq

42.31 PBq

0.02 PBq

399

400

UNSCEAR 2008 REPORT: VOLUME I

Figure XXXI. Locations where radioactive waste was dumped in the Kara Sea [I11]

Reactors
containing
fuel

Reactors
withoutfuelSolid
low-level
waste

Liquidlow-level
waste

Figure XXXII. Spacecraft missions utilizing nuclear and/or radioactive material
RHU: radioisotope heating unit
6

NUMBER OF DEVICES

5
4
RHU
3

Reactor
RTG

2
1
0
1960

1970

1980

1990
YEAR

2000

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXXIII.
of devices

401

Current status of devices utilizing nuclear and/or radioactive material in space: number of missions and number

50

NUMBER OF MISSIONS

40

30

RHU
RTG

20

Reactor100Long-term On the Moon
storage orbit

On Mars

On Jupiter

On Saturn

Other orbits

Re-entered
Earth’s
atmosphere

140

NUMBER OF DEVICES

120
100
80

RHU
RTG

60

Reactor40
20
0Long-term On the Moon
storage orbit

On Mars

On Jupiter

On Saturn

Other orbits

Re-entered
Earth’s
atmosphere

402

UNSCEAR 2008 REPORT: VOLUME I

Figure XXXIV. Estimated 137Cs deposition density (Bq/m2) from NTS fallout (top figure) and from global fallout (bottom figure)across the continental United States [S23]

1 200
1 100
1 000
900
800
700
600
500
400
300
200
100
0

9 000
8 000
7 000
6 000
5 000
4 000
3 000
2 000
1 000
0

Figure XXXV. Sites of peaceful underground nuclear explosions in the former Soviet Union
Purposes of explosions: 1, obtaining seismic profiles; 2, creation of reservoirs in salt; 3, oil extraction; 4, creation of oil and gas inflow; 5, liquidation of oil wells; 6, burial of liquid toxic waste; 7, ore
crushing; 8, creation of underground storage facilities in clay; 9, ground excavation; 10, ground loosening [V1]

1
11

1

1

2

6
6
11 33
2

33

5

1
1

11
1

33 9

4
8

1
1

1

11

10
1

1

1

1

1

1 3

1

11

144
2

1
1

1

1

1

1

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

77

403

404

UNSCEAR 2008 REPORT: VOLUME I

Figure XXXVI. Estimated contributions to public exposure from different sources for different countries, and UNSCEAR
estimates of worldwide average exposures
Figures for the United States from references [M23, N8], for Germany from [B49], for the United Kingdom from [W6]. Different distributions
can be expected for other countries, as all countries considered here have a high level of development. For Germany, “Other” includes exposuredue to fallout resulting from nuclear tests, to the Chernobyl accident and to releases from nuclear power plants
GLOBAL (UNSCEAR 2000)
0.4
0.01

GLOBAL (UNSCEAR 2008)
0.6
0.01

0.3
0.29

1.2
0.5

1.26

0.48

0.39

0.4

UNITED STATES (1987)
0.54

UNITED STATES (2006)

0.11

3

0.04
0.4

0.11
0.04

0.29

1.98

0.4

0.29

1.98
0.29
0.29

GERMANY (2005)

UNITED KINGDOM (2005)
0.41

1.9

0.1

0.01

0.25
0.3
0.35
0.4

1.3

0.04
0.3

Radon

0.33

1.1

CosmicExternalterrestrial

Ingestion

MedicalConsumerproducts

Other

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

405

Figure XXXVII. Example of different dose distributions affecting public exposure in the United States(a) Altitude and latitude effects on cosmic radiation dose [U26]; (b) external gamma exposure [U28]; (c, d, e) distributions of the natural
terrestrial radionuclides Th, U and K, which contribute to ingestion and inhalation doses [U28]; (f) indoor radon, main contributor to public
exposure from natural sources via inhalation [U28]; (g) doses from fallout resulting from nuclear tests [N6]; (h) location of nuclear power plants[U39]. All these source distributions would be combined with (i), population distribution [U28]

(a)

(b)

(c)

(d)

(e)

(f )

(g)

(h)

(i)

406

UNSCEAR 2008 REPORT: VOLUME I

Figure XXXVIII. Worldwide trends in occupational exposure due to uranium mining
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
350

MONITORED WORKERS �103�

300
250
200
150
100
50
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

1 800
1 600
1 400
1 200
1 000
800
600
400
200
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

6
5
4
3
2
1
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XXXIX. Worldwide trends in occupational exposure due to uranium milling
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

25

20

15

10

5

0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

12
10
8
6
4
2
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD

407

408

UNSCEAR 2008 REPORT: VOLUME I

Figure XL. Worldwide trends in occupational exposure due to uranium enrichment and conversion
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
20

MONITORED WORKERS �103�

18
16
14
12
10
8
6
4
2
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

6
5
4
3
2
1
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

0.5

0.4

0.3

0.2

0.1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XLI. Worldwide trends in occupational exposure due to nuclear fuel production
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

30
25
20
15
10
5
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

40
35
30
25
20
15
10
5
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

2

1.5

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD409

410

UNSCEAR 2008 REPORT: VOLUME I

Figure XLII. Worldwide trends in occupational exposure due to reactor operation
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
350

MONITORED WORKERS �103�

300
250
200
150
100
50
0

PWR

BWR

HWR

GCR

LWGR

GCR

LWGR

GCR

LWGR

REACTOR TYPE

COLLECTIVE EFFECTIVE DOSE �man Sv�

600
500
400
300
200
100
0

PWR

BWR

HWR

AVERAGE EFFECTIVE DOSE �mSv�

REACTOR TYPE
14
13
12
11
10
9
8
7
6
5
4
3
2
1
0

PWR

BWR

HWR
REACTOR TYPE

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

411

NORMALIZED COLLECTIVE EFFECTIVE DOSEPER UNIT ELECTRICAL ENERGY�man Sv per GW a�

Figure XLIII. Worldwide trends in collective effective dose due to reactor operation, and in normalized collective effective
dose per reactor and per unit electrical energy
PWR
BWR
HWR
GCR
LWGR
All

25

20

15
10

5

0

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

NORMALIZED COLLECTIVE EFFECTIVE DOSEPER REACTOR �man Sv�

PERIOD
PWR
BWR
HWR
GCR
LWGR
All

10
9
8
7
6
5
4
3
2
1
0

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

PERIOD
1 200

LWGR

AVERAGE ANNUAL COLLECTIVE
EFFECTIVE DOSE �man Sv�

GCR
1 000

HWR
BWR

800

PWR

600
400
200
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD

1995–1999

2000–2002

412

UNSCEAR 2008 REPORT: VOLUME I

Figure XLIV. Worldwide trends in occupational exposure due to fuel reprocessing
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
80

MONITORED WORKERS �103�

70
60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

80
70
60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

8

6

4

2

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XLV. Worldwide trends in occupational exposure due to research related to the nuclear fuel cycle
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
140

MONITORED WORKERS �103�

120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

180
160
140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

2

1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD413

414

UNSCEAR 2008 REPORT: VOLUME I

Figure XLVI. Worldwide trends in the number of monitored workers, and in collective effective doses and effective doses toworkers for different practices of the nuclear fuel cycle

MONITORED WORKERS �103�

600
500
400
300
200
100
0

MiningMilling

Enrichment

Fuelfabrication

Reactoroperation

Reprocessing

Research

Reactoroperation

Reprocessing

Research

Reactoroperation

Reprocessing

Research

COLLECTIVE EFFECTIVE DOSE �man Sv�

PRACTICE
1 800
1 600
1 400
1 200
1 000
800
600
400
200
0

MiningMilling

Enrichment

Fuelfabrication

AVERAGE EFFECTIVE DOSE �mSv�

PRACTICE
11
10
9
8
7
6
5
4
3
2
1
0

MiningMilling

Enrichment

Fuelfabrication

PRACTICE

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

415

Figure XLVII. Worldwide trends in the number of monitored workers, and in collective effective doses and effective doses toworkers in the nuclear fuel cycle
1 000

MONITORED WORKERS �103�

900
800
700
600
500
400
300
200
100
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

3 500
3 000
2 500
2 000
1 500
1 000
500
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

5

4

3

2

1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD416

UNSCEAR 2008 REPORT: VOLUME I

Figure XLVIII. Dose levels for (a) interventional radiologist and (b) interventional cardiologist
Average values of 83 procedures performed by ten specialists in six laboratories [V7]296 µSv
296 µSv

236 µSv

284 µSv

325 µSv

283 µSv

260 µSv

294 µSv

269 µSv

252 µSv

365 µSv

618 µSv

326 µSv

646 µSv

396 µSv(a)

167 µSv

191 µSv

364 µSv(b)

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure XLIX. Worldwide trends in occupational exposure due to diagnostic radiology
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
8 000

MONITORED WORKERS �103�

7 000
6 000
5 000
4 000
3 000
2 000
1 000
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

4 000
3 500
3 000
2 500
2 000
1 500
1 000
500
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD

417

418

UNSCEAR 2008 REPORT: VOLUME I

Figure L.

Occupational exposures due to diagnostic radiology in Greece for various job categories
Conventional proceduresInterventional procedures(a) Numbers of monitored workers for conventional and interventional procedures
Doctors
1 715 (32%)

Nurses222 (21%)

Others395 (7%)

Technicians
3 284 (61%)

Doctors(orthopaedists)
297 (29%)

Others96 (9%)

Doctors(cardiologists)
422 (41%)(b) Collective effective doses for conventional and interventional procedures
Others
0.09 man Sv

Others
0.06 man Sv
Nurses0.16 man Sv

Doctors0.96 man Sv(c)Technicians
0.35 man Sv

Doctors(orthopaedists)
0.16 man Sv

Doctors(cardiologists)
1.68 man Sv

Average effective doses to monitored workers
Others0.22 mSv

Doctors0.56 mSv

Others
0.67 mSv

Nurses0.70 mSv

Doctors(cardiologists)
3.97 mSv

Technicians
0.11 mSv
Doctors(orthopaedists)
0.54 mSv(d) Average effective doses to measurably exposed workers
Others1.66 mSv

Others1.16 mSv

Nurses1.42 mSv

Doctors
3.05 mSv
Technicians
0.83 mSv

Doctors(orthopaedists)
1.63 mSv

Doctors(cardiologists)
6.43 mSv

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LI. Worldwide trends in occupational exposure due to dental practice
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

600
500
400
300
200
100
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

0.4

0.3

0.2

0.1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD419

420

UNSCEAR 2008 REPORT: VOLUME I

Figure LII. Worldwide trends in occupational exposure due to nuclear medicine
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
140

MONITORED WORKERS �103�

120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

100
90
80
70
60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

421

Figure LIII. Occupational exposures due to nuclear medicine in Greece for various job categories for the periods 1995–1999
and 2000–2002(1995–1999)(2000–2002)(a) Numbers of monitored workers
Doctors
179 (33%)

Doctors
208 (33%)
Nurses77 (12%)

Nurses38 (7%)

Others60 (11%)

Technicians
261 (49%)

Others71 (11%)

Technicians279 (44%)(b) Collective effective doses
Doctors0.12 man Sv

Doctors0.13 man Sv

Nurses0.05 man Sv

Nurses0.02 man Sv
Others0.02 man Sv

Technicians
0.20 man Sv

Technicians
0.23 man Sv(c)Others0.02 man Sv

Average effective doses for monitored workers
Doctors
0.71 mSv
Nurses0.48 mSv

Doctors
0.58 mSv

Others0.33 mSv
Technicians
0.89 mSv

Nurses0.59 mSv

Others0.23 mSv
Technicians
0.73 mSv(d) Average effective doses for measurably exposed workers
Doctors
1.53 mSv
Nurses1.41 mSv

Technicians
2.03 mSv

Others1.38 mSv

Doctors
1.08 mSv

Technicians
1.41 mSv

Nurses1.49 mSv

Others0.75 mSv

422

UNSCEAR 2008 REPORT: VOLUME I

Figure LIV. Worldwide trends in occupational exposure due to radiotherapy
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
140

MONITORED WORKERS �103�

120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

200
180
160
140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

2.5

2

1.5

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LV. Worldwide trends in occupational exposure due to all medical uses of radiation
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
8 000

MONITORED WORKERS �103�

7 000
6 000
5 000
4 000
3 000
2 000
1 000
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

4 000
3 500
3 000
2 500
2 000
1 500
1 000
500
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD423

424

UNSCEAR 2008 REPORT: VOLUME I

Figure LVI. Trends in occupational exposure due to industrial irradiation in China
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
1.6

MONITORED WORKERS �103�

1.4
1.2
1
0.8
0.6
0.4
0.2
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

1.4
1.2
1
0.8
0.6
0.4
0.2
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

1.5

1

0.5

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LVII. Worldwide trends in occupational exposure due to industrial radiography
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
140

MONITORED WORKERS �103�

120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

250

200

150

100

50

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

3

2

1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD425

426

UNSCEAR 2008 REPORT: VOLUME I

Figure LVIII. Trends in occupational exposure due to luminizing in Switzerland
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
0.40

MONITORED WORKERS �103�

0.35
0.30
0.25
0.20
0.15
0.10
0.05
0.00
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

2.5

2.0

1.5

1.0

0.5

0.0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

12

9

6

3

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LIX. Worldwide trends in occupational exposure due to radioisotope production
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
100

MONITORED WORKERS �103�

90
80
70
60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

3

2

1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD

427

428

UNSCEAR 2008 REPORT: VOLUME I

Figure LX. Trends in occupational exposure due to well logging in Canada
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
1.6

MONITORED WORKERS �103�

1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
1975–1979

1980–1984

1985–1989

1990–1994

AVERAGE EFFECTIVE DOSE �mSv�

PERIOD10
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LXI. Trends in occupational exposure due to accelerator operation in Canada
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

1.2
1.0
0.8
0.6
0.4
0.2
0.0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

1.2
1.0
0.8
0.6
0.4
0.2
0.0
1975–1979

1980–1984

1985–1989

1990–1994

AVERAGE EFFECTIVE DOSE �mSv�

PERIOD10
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD429

430

UNSCEAR 2008 REPORT: VOLUME I

Figure LXII. Worldwide trends in occupational exposure due to all industrial uses of radiation
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
1 000

MONITORED WORKERS �103�

900
800
700
600
500
400
300
200
100
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

1 000
900
800
700
600
500
400
300
200
100
0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODAVERAGE EFFECTIVE DOSE �mSv�

2

1

0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LXIII. Worldwide trends in occupational exposure due to uses of radiation in educational establishments
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
500

MONITORED WORKERS �103�

450
400
350
300
250
200
150
100
50
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

80
70
60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

AVERAGE EFFECTIVE DOSE �mSv�

PERIOD

0.5

0.0
1975–1979

1980–1984

1985–1989

1990–1994

PERIOD

431

432

UNSCEAR 2008 REPORT: VOLUME I

Figure LXIV. Worldwide trends in occupational exposure due to uses of radiation in veterinary medicine
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
180

MONITORED WORKERS �103�

160
140
120
100
80
60
40
20
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

60
50
40
30
20
10
0
1975–1979

1980–1984

1985–1989

1990–1994

AVERAGE EFFECTIVE DOSE �mSv�

PERIOD

0.5

0.0
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

433

1

0.8

0.6
<1 mSv
1-5 mSv

0.4>5 mSv
0.2

gh
w
ay
Hi

ilw
ay
Ra

t
Po
r

tu
re
r
de
pa
rtm
en
t
Ho
sp
ita
l
Un
iv
er
sit
y
Ai
rc
ar
go

r

uf

ac

ke
M

an

Tr
uc

rie

r

0

Co
u

FRACTION OF ANNUAL DOSES IN THREE RANGES

Figure LXV. Fraction of annual doses in three dose ranges for all companies participating in study of occupational doses due
to transport in Canada [E2]

434

UNSCEAR 2008 REPORT: VOLUME I

Figure LXVI. Worldwide trends in occupational exposure due to military activities
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers
450

MONITORED WORKERS �103�

400
350
300
250
200
150
100
50
0
1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

1995–1999

2000–2002

1995–1999

2000–2002

PERIOD

COLLECTIVE EFFECTIVE DOSE �man Sv�

450
400
350
300
250
200
150
100
50
0
1975–1979

1980–1984

1985–1989

1990–1994

AVERAGE EFFECTIVE DOSE �mSv�

PERIOD10
1975–1979

1980–1984

1985–1989

1990–1994

PERIODANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

435

Figure LXVII. Worldwide trends in numbers of monitored workers, and in collective effective doses and effective doses tomonitored workers

MONITORED WORKERS �103�

14 000
12 000
10 000
8 000
6 000
4 000
2 000
0

Naturalradiation

Nuclear fuelcycle

Medicaluses

Industrial
uses

Military
activities

Miscellaneous

Industrial
uses

Military
activities

Miscellaneous

Industrial
uses

Military
activities

Miscellaneous

1995–1999

2000–2002

COLLECTIVE EFFECTIVE DOSE �man Sv�

SECTOR
40 00035 00030 00025 000
20 000
15 000
10 0005 000
0Naturalradiation

Nuclear fuelcycle

Medicaluses
SECTORAVERAGE EFFECTIVE DOSE �mSv�

5
4
3
2
1
0

Naturalradiation

Nuclear fuelcycle

Medicaluses
SECTOR

1975–1979

1980–1984

1985–1989

1990–1994

436

UNSCEAR 2008 REPORT: VOLUME I

Figure LXVIII. Worldwide trends in occupational exposure due to natural sources of radiation
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

8 000
7 000
6 000
5 000
4 000
3 000
2 000
1 000
0

Coal mines

Other mines

Processing

Workplaces
other thanmines

Aircrew

Workplaces
other thanmines

Aircrew

Workplaces
other thanmines

Aircrew

COLLECTIVE EFFECTIVE DOSE �man Sv�

SECTOR
18 000
16 000
14 000
12 000
10 000
8 000
6 000
4 000
2 000
0

Coal mines

Other mines

Processing
SECTORAVERAGE EFFECTIVE DOSE �mSv�

6
5
4
3
2
1
0

Coal mines

Other mines

Processing

SECTOR

1990–1994

1995–1999

2000–2002

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

Figure LXIX. Worldwide trends in occupational exposure due to man-made sources of radiation
Average annual numbers of monitored workers, and collective effective doses and effective doses to monitored workers

MONITORED WORKERS �103�

8 000
7 000
6 000
5 000
4 000
3 000
2 000
1 000
0

Nuclear fuelcycle

Medicaluses

Industrial
uses

Military
activities

Miscellaneous

Military
activities

Miscellaneous

Military
activities

Miscellaneous

COLLECTIVE EFFECTIVE DOSE �man Sv�

SECTOR
4 000
3 500
3 000
2 500
2 000
1 500
1 000
500
0

Nuclear fuelcycle

Medicaluses

Industrial
uses
SECTORAVERAGE EFFECTIVE DOSE �mSv�

5
4
3
2
1
0

Nuclear fuelcycle

Medicaluses

Industrial
uses
SECTOR

1975–1979

1980–1984

1985–1989

1990–1994

1995–1999

2000–2002

437

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATION

439

References
PART A
Responses to the UNSCEAR
Global Survey of Public Radiation Exposuresand Global Survey of Occupational Radiation Exposures
Country

Respondent

Albania

Kostandin Dollani. Institute of Nuclear Physics, Tirana

Argentina

Adriana Curti and Abel González. Nuclear Regulatory Authority, Buenos Aires

ArmeniaA. Martirosyan. Armenian Nuclear Regulatory Authority, Yerevan

Australia

Peter Thomas and Owen Wilson. Personal Radiation Monitoring Service, Australian Radiation Protection and Nuclear
Safety Agency, Yallambie, Victoria

Austria

J. Fuchs. Physikalisch-Technische Prüfanstalt für Radiologie und Elektromedizin, Vienna
V. Karg. Federal Ministry of Agriculture, Forestry, Environment and Water Management, Vienna
H. Stadtmann. Health Physics Division, ARC Seibersdorf Research GmbH, Seibersdorf
H.M. Spreizer and E.M. Plantosar. Staatlich akkreditierte Prüfstelle für Strahlenschutz, LKH-Univ. Klinikum, Graz
T. Schöpf. Institut für medizinischen Strahlenschutz und Dosimetrie am Landeskranken – Universitätskliniken – Innsbruck

Azerbaijan

A.K. Qazibov. National Academy of Sciences, Baku

Bangladesh

M.N. Alam and Mantazul Islam Chowdhury. Radioactivity Testing and Monitoring Laboratory, BAEC, Chittagong
Kazi Obaidul Awal. Nuclear Safety and Radiation Control Division, BAEC, Dhaka
Mahfuza Begum. Bangladesh Atomic Energy Commission, Dhaka

BelarusNataliya Mahotina. Republic Centre of Hygiene, Epidemiology and Public Health, Minsk

Belgium

Hans Vanmarcke. SCK-CEN, Nuclear Research Centre, Boeretang, Mol

Bulgaria

Radostina Georgieva, Vanda Rupova and G. Vasilev. National Centre of Radiobiology and Radiation Protection, Ministry of
Health, Sofia
Radelina Tranteeva. Safety Management Department, Kozloduy Nuclear Power Plant, Kozloduy

CanadaSonia Jonson and Willem Sont. Radiation Protection Bureau, Health Canada, Ottawa
Kevin Bundy, Bernie Ellaschuk, Norman Gentner and D.W. Whillans. Canadian Nuclear Safety Commission, Ottawa

Chile

J.E. Gamarra Chamorro. Comisión Chilena de Energía Nuclear, Santiago

China

Ziqiang Pan. Committee on Science and Technology, China National Nuclear Corporation, Beijing
Jizeng Ma. Department of Health Physics and Nuclear Safety, China Institute for Radiation Protection, Taiyuan

China (Taiwan)

Ching-Chung Huang. Radiation Monitoring Center (AEC), Kaohsiung
Hsueh-Li Yin. Department of Radiation Protection, Atomic Energy Council, Yonghe City

Costa Rica

Patricia Mora. Dosimetry Section, Nuclear Physics Laboratory, University of Costa Rica, Atomic and Nuclear Sciences
Research Center, San Pedro de Montes de Oca

Croatia

Gordana Marović. Institute for Medical Research and Occupational Medicine, Zagreb
Nikša Sviličić. State Office for Radiation Protection, Zagreb

Cuba

Juan Tomás Zerquera. Center for Radiation Protection and Hygiene, Havana

CyprusStelios Christofides. Medical Physics Department, Nicosia General Hospital, Nicosia

Czech Republic

Jiri Hulka, Josef Thomas, Irene Malátová and Ladislav Tomasek. National Radiation Protection Institute, Prague
František Spurný. Nuclear Physics Institute ASCR, PragueZdeněk Prouza, Karla Petrova and Miluse Budayova. State Office for Nuclear Safety, PragueBožena Jurochová and Jiří Pařízek. Dukovany Nuclear Power Plant, DukovanyJiří Čumpl and Zdeněk Gregor. DIAMO, s.p.,o.z. GEAM, Dolní Rožinka
Juraj Štofánik. DIAMO, s.p., o.z. TÚU, Stráž pod Ralskem
Josef Koc. Temelín Nuclear Power Plant, Temelín

Denmark

Henrik Roed, Kaare Ulbak and J. Munk. National Board of Health, National Institute of Radiation Protection, Herlev

440

UNSCEAR 2008 REPORT: VOLUME ICountry

Respondent

El Salvador

Ronald Enrique Torres Gómez. Ministerio de Salud Publica y Asistencia Social, San Salvador

Estonia

Merle Lust. Estonian Radiation Protection Centre, Tallinn

Finland

Hannu Arvela, Maaret Lehtinen, Päivi Kurttio and Eija Vartiainen. Radiation and Nuclear Safety Authority—STUK, Helsinki

FranceAnne Flüry-Herard. DSV/DIR/CARMIN et cab. HC, Fontenay-aux-Roses

Germany

T. Beck, E. Ettenhuber, G. Frasch and L. Kammerer. Federal Office for Radiation Protection, Berlin

GreeceVasiliki Kamenopoulou and Virginia Koukouliou. Greek Atomic Energy Commission, Aghia Paraskevi
Eleftheria Carinou. Department of Dosimetry, Greek Atomic Energy Commission, Aghia Paraskevi

Hungary

Andor Kerekes. Department of Radiohygiene, National Research Institute for Radiobiology and Radiohygiene, Budapest
Nándor Ötvös. Hungarian Atomic Energy Authority, Budapest

IcelandSigurður Emil Pálsson and Gudlaugur Einarsson. Icelandic Radiation Protection Institute, Reykjavik

Indonesia

Dadong Iskandar. Center for Technology of Radiation Safety and Metrology, National Nuclear Energy Agency, Jakarta
Kunto Wiharto. Centre for Research and Development of Radiation Safety and Nuclear Bio-Medicine, National Nuclear
Energy Agency, Jakarta

Iran, Islamic Republic of

M. Ghiassi-Nejad. National Radiation Protection Department, Iranian Nuclear Regulatory Authority

Iraq

Souad Mohammad. Radiation Prevention Centre, Baghdad

IrelandCatherine Organo, Lorraine Currivan and Tony Colgan. Radiological Protection Institute of Ireland, Dublin

Italy

Francesco Bochicchio and Serena Risica. Istituto Superiore di Sanità, Roma
R. Rusconi. ARPA Lombardia, Milano

Japan

Japan Expert Panel for UNSCEAR. Regulatory Sciences Research Group, National Institute of Radiological Sciences,
Chiba
Masahiro Doi, Yoshiharu Yonekura and Shinji Yoshinaga. National Institute of Radiological Sciences, Chiba
Masanao Nakano. Japan Atomic Energy Agency, Ibaraki

Kazakhstan

German Fyodorov. Atomic Energy Committee, Almaty

KuwaitDhari Al Ajmi and Abdulhadi S. Al-Otaibi. Environment and Urban Development Division, Kuwait Institute for Scientific
Research, Safat

LatviaDmitrijs Lvovs. TLD Sector, Radiation Safety Centre, Riga

Libyan Arab Jamahiriya

Mohamed Ahmed Shenber. Environmental Radiation Measurements Division, Radiation Protection and Health Physics
Department, Tajoura Nuclear Research Centre, Tripoli

Liechtenstein

Manfred Frick. Office for Environmental Protection, Vaduz

Lithuania

B. Gricienė, G. Morkūnas, D. Jankauskiene, L. Pilkytė, A. Urbonienė and V. Vlaskin. Radiation Protection Centre, Vilnius
B. Vilimaitė-Šilobritienė. Radiology Division, Environmental Protection Agency, Vilnius

Luxembourg

Patrick Breuskin, Marielle Lecomte, Alexandra Schreiner and Ferid Shannoun. Department of Radiation Protection,
Ministry of Health

Malta

Paul Brejza. Radiation Protection Board, Pietà

Mauritius

Reza Pooloo. Physics Section, Ministry of Health, Victoria Hospital, Quatre Bornes

MexicoHermenegildo Maldonado, Mario Mejía López and O. Aguilar. National Commission of Nuclear Safety and Safeguards,
Mexico City

Montenegro

P. Vukotic and T. Andjelic. Center for Ecotoxicological Researches of Montenegro, Montenegro

Myanmar

Lwin Lwin Wai. Department of Atomic Energy, Yangon

Netherlands

J.W.E. van Dijk. National Dose Registration and Information System, NRG, Ministry of Social Affairs and Employment,
Arnhem
C.P. Tanzi. National Institute for Public Health and the Environment, BA Bilthoven

New ZealandAndrew McEwan and Matthew Aileone. National Radiation Laboratory, Christchurch

NorwayGudrun Uthaug Paulsen. Norwegian Radiation Protection Authority, Østerås

ANNEX B: EXPOSURES OF THE PUBLIC AND WORKERS FROM VARIOUS SOURCES OF RADIATIONCountry441

Respondent

Oman

Ministry of Regional Municipalities, Environment and Water Resources

Peru

J.F. Márquez Pachas. Instituto Especializado de Enfermedades Neoplasicas, Lima
R. Ramírez Quijada. Instituto Peruano de Energía Nuclear, Lima

Philippines

Teresa Nazarea and Eliza Enriquez. Health Physics Research Section, Philippine Nuclear Research Institute, Quezon City
Estrella Caseria. Radiation Protection Services, Philippine Nuclear Research Institute, Quezon City

PolandA. Koczynski, J. Henschke and M. Biernacka. Central Laboratory for Radiological Protection, Warsaw

Republic of Korea

Ki-Juung Lim. Korea Radioscope Association, Seoul
Saeng-Ki Kim. Korea Hydro and Nuclear Power Co. Ltd., Seoul
Dong Myung Lee. Korea Institute of Nuclear Safety, Daejeon

Romania

A. Baciu. National Commission for Nuclear Activities Control, Bucharest
E. Botezatu and O. Iacob. Institute of Public Health, Iasi

Slovakia

Emil Bédi. Public Health Authority of the Slovak Republic, Bratislava

Slovenia

Michel Cindro and Milko Križman. Slovenian Nuclear Safety Administration, Ljubljana
Nina Jug. Slovenian Radiation Protection Administration, Ljubljana

Spain

J. Arteche, I. Fuente, J. Gómez, J.L. Martin Matarranz, J. Nicolás, L. Quindós, L.S. Quindós Poncela, C. Sainz. Grupo
Radon, University of Cantabria, Santander, Cantabria
Sara González Veci and Juan Carlos Lentijo. Nuclear Safety Council, Madrid
María José Barahona and José Luís Martín. Consejo de Seguridad Nuclear

SwedenPål Andersson, K. Bergman, T. Cederlund, Ingemar Lund, Maria Lüning and Leif Nyblom. Swedish Radiation Protection
Authority, Stockholm

Switzerland

Christophe Murith. Radiological Risks Section, Division of Radiation Protection, Swiss Federal Office of Public Health, Bern
Sybille Estier. Environmental Radioactivity Section, Division of Radiation Protection, Swiss Federal Office of Public Health,
Bern
Daniel Frei. Division of Radiation Protection, Swiss Federal Office of Public Health, ,Bern
Johannes Hammer. Swiss Federal Nuclear Safety Inspectorate, Villigen

Syrian Arab Republic

M.H. Kharita and Atef Al Bazal. Atomic Energy Commission, Damascus

The former Yugoslav Republic of Macedonia

Lidija Nikolovska and Rumen Stamenov. Radiation Safety Directorate, Skopje

Trinidad and Tobago

Sue-Jaan Mejias. National Oncology Programme, Ministry of Health, Port of Spain
Rohit Doon. Ministry of Health, Port of Spain

Tunisia

Latifa Ben Omrane and Sadok Mtimet. Centre National de Radioprotection, Tunis

TurkeyLevent Aksu. Turkish Atomic Energy Authority, Ankara Nuclear Research and Training Centre Dosimetry Group, Ankara

UkraineAla Govor. Radiation Protection of Population and Ecology Division, Executive Committee of the Town Council, Zhovty Vody
Proskura Mukola Ivanovich. Ministry of Ukraine for Emergencies and Protection of the Population from the Consequencesof the Chernobyl Catastrophe
Oleg A. Karpenko. Statistical Section for Municipal Environmental Office, Dneprodzerzinsk City
Olena Bober. Safety of Radiation Technology Department, State Nuclear Regulatory Committee of Ukraine
V. Gryschenko. State Nuclear Regulatory Committee of Ukraine
G. Burtak. Ministry of Health of Ukraine

United Kingdom

J.S. Hughes and Richard Paynter. Health Protection Agency, Chilton, Didcot

United Republic of Tanzania

W.E. Muhogora. National Radiation Commission, Arusha

United States

D. Miller. Department of Nuclear Engineering, University of Illinois, Urbana-Champaign
E. Vincent Holahan. Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington D.C.
N.H. Harley. Department of Environmental Medicine, New York University School of Medicine

442

UNSCEAR 2008 REPORT: VOLUME I

PART B
A1
A2
A3

A4

A5

A6

A7
A8

A9

A10

A11

A12

A13

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