HUNDREDS IF not thousands of radiation accidents have occurred in nuclear facilities, as well as in industrial, medical, research, and academic facilities, that make use of sealed radioactive sources or of machine-generated radiation. A detailed survey of accidents “whereby exposure to radioactive material affected workers or members of the public in a fashion that results in acute (i.e., deterministic) health effects” was conducted by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR 2011a), and the medical management of such types of radiation accidents was described extensively by Gusev et al. (2001). This paper will present and discuss the increased occupational doses that were received as a consequence of severe reactor accidents, defined as those that resulted in irreparable damages to the plant (Windscale in the United Kingdom in 1957; Three Mile Island in the United States in 1979; Chernobyl in the former Soviet Union in 1986; Fukushima in Japan in 2011). The facilities of the “military” fuel cycle [i.e., the reactors used for the production of plutonium and the plants involved in the processing of plutonium (Hanford in the United States and Mayak in the former Soviet Union)] have been considered as well, even though the increased doses that were received in the early years of operation in the late 1940s and early 1950s at Mayak were not due to accidents but to the conduct of routine operations.
External doses to workers are usually derived from the reading or processing of personnel dosimeters. Over the course of the years, the external doses have been reported in terms of exposure, absorbed dose, or dose equivalent. For regulatory practices, the operating principle is to estimate the dose conservatively in order to protect the worker. However, when the doses are to be used in the framework of epidemiologic studies, care must be taken to ensure that unbiased estimates of individual dose are obtained. In that respect, considerable efforts have been made to evaluate the validity of the reported doses resulting from the Chernobyl nuclear reactor accident and from the early operations of the Mayak Production Association (MPA). In this paper, the doses recorded for regulatory practice are presented in terms of dose equivalent (sievert or millisievert); when the doses are used for epidemiologic purposes, they are presented in terms of absorbed dose (gray or milligray).
Substantial internal doses were also received by some of the workers at Mayak, as well as those involved in emergency and cleanup activities related to the Chernobyl and Fukushima accidents. The internal occupational doses are estimated retrospectively and are derived from the analysis of the available environmental and bioassay measurements. Based on the experience of Mayak and Chernobyl, the assessment of the internal doses improves gradually with time and takes a number of years to be implemented satisfactorily. In this paper, the internal doses are expressed in terms of absorbed dose (gray or milligray), except for Fukushima, where the use of effective doses seemed to be more appropriate.
Following a brief summary of the average annual occupational doses that have been reported worldwide for routine operations in facilities of the nuclear fuel cycle, a detailed presentation of the increased exposures among the workers involved in early operations of plutonium-producing facilities and in the emergency and mitigating activities related to the four major nuclear reactor accidents will be made, and the efforts made to improve the dose estimates for the purposes of epidemiologic studies will be described.
AVERAGE ANNUAL DOSES TO WORKERS DUE TO ROUTINE OPERATION OF THE NUCLEAR FUEL CYCLE
Doses to workers in the whole range of facilities and activities involving radiation exposure have been compiled and presented in a number of reports, notably by the National Council on Radiation Protection and Measurements (NCRP 1989) and UNSCEAR (1993, 2010). For epidemiologic purposes, occupational doses covering a large number of reactor workers were collected and analyzed also (Cardis et al. 2007; Thierry-Chef et al. 2007). In an ongoing study, Boice (2012) is collecting and analyzing the dosimetric information for about one million U.S. workers who were exposed to radiation in facilities of various types or a as a result of work in contaminated areas.
With regard to the nuclear fuel cycle, which includes the facilities that are covered in this paper, UNSCEAR has reported information periodically on the number of workers involved worldwide in each component of the nuclear fuel cycle and on the related average annual effective doses. The results for two time periods (1975–1979 and 2000–2002) are presented in Table 1 (UNSCEAR 2010). The number of workers in uranium mining and milling decreased from one time period to the other, while the number of workers involved in reactor operations increased and the workers in other components of the nuclear fuel cycle were relatively constant. The average annual effective doses during the 1975–1979 time period were <5 mSv in uranium enrichment, fuel fabrication, and reactor operation, and between 5–10 mSv in uranium mining, uranium milling, and fuel reprocessing; during the 2000–2002 time period, average annual effective doses had decreased to <2 mSv in all components of the nuclear fuel cycle.
INCREASED EXPOSURES DURING ROUTINE OPERATION: EARLY YEARS
At the beginning of the nuclear era in the 1940s, nuclear energy was developed for military purposes. It was deemed interesting to present the occupational doses at two sites with very similar activities: Hanford in the United States and Mayak in the former Soviet Union.
The Hanford site, located in south-central Washington State, was selected in 1943 for the production of plutonium and other nuclear materials in support of the World War II effort. Reactor operation started in 1944. Over several years, nine nuclear reactors were constructed for the production of plutonium (Shipler et al. 1996), which was used first for the Trinity nuclear weapons test of July 1945. A few thousand workers were monitored for radiation at the Hanford site during the mid- and late-1940s. The personnel dosimetry program at Hanford has been described and evaluated in detail by Wilson et al. (1990). The predominant source of exposure, which was external irradiation from high-energy photons (>100 keV), was judged to have been measured adequately for all years of operation. From 1944–1957, a film badge dosimeter was used. In 1957, a multi-element film badge dosimeter was introduced. This led to a significant improvement in measurement of low-energy photons, the dose from which had previously been underestimated. In 1972, the film badges were replaced with thermo-luminescent dosimeters, thus allowing for the estimation of the neutron doses, which had also been underestimated until that time.
Following the detonation of nuclear weapons by the United States in 1945, large efforts were undertaken by the former Soviet Union to develop their own nuclear weapons technology. The MPA, which is located in the Southern Urals, was the first industrial complex in the former Soviet Union that was built for the production of plutonium. The initial industrial complex included nuclear reactors, a radiochemical plant for separation of plutonium from irradiated fuel, and a plutonium production plant (Vasilenko et al. 2007). The construction of the first reactor began in 1945 and became operational in 1948. The construction of the chemical processing plant began in 1946, and plutonium ready for use in nuclear weapons was available in 1949. Approximately 20,000 workers were exposed to radiation in the early years of operation. Occupational doses were due to external irradiation, mainly from photons, and internal irradiation arising from intakes of plutonium. External exposures were monitored by means of film badges, which were without filters from 1948–1953 and not suitable for appropriately measuring photons with energies below 400 keV. A compensating filter (0.75 mm of lead) was added to the film badges to measure external doses from 1954–1960. In 1961, a 0.5-mm aluminum filter was added to the lead filter (Romanov et al. 2002).
Estimates of annual effective doses from external irradiation to workers at the Hanford Works (Buschbom and Gilbert 1993) and at MPA (Romanov et al. 2002; Vasilenko et al. 2007) are shown in Table 2 and in Fig. 1. For comparison purposes, it was assumed that the reported doses, which were expressed as whole-body penetrating doses (Hanford) and tissue-equivalent dose in free air (MPA), are approximately equal to the effective doses. The differences in the magnitude of the average annual effective doses from external irradiation at the Hanford Works and at MPA are striking. Annual doses at the Hanford Works were generally <5 mSv, and there were only three workers with annual doses >50 mSv during the early years: one in 1947 with an annual effective dose of 60 mSv, another one in 1951 with a dose of 55 mSv, and the third one in 1954 with an annual dose of 144 mSv. By contrast, at MPA, the average annual effective dose was ∼1,000 mSv or so around 1950 and did not decrease to 10 mSv until 1970. Reasons for these high doses from external irradiation at MPA include:
1. the fact that reactor, chemical processing, and plutonium chemical-metallurgical facility technologies were emerging rapidly;
2. there were limitations in MPA resources and capabilities to protect workers; and
3. there was poor understanding of the consequences of relatively high occupational radiation doses.
It is worth noting that important efforts have been made during the last 20 y to reconstruct the doses to the MPA workers, resulting in a database called “Doses-2005” (Vasilenko et al. 2007). It is now established that the initial doses shown in Table 2 and Fig. 1 were mostly overestimated when the film badges without filtration were used during 1948–1953, in some areas by a factor of 2.8. With regard to the Hanford Works, it seems that the recorded doses are biased to some extent but that they are reasonably adequate for the purposes of epidemiologic studies (Gilbert and Fix 1995).
Another important feature of the working situation at MPA is the high internal doses that were received as a result of plutonium intakes. Cumulative organ doses to workers were calculated using the Doses-2005 internal dosimetry model (Vasilenko et al. 2007). As expected, the greatest cumulative doses were estimated for the lung (mean of 205 mGy, median of 23.5 mGy, 53 workers with cumulative doses >500 mGy), liver (mean of 284 mGy, median of 42.2 mGy, 77 workers with cumulative doses >500 mGy), and bone surface (mean of 1040 mGy, median of 156 mGy, 243 workers with cumulative doses >500 mGy).
Four major nuclear reactor accidents, which resulted in irreparable damages to the plant, occurred in four different countries: the United Kingdom in 1957 (Windscale), the United States in 1979 (Three Mile Island), the former Soviet Union in 1986 (Chernobyl), and Japan in 2011 (Fukushima). In each case, the type of reactor was different, and the main cause of the accident was, to some extent, different.
The Windscale reactors, also called “Piles,” used uranium metal as fuel, were moderated by graphite, and were air-cooled. Their main purpose was the production of plutonium for the U.K. atomic weapons program. The heat produced by nuclear fission was not used to generate electricity. Pile No. 1, on which the accident happened, was operational in October 1950 (Wakeford 2007). Since the potential energy stored in the graphite (the Wigner energy) needed to be released in a controlled manner, anneals were organized periodically. Unfortunately, during the ninth anneal, a fire broke out in the reactor core on 10 October 1957 and resulted in a partial core meltdown (Arnold 2007).
Within the framework of an epidemiologic study, McGheoghegan and Binks (2000) collected the recorded external doses relative to the 471 workers who were involved in fire activities. For October 1957, the median of the recorded dose was 4.5 mSv, while the 95th percentile was 16 mSv, and the maximum dose was 44 mSv. Over a 3-mo period encompassing the time of the accident, 14 workers received a dose >30 mSv, the highest dose being 47 mSv. No information could be found in the open literature on the occupational exposures related specifically to the cleanup activities.
Three Mile Island accident
Unit 2 of the Three Mile Island reactor was a pressurized-water reactor with an installed capacity of 906 MW of electricity. It had started operating in December 1978. An accident occurred on 28 March 1979, with failures in the non-nuclear secondary system, followed by a stuck-open relief valve in the primary system, which allowed loss of coolant to occur and resulted in melting at least 45% of the reactor core.
Detailed information on the occupational doses related to the accident and to the cleanup of the damaged reactor could not be found in the open literature. Occupational external doses, which were reported together for the workers of Units 1 and 2 until 1985, show for 1979 an average of 3.5 mSv among the 3,975 workers with measurable doses and a maximum dose of 45 mSv (Table 3). Because the containment of the reactor had held up, there was no urgency to clean up the reactor. After 6 y of preparation, defueling began in October 1985, and decontamination activities took place until December 1993. About 1,000 workers were involved in those operations. The reported average annual external doses for the workers with measurable doses over the 1979–1993 time period are presented in Table 3; they were a few millisievert each year during that time period.
Chernobyl nuclear reactor accident
The reactor was a graphite-moderated, light water-cooled system known as RBMK-1000. With an installed electrical generating capacity of 1,000 MW, it was used to produce electricity for commercial purposes; it had started operating in December 1983. The accident occurred on 26 April 1986 during a low-power engineering test of Unit 4. Improper, unstable operation of the reactor, which had design flaws, allowed an uncontrollable power surge to occur, resulting in successive steam explosions, which destroyed the reactor and part of the building in which the reactor core was housed (UNSCEAR 1988, 2000, 2011b). It is the most severe accident that has ever occurred in the nuclear power industry.
With regard to occupational exposure, a distinction is made between the emergency workers, including the persons who were on the site during the day of the accident, and the recovery operation workers, who performed a variety of tasks at the site and in the 30-km zone surrounding the site from 1986–1990. Among the emergency workers, who included 374 reactor staff, 69 firemen, 113 guards, and 10 medical staff, two workers died in the immediate aftermath and 134 reactor staff and firemen suffered from acute radiation sickness (ARS) (Mettler et al. 2007; UNSCEAR 2011b). The distribution of the external doses received by the workers with ARS and their outcomes are shown in Table 4. The external whole-body doses that are presented in Table 4 are based on biological measurements and clinical symptoms, as the dosimeters worn by the reactor personnel were all overexposed, and the firemen were not equipped with dosimeters. The external whole-body doses ranged from 0.8–16 Gy; most of the 28 ARS victims who died within a few months after the accident had received doses between 6.5 and 16 Gy. The skin doses resulting from beta exposures may have been much higher than the external whole-body doses; an evaluation of the skin doses for eight patients showed that the skin doses ranged from 10–30 times the external whole-body doses (Barabanova and Osanov 1990). On the other hand, the internal doses, based on whole-body counting and bioassay measurements performed while the patients were under treatment, were found generally to be much smaller than the external whole-body doses (Mettler et al. 2007; UNSCEAR 2000).
Following the acute emergency phase of the accident, ∼530,000 recovery operation workers were called from 1986 to 1990 to carry out a variety of tasks, including the decontamination of the reactor block and of the reactor site, and the construction of the entombment of the reactor (known as object Shelter, sarcophagus, and Ukrytie) (UNSCEAR 2011b). The enormous scale of the problems that had to be faced necessitated a massive engagement of several ministries of the former Soviet Union, most notably the Ministry of Defense, the Ministry of Atomic Energy, and the Ministry of Medium Machinery. The numbers of recovery operation workers decreased from year to year, from ∼300,000 in 1986 to ∼6,000 in 1990. Altogether, ∼200,000 recovery operation workers were from Ukraine; ∼200,000 also from Russia; ∼100,000 from Belarus; and ∼5,000 from each of the Baltic countries (Estonia, Latvia, and Lithuania). The time spent by the workers on the site was extremely variable but was generally less than a year. The dosimetry of the recovery operation workers, also called cleanup workers or liquidators, proved to be very challenging, in part because of the very large number of workers originating from a variety of organizations. The “official” doses presented in Table 5 are based on doses that were recorded for about half of the recovery operation workers, using the assumption that the mean doses obtained for the workers with recorded doses apply to the entire population of workers. The recorded doses, which represent external irradiation from photons only, were usually obtained by means of one of three methods:
1. reading of a personal dosimeter;
2. group dosimetry: assignment of the same dose to a group of workers performing a given task, based on the reading of a personal dosimeter worn by a member of the group (in some cases, no member of the group, in which case the dose was assigned on the basis of previous experience); or
3. group estimation: crude time-and-motion analysis (measurements of gamma-radiation levels were made at various points of the reactor site, and the dose was estimated as a function of the locations where work was to be done and of the time spent at those locations) (Pitkevitch et al. 1997; UNSCEAR 2000).
As shown in Table 5, the annual averages of the officially recorded external whole-body doses decreased from 146 mGy in 1986 to 96 mGy in 1987 and 43 mGy in 1988 and remained approximately stable in 1989 (41 mGy) and 1990 (47 mGy); the average annual official external whole-body dose over the 1986–1990 time period was 117 mGy. Neither the external doses to the skin nor to the lens of the eye that were due to beta exposure nor the internal doses due to intakes of radionuclides were recorded.
The extent to which the officially recorded whole-body doses are valid has been discussed in many publications (e.g., Ilyin et al. 1995; Pitkevitch et al. 1997; Auvinen et al. 1998; UNSCEAR 2000; Chumak 2007; Kryuchkov et al. 2012), mainly in relation to their usefulness in epidemiologic studies. The quality of the dosimetry depends to a large extent on the worker category and on the way in which his or her dose was estimated. As a result of the broad variety of the tasks related to Chernobyl activities, the population of the recovery operation workers was very heterogeneous. From the point of view of exposure conditions and dosimetric monitoring, the population of recovery operation workers can be divided into the following categories (Chumak 2007; Kryuchkov et al. 2012), presented in Fig. 2 in terms of percentages of the total number of workers. It should be noted that the numbers of workers given below for each category are taken from Kryuchkov et al. (2012) and may not be in complete agreement with those found in other publications (e.g., UNSCEAR 2011b); they give, however, a good idea of the population size of each category of recovery operation workers:
* Witnesses and victims of the accident (∼2,000): the personnel of the Chernobyl Nuclear Power Plant (ChNPP) and of other organizations, who were at the site at the time of the accident or arrived at the site before 30 April 1986. This category includes the emergency workers previously discussed. Doses during the period of exposure were not recorded for those workers. The official doses, when available, are based on clinical and biological monitoring performed while the workers were hospitalized;
* Early liquidators (∼21,600): civilian and military workers, who were used to decontaminate the site and the 30 km zone between 27 April and 31 May 1986. A consequence of the chaotic situation at the ChNPP site during the few weeks following the accident was that all information related to reading personal dosimeters until mid-May 1986 was either inadequate or lost. The official doses, when available, were based on a conservatively applied time-and-motion analysis;
* ChNPP personnel (2,358 in 1986; 4,498 in 1987): professional atomic workers, charged with the control of operations in Unit 4 and the operation of Units 1, 2, and 3. Good quality dosimetric monitoring was established in mid-May 1986. However, all dose records related to May–June 1986, presumably the period of highest radiation doses, were lost and were never recovered. Personnel who started their work after June 1986 have adequate personal dosimetry;
* Sent to ChNPP (∼2,000 in 1986; 3,458 in 1987): personnel temporally assigned to ChNPP from other nuclear power plants. The dosimetry system was identical to that for the ChNPP personnel;
* Sent to the 30 km zone (31,000 in 1986; 32,000 in 1987): the most diverse category of workers, who were involved in Chernobyl cleanup activities on a task-oriented basis and who visited the 30 km zone only for the duration of their mission; to this category belong, for instance, drivers who delivered equipment and supplies, specialists who were engaged to solve specific technological problems, as well as all kinds of inspectors and representatives of authorities, research institutes, etc. The dosimetric monitoring system depended on the task to be performed and was very uneven;
* AC-605 (21,500 in 1986; 5,376 in 1987): the personnel of Administration of Construction (AC) No.605 [specialized enterprise of the Ministry of Medium Machinery, which was established for the purpose of construction of the sarcophagus (“Object Ukrytie”)]. Except for the period before mid-June 1986, the dosimetric management of that category of workers was performed adequately by means of personal dosimeters;
* Ukrytie personnel (several hundreds in 1990): personnel monitoring the sarcophagus. Dosimetric monitoring was controlled by the ChNPP dosimetry service (personal dosimeters);
* IAE personnel (3,521 in 1988): personnel of the Kurchatov Institute of Atomic Energy (IAE) and of other organizations, involved in various activities inside the sarcophagus. The dosimetric monitoring of that category of workers is similar to that of the AC-605 personnel, and, therefore, of high quality;
* Military workers (61,762 in 1986; 63,751 in 1987): the most numerous category of recovery operation workers, involved in most of the decontamination activities (including manual removal of reactor debris from the roofs of ChNPP in September–October 1986 and January–February 1987), as well as the demolition of abandoned villages, transportation of contaminated materials, etc. The military had its own dosimetric monitoring system, where each unit of battalion level had its own radiation protection officer who kept the dose records and conducted liaison with the dosimetrists on the spot. Due to lack of adequate personal dosimeters (wartime dosimeters had a sensitivity threshold of ∼100 mSv) and because of numerous cases of misuse of the dosimeters, the group dosimetry and the group-estimation methods were used for dose assignment to all team members, often numbering several dozen. Both methods were applied in a conservative manner (e.g., overestimated the doses);
* “Combinat” personnel (6,281 in 1987): civilian staff permanently employed to perform and supervise all offsite activities within the 30 km zone (e.g., activities not related to the recovery and operation of the ChNPP itself), including mainly radiation monitoring, handling of radioactive waste, decontamination, and life supporting infrastructure within the 30 km zone. Because of organizational problems, the dosimetric monitoring of that category of workers was largely not conducted in 1986 and part of 1987. The quality and completeness of the dosimetric information regarding the “Combinat” personnel and the visitors to the 30 km zone became adequate in mid-1987; and
* Belarusian workers (∼24,000 in 1986; ∼28,000 in 1987): civilian personnel who worked in the Belarusian part of the 30 km zone. Only 9% of the Belarusian workers have official recorded doses.
In summary, the population of recovery operation workers was extremely heterogeneous, with durations of exposure that could vary from hours to years, locations of work with very low to very high radiation levels, and activities varying from manual removal of reactor debris to working as a cook or in an office. Also, many types of dosimeters (Fig. 3) were used, because the workers came from a number of organizations with their own dosimetric monitoring systems with little coordination. The quality of the official recorded doses was relatively low in 1986 because the authorities were not prepared to monitor a very large number of workers, but it improved with time and was adequate after mid-1987. Unfortunately, the highest doses were received during the first year following the accident.
A number of risk projections and epidemiologic studies have been devoted to the risk of leukemia among recovery operation workers (e.g., Ivanov 2007; Kesminiene et al. 2008; Romanenko et al. 2008). The case control studies sponsored by the International Agency For Research on Cancer (Kesminiene et al. 2008) and by the National Cancer Institute (Romanenko et al. 2008) required the estimation of individual bone marrow doses for all study subjects. Because external doses are not available for about half of the recovery operation workers, and because of doubts regarding the quality of the external doses for some categories of workers, it was decided not to rely on the official recorded doses and to develop a method of dose estimation that could be applied to all study subjects, whether dead or alive, irrespective of their radiation exposure. The method of dose estimation that was selected for the case control studies sponsored by the International Agency For Research on Cancer and by the National Cancer Institute is a sophisticated time-and-motion analysis called RADRUE, which is an acronym for Realistic Analytical Dose Reconstruction with Uncertainty Estimates (Kryuchkov et al. 2009). Briefly, results of exposure rate and nuclide deposition measurements were embedded in RADRUE and were used to derive exposure rates at places where liquidators lived and worked and thus to calculate external dose to bone marrow according to the liquidators’ itineraries. Liquidators’ routes were reconstructed by dosimetry experts familiar with the organization and conditions of work in the 30-km zone based on information obtained through a study questionnaire (Kryuchkov et al. 2009).
The individual bone marrow dose estimates, along with their uncertainties, that were obtained for the 572 subjects of Phase I of the study of leukemia among the Ukrainian recovery operation workers (Romanenko et al. 2008) are presented in Table 6. In that table, the workers have been classified according to the worker categories given above. The individual bone marrow dose estimates vary in a large range, from <0.01 to 3,260 mGy (even in the same category of workers; for example, the military worker category or sent to the 30-km zone category). The variability of the doses can be important in comparing differing results. The overall average bone-marrow dose obtained for the Ukrainian study subjects is 87 mGy, which is somewhat lower than the estimated average whole-body dose of 117 mGy for all recovery operation workers (UNSCEAR 2011b). Kryuchkov et al. (2012) compared the official doses and the doses calculated by RADRUE for the 1986 recovery operation workers from Ukraine, Russia, and the Baltic countries, and concluded that:
* the RADRUE-calculated dose distributions for the red marrow are wider than the distribution of the official doses;
* the mean values of the RADRUE-calculated doses for the 1986 workers from Ukraine, Russia, and the Baltic countries are lower than the mean value of the official doses for the 1986 workers, the average ratio being 0.6; and
* a larger percentage of doses exceeding 250 mGy is predicted by the RADRUE calculations.
Similar conclusions had been reached previously by Ilyin et al. (1995). Also, Littlefield et al. (1998) performed a cytogenetic analysis of lymphocyte cultures from 118 Estonian workers (mean recorded dose of 103 mGy and maximum recorded dose of 250 mGy) and concluded that it is likely that recorded doses for these workers overestimate their average bone marrow doses, perhaps substantially. Finally, Pitkevitch et al. (1997) performed a statistical analysis of the doses registered for the Russian workers and did not find a major evidence of unreliability, but they recommended the use of a time-and-motion analysis to verify the individual values.
In addition to whole-body doses from external gamma irradiation, the workers received doses to the skin and to the lens of the eye from external beta irradiation. In a cohort study of cataracts among Ukrainian Chernobyl recovery operation workers (Worgul et al. 2007), individual doses from beta particles had to be added to the external doses from photons for the 8,607 cohort members. The dose reconstruction process was very complex (Chumak et al. 2007). In summary and in a simplified way, it involved:
* the recalibration of the official doses from photons, which was obtained by means of a comparison of a high-quality set of EPR measurements of workers’ teeth against their official dose records;
* the assessment of the beta-to-gamma dose ratios for a variety of exposure scenarios, which was obtained using Monte Carlo calculations; and
* the distribution of the official doses according to a range of work locations with similar beta-to-gamma dose ratios, which was derived from a questionnaire addressed to the study subjects.
It was found that:
* the official external doses were biased high by a factor of ∼2.2;
* the beta-to-gamma dose ratios varied as a function of time after the accident but were similar for all work locations; and
* the beta-to-gamma dose ratios varied substantially from one worker to another, with ∼32% of the study subjects with beta doses as large as or larger than the gamma doses and ∼56% with beta doses less than half as large as the gamma doses (Chumak et al. 2007).
The overall distribution of the total doses (beta + gamma) is presented in Table 7. The median dose to the lens of the eye was estimated to be in the range from 100–200 mGy, while ∼200 subjects had estimated doses exceeding 700 mGy.
Because of the abundance of 131I and of shorter-lived radioiodines in the environment of the reactor during the accident, the workers who were on the site during the first few weeks after the accident may have received substantial thyroid doses from internal irradiation. Kesminiene et al. (2012) performed a case-control study among 530 workers from Belarus, Russia, Estonia, Latvia, and Lithuania with the objective of evaluating more precisely the relationship between the dose to the adult thyroid from both external irradiation and internally incorporated 131I and risk of thyroid cancer. For each study subject, individual doses to the thyroid were reconstructed by considering the following pathways of exposure: (1) external irradiation from gamma-emitting radionuclides; and (2) internal irradiation arising from the intake of 131I via inhalation of contaminated air or ingestion of contaminated foodstuffs. To estimate the external doses received by the workers during their cleanup missions, the RADRUE method, which had been selected for the leukemia studies (Kesminiene et al. 2008; Romanenko et al. 2008), was also used. The main difficulty resided in the estimation of the thyroid doses resulting from intakes of 131I. Internal doses due to inhalation intake of 131I during the period of work as a liquidator were calculated for six study subjects who worked on the ChNPP site during the first few weeks after the accident. The approach was based on the data on concentration of 131I in air in settlements in the 30-km zone (Prohl et al. 2000) and was validated by measurements of the dose rate near the neck taken in 30 April–5 May 1986 in a group of 624 early liquidators who were not study subjects. In addition to the dose received during their work, Belarusian workers who were residents of contaminated settlements of the Gomel and Mogilev oblasts and were returning home every evening or after weekly shift work may also have received substantial dose to the thyroid from 131I through consumption of locally produced contaminated milk and/or vegetables. The so-called “residential” doses from intakes of 131I were estimated up to 20 June 1986, whereas those from external irradiation were estimated for the entire period of work as a liquidator. Residential doses were not calculated for Russian and Baltic workers, as they are thought to have consumed foodstuffs that were produced in and imported from non-contaminated areas. The estimated thyroid doses from external and internal irradiation are presented in Table 8. The average external dose received by the Belarusian subjects, who worked away from the reactor site in the Belarusian part of the 30-km zone, is much smaller than that received by the Russian and the Baltic subjects, who worked at the ChNPP site. On the other hand, the average internal dose received by the Belarusian subjects was much higher than those due to the Russian and the Baltic subjects, because the Belarusian subjects consumed locally produced milk and/or leafy vegetables that were contaminated with 131I.
The Fukushima-Daiichi Nuclear Power Station (FDNPS) of the Tokyo Electric Power Company (TEPCO) consisted of six units of boiling-water reactors, with a total power generating capacity of 4,700 MW of electricity. Unit 1 started operating in 1971, Unit 2 in 1974, Unit 3 in 1976, Units 4 and 5 in 1978, and Unit 6 in 1979. On 11 March 2011, an earthquake of magnitude 9.0, the largest ever recorded in Japan, occurred along the Japan Trench. The earthquake created a series of tsunami waves that struck the east coast of Japan. The earthquake and the tsunamis knocked out the power supply to the FDNPS and, consequently, the means to control and cool the reactor. In the days that followed, reactor meltdown and hydrogen gas explosions resulted in serious damage to the facility (WHO 2013).
Similarly to the Chernobyl nuclear reactor accident, occupational exposure to radiation of the FDNPS workers included external irradiation from radiation sources within the damaged reactor and radioactive material deposited in the workplace, and internal irradiation from inhalation of radioactive material. The assessment of the radiation doses is under the responsibility of TEPCO. Doses from external irradiation were mainly monitored using alarm personal detectors that were changed every day. However, because there was a shortage of monitoring equipment at the early stage of emergency response, groups of workers were provided with a single personal dosimeter, and the resulting measurements were taken to be representative of the external doses received by all members of the group (WHO 2013). Once monitoring equipment was available for all workers, external dose assessment was based on the alarm personal detector measurements. The distribution of the monthly effective doses from external irradiation received by the TEPCO workers and their contractors in March and April 2011 is presented in Table 9 (TEPCO 2011). The external effective doses are found to be, on average, lower in April than in March by a factor of about three for the TEPCO workers and two for the contractors. The average doses in March 2011 were 19 mSv for the TEPCO workers and 6 mSv for the contractors, while the maximum dose was 200 mSv for both the TEPCO workers and the contractors. It is worth noting that other categories of workers who may have been exposed to radiation during the response to the accident (e.g., rescue workers, firemen, policemen, etc.) are not included in Table 9 (WHO 2013).
The internal dose assessment was based on in vivo measurements performed with whole-body counters (WBCs). The Japan Atomic Energy Agency and the National Institute of Radiological Sciences cooperated with TEPCO to assess the occupational doses during the first few months after the accident (Kurihara et al. 2012; Nakano et al. 2012; Takada et al. 2012). The initial screening was performed using WBCs equipped with plastic scintillators. In a second step, WBCs with sodium iodide scintillators were used to identify the radionuclides present in the body of workers with a predicted internal dose >20 mSv. Finally, WBCs with germanium semiconductor detectors were used for more precise measurements on workers with predicted internal doses >250 mSv (WHO 2013). The distribution of estimated internal effective doses is shown in Table 10 (WHO 2013). Among the population of 23,000 workers that was considered, only 12 TEPCO workers and no contractors were found with estimated effective doses >100 mSv. TEPCO concluded that workers with the highest internal doses were those working in a central control room; for these workers, 131I was by far the major contributor to the internal dose. Consequently, the thyroid doses are very high for those workers and are estimated to be >10 Gy for two workers and in the range from 2–10 Gy for the other 10 (WHO 2013). Because of the short physical half-life of 131I (∼1 wk), practically all of the thyroid doses had been delivered by the end of April 2011.
The cumulative effective doses from external and internal irradiation to the ∼25,000 Fukushima workers (TEPCO and contractors) that were involved in mitigation activities during the time period from March 2011 through December 2012 are presented in Table 11 (TEPCO 2011). Although many more contractors than TEPCO workers had been involved in mitigation activities up to the end of 2012, the TEPCO workers received most of the effective doses >100 mSv (161 versus 20 for the contractors). The average effective dose was 25 mSv among the TEPCO workers and 10 mSv among the contractors. A comparison of the effective dose estimates presented in Table 11 on the one hand and in Tables 9 and 10 on the other shows that the highest doses were received during the first 2 mo following the accident (March and April 2011).
Increased occupational exposures that resulted from the four major reactor accidents that happened since the beginning of the nuclear era and from the early operations of two plutonium-production facilities for military purposes are presented and discussed in this paper. Particular attention is paid to the increased occupational exposures resulting from the Chernobyl nuclear reactor accident that occurred in Ukraine in April 1986, the reactor accident of Fukushima that took place in Japan in March 2011, and the early operations in the 1940s and 1950s of the MPA, which is located in Russia.
The Chernobyl nuclear reactor accident is the most serious that ever occurred in the nuclear industry. In addition to the ∼800 emergency workers involved during the first few days after the accident in firefighting and closing down unaffected units of the power plant, >500,000 cleanup workers took part in 1986–1990 in the mitigation of the consequences of the accident, including decontamination and construction of the sarcophagus. Among the emergency radiation workers, special attention is paid to the 134 persons who had been diagnosed with ARS: They received bone marrow doses due to external gamma radiation ranging from 0.8 to 16 Gy. The average effective dose received by the 530,000 cleanup workers, also called liquidators or recovery operation workers, was mainly due to external irradiation and is estimated to have been ∼0.12 Sv. The recorded worker doses varied from <0.01 to >1 Sv, although ∼85% of the recorded doses were in the range from 0.02–0.5 Sv.
The accident at FDNPS was the consequence of an earthquake of magnitude 9.0, which triggered a major tsunami that submerged the emergency diesel generators, resulting in serious damage to the reactor. From the time of the accident until the end of 2012, ∼25,000 workers were involved in activities on the reactor site; six of those workers received effective doses (external + internal) >0.25 Sv, 167 workers received effective doses >0.1 Sv, and about two-thirds of the workforce received effective dose equal to or below 0.01 Sv.
The MPA was the first industrial complex in the former Soviet Union built for the production of plutonium. The complex included reactors, chemical processing plants, and plutonium production facilities. In the early years, there was poor understanding of the consequences of relatively high occupational radiation exposures. The highest external gamma doses were recorded in 1948–1952; i.e., during the start-up and adaptation phase of the reactor and radiochemical plants. Average values of annual doses amounted to 1 Gy, and maximum individual annual doses were up to 8 Gy. High internal doses were due to the exposure to plutonium.
Both for the Chernobyl nuclear reactor accident and the routine operations at Mayak, the considerable efforts made to reconstruct individual doses from external irradiation to a large number of workers revealed that the recorded doses had been overestimated by a factor of about two.
The authors thank Ethel Gilbert and Vladimir Drozdovitch (National Cancer Institute), Kazuo Sakai and Shin Saigusa (National Institute of Radiological Sciences), Derek Hagemeyer (Oak Ridge Associated Universities), and Ellen Anderson (Nuclear Energy Institute) for providing information on occupational doses that is not readily available in the open literature.
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