Radiological emergency
A radiological emergency includes events that result in large-scale exposure to ionizing radiation that can adversely affect health and the environment. In the context of this article, these are supraregional or regional emergencies within the meaning of Section 5 (26) of the German Radiation Protection Act (StrlSchG) . Local emergencies within the meaning of the German StrlSchG, which are assessed and fought on site by the general hazard prevention and disaster control of the German federal states, are not the subject of this article.
Radiological emergency management
In radiological emergencies, Part 3 of the StrlSchG is used in Germany to protect the population from radiation. If necessary, early emergency response measures are taken to avert threats to individuals in the population. Criteria are provided by the Emergency Dose Value Ordinance ( NDWV ). Supplementary measures stipulate statutory ordinances, which are issued depending on the situation, if necessary. European legal acts have priority. If the situation so requires, temporary emergency ordinances can also be issued by certain federal ministries.
Measures are taken within the framework of an emergency management system that has been set up in Germany based on experience as a result of the Chernobyl reactor accident and has been continuously optimized since then. It provides for a federal radiological situation center. This coordinates the responsible federal and state authorities and specialist agencies in their protective measures. As a precaution, such measures are planned in advance in a coordinated system of emergency plans of the federal government and the federal states. They are based on uniform assessment principles in the form of dose and contamination values.
The radiological situation center maintains a radiological situation report to which u. a. the integrated measurement and information system contributes to the monitoring of environmental radioactivity ( IMIS ). This picture of the situation is decisive for all acute and strategic decisions to be made in the course of the emergency.
A local, e.g. The authorities responsible for public safety and disaster control provide security and protection against hazards, for example, in the vicinity of a nuclear facility and are the subject of disaster control planning in the federal states. Disaster prevention and emergency management are closely interlinked and complement one another.
Radioactivity in radiological emergencies
Radioactivity is the spontaneous transformation of unstable atomic nuclei ( nuclides ) with the emission of ionizing radiation . Such unstable nuclides are called radionuclides .
Stable nuclides and radionuclides are characterized by specifying the chemical element and the mass number of their atomic nuclei. For example, the element cesium (chemical abbreviation Cs) includes nuclides with mass numbers from 113 to 148. Only one of these many nuclides is stable, namely the one with mass number 133 (Cs-133). It consists of 55 protons, like all other nuclides of the element cesium, and 78 neutrons. In total, this is - according to the mass number - 133 core components. The other nuclides of the element cesium are radionuclides, including the radionuclide Cs-137, which is particularly important for radiological emergencies. They only exist when they are artificially created, are then subject to radioactive decay and emit ionizing radiation.
Radionuclides can be of natural origin, arise from nuclear fission in nuclear reactors or from nuclear weapon explosions . Radionuclides of natural origin are either created together with the stable nuclides ( primordial radionuclides ) or they are continuously generated in natural processes. Potassium-40 (K-40) is an example of a primordial radionuclide. The radioactive tritium (H-3), on the other hand, is a cosmogenic radionuclide . It is continuously generated by the interaction of cosmic rays with the earth's atmosphere.
The nuclide that arises during radioactive decay, the so-called daughter nuclide , a decay product, can again be a radionuclide or a stable nuclide. The conversion processes are characteristic of the respective radionuclides. They run off spontaneously, follow statistical laws and, under certain circumstances via decay series , finally always end with stable nuclides.
Characteristic of the radioactive decay of a radionuclide is its half-life and the radiation that occurs. There are mainly two types of particle radiation , alpha rays and beta rays . In addition, there is gamma radiation , which is not particle radiation, but rather electromagnetic radiation of relatively high energy, which almost always follows the particle radiation. There are other types of radiation, which, however, have no further significance in relation to the subject matter dealt with below, such as " positron radiation " ("Beta +") and " K capture ". B. for K-40 together make up 11% of the decays.
A special form of radioactive conversion is spontaneous nuclear fission . This is observed with a few, particularly heavy nuclides, e.g. B. in the primordial radionuclide uranium -235 (U-235). The resulting fission products are almost always radioactive. Cs-137 is e.g. B. such a fission product. Fission products are created to a considerable extent artificially in nuclear reactors in which humans control the nuclear fission process. Other important examples of radionuclides which, like Cs-137, can be important in radiological emergencies, are the cleavage products Sr -90, Mo -99, Ru -106, Te -132, I -131, I-132, I-133 and Ba -140 and the heavy nuclides U-238 and Pu -239. Radionuclides that are artificially produced in other ways and are widely used in technology and medicine can also be important in radiological emergencies, such as B. Co -60.
Radiological emergencies are determined by the radionuclides present, their activity and their half-life . Activity information, measured in the unit Becquerel (unit symbol: Bq), must always express where the activity was measured, which radionuclides it relates to and how the activity is distributed. This can be inaccessible or evenly distributed over a large area on the floor, restricted to a certain area, concentrated in the smallest of spaces (“hot spots”). Examples of activity information are e.g. B. Bq per m², Bq per m³, Bq per kg, Bq per liter, Bq per unit of use and Bq per hot spot.
At first glance, radiological emergencies are characterized by gamma radiation and the external radiation it causes. Radionuclides can also get into the body (incorporation), e.g. B. by breathing in ( inhalation ) or by swallowing when eating ( ingestion ). With regard to the radiation effect, particle radiation then dominates. It can develop its particularly damaging effect in biological tissue, while it is otherwise hardly ever reached from the outside or is at the latest shielded by the skin.
Table 1 shows data from radionuclides, the selection of which was specifically made with the help of the literature with regard to this article. In particular, the general administrative regulation for the integrated measurement and information system for monitoring radioactivity in the environment (IMIS) , the publication UNSCEAR 2008 and the list of isotopes were used . The table therefore mainly includes radionuclides that are included in the routine measurement program of IMIS or that were released in the Chernobyl reactor accident. Mostly they are fission products. In addition to gamma radiation, beta radiation is the predominant type of radiation.
| Z | element | Order code drawing |
half-life time |
Type of decay | Decay energy in keV |
Dominant particle energy in keV |
Gamma radiation |
Hints |
|---|---|---|---|---|---|---|---|---|
| 1 | hydrogen | H-3 | 12.3 a | beta | 19th | 19th | - | - |
| 19th | potassium | K-40 | 1.3 billion a | Beta K, Beta + |
1311 1505 |
1311 44 |
- gamma |
89% of the decays 11% of the decays |
| 27 | cobalt | Co-60 | 5.3 a | beta | 2824 | 318 | gamma | - |
| 36 | krypton | Kr-85 | 10.7 a | beta | 687 | 687 | gamma | - |
| 38 | strontium | Sr-89 Sr-90 |
50.5 d 28.8 a |
Beta beta |
1495 546 |
1495 546 |
- - |
- disintegrates into Y-90 |
| 39 | yttrium | Y-90 | 2.7 d | beta | 2280 | 2280 | - | Decay product of Sr-90 |
| 40 | Zirconium | Zr-95 | 64 d | beta | 1125 | 368 | gamma | breaks down into Nb-95 |
| 41 | niobium | Nb-95 | 35 d | beta | 926 | 160 | gamma | Decay product of Zr-95 |
| 42 | molybdenum | Mon-99 | 2.7 d | beta | 1215 | 1215 | - | breaks down into Tc-99m |
| 43 | Technetium | Tc-99m | 6 h | gamma | 140 | - | gamma | Decay product of Mo-99 |
| 44 | Ruthenium | Ru-103 Ru-106 |
39.3 d 1 a |
Beta beta |
726 39 |
726 39 |
- - |
disintegrates into Rh-103 disintegrates into Rh-106 |
| 45 | Rhodium | Rh-103m Rh-106 |
56 min 30 s |
Gamma beta |
40 3541 |
- 3541 |
Gamma gamma |
Decay product of Ru-103 Decay product of Ru-106 |
| 52 | Tellurium | Te-129m Te-129 Te-132 |
33.6 d 1.2 h 3.2 d |
Gamma Beta Beta |
105 1498 493 |
- 1470 215 |
Gamma Gamma Gamma |
- The ground state of Te-129m breaks down into I-132 |
| 53 | iodine | I-131 I-132 I-133 |
8 d 2.3 h 20.8 h |
Beta beta beta |
971 3580 1771 |
606 2140 1241 |
Gamma Gamma Gamma |
- decay product of Te-132 - |
| 54 | xenon | Xe-133 | 5.2 d | beta | 427 | 346 | gamma | - |
| 55 | Cesium | Cs-134 Cs-136 Cs-137 |
2 a 13.6 d 30.2 a |
Beta beta beta |
2059 2548 514 |
658 2208 514 |
Gamma Gamma - |
- - breaks down into Ba-137m |
| 56 | barium | Ba-137m Ba-140 |
2.6 min 12.8 d |
Gamma beta |
662 1050 |
- 1006 |
Gamma gamma |
Decay product of Cs-137 decays into La-140 |
| 57 | Lanthanum | La-140 | 1.7 d | beta | 3762 | 1350 | gamma | Decay product of Ba-140 |
| 58 | cerium | Ce-141 Ce-144 |
32.5 d 284.7 d |
Beta beta |
581 319 |
435 319 |
Gamma gamma |
- breaks down into Pr-144 |
| 59 | Praseodymium | Pr-144 | 17.3 min | beta | 2998 | 2998 | - | Decay product of Ce-144 |
| 86 | radon | Rn-222 | 3.8 d | alpha | 5590 | 5490 | - | Decay product of Ra-226 |
| 88 | radium | Ra-226 | 1602 a | alpha | 4871 | 4784 | gamma | breaks up into marginal 222 |
| 92 | uranium | U-234 U-235 U-238 |
250,000 a 700 mill. A 4.5 bill. A |
Alpha Alpha Alpha |
4858 4679 4270 |
4775 4397 4198 |
Gamma Gamma Gamma |
- Start of the uranium-actinium decay series Start of the uranium-radium decay series |
| 93 | neptunium | Np-239 | 2.4 d | beta | 722 | 436 | gamma | - |
| 94 | plutonium | Pu-238 Pu-239 Pu-241 |
87.7 d 24 110 a 14.4 a |
Alpha Alpha Beta |
5593 5244 21 |
5499 5157 21 |
Gamma Gamma - |
Decay product of Cm-242 - decays to Am-241 |
| 95 | Americium | Am-241 | 432.2 a | alpha | 5638 | 5486 | gamma | Decay product of Pu-241 |
| 96 | Curium | Cm-242 | 162.8 d | alpha | 6216 | 6113 | gamma | breaks down into Pu-238 |
Radiation effect
Both particle radiation and gamma radiation ionize atoms in the tissue that they hit. Chemical compounds, including genetically important molecules such as B. the DNA , up to higher dimensional tissue structures damaged or destroyed. The result is stochastic damage , such as an increased risk of cancer and - in the case of very intensive radiation exposure - deterministic damage (tissue reactions up to radiation sickness ).
The effect of ionizing radiation is determined by the absorbed dose . Its unit gray (unit symbol: Gy) relates to the unit of mass of the exposed substance and describes the energy that the radiation has transferred to it. The stochastic effects in humans are then quantified by the dose equivalent . With regard to such risks, it also takes into account the different effectiveness of the types of radiation through radiation weighting factors .
Body doses with the dose unit Sievert (unit symbol: Sv) are decisive in radiation protection . Specifically, these are the organ equivalent dose and the effective dose or, in the case of the incorporation of radionuclides, the subsequent organ equivalent dose and the effective subsequent dose .
In the context of this article, the effective dose (or effective subsequent dose ) is mainly addressed. It is related to the whole body and represents a balance of the dose. The different radiation sensitivity of the affected organs is taken into account by tissue weighting factors (see article Effective dose ). By specifying the effective dose, different types of exposure can be compared directly with one another with regard to the stochastic risk.
Body doses cannot be measured directly. Their determination is based on measured variables such as B. the personal dose, the local dose, the local dose rate, the concentration of radioactive substances in the air, the contamination of the workplace, the activity of excretions and the properties of the radiation source or the radiation field (see article equivalent dose ).
Scenarios
The only radiological emergency to date that has affected Germany was the effects of the Chernobyl reactor accident in 1986.
Radiological emergencies in this sense should always be preceded by a warning. An unplanned event has occurred which leads to a release of radioactivity at the site of the event. U. has led or may lead to a disaster. If this can subsequently lead to an entry of radionuclides at one's own location, there is a potential hazard. If significant adverse effects from ionizing radiation on people, the environment or property must be expected, this is an “emergency” as defined by the StrlSchG. The responsible authorities inform the affected population about the situation, give recommendations on behavior and proceed as part of the radiological emergency management according to their emergency plans.
If radionuclides are introduced, this is precisely monitored in Germany with the help of the integrated measurement and information system for monitoring environmental radioactivity IMIS. It can be assumed that the relevant data will be published quantitatively up-to-date on the Internet.
The entry takes place mainly with the air. Possible radiation exposures are direct irradiation from the air (submersion), radiation from the ground through radionuclides that have been deposited, and the incorporation of radionuclides through inhalation and ingestion.
According to the International Commission on Radiation Protection (ICRP) , it is primarily external radiation, mainly ground radiation, that is important, while submersion is only of secondary importance. In the second place, the incorporation is decisive, in the long term the ingestion, in the short term and overall of less importance the inhalation.
If the radiological situation has stabilized after an emergency, there is an "existing exposure situation" in the sense of Part 4 Part 4 of the StrlSchG. Protective measures taken can be adapted on this basis, e.g. U. be repealed and supplemented by remedial measures.
This article deals with supraregional and regional, i. H. large-scale radiological emergencies. Disaster scenarios and local emergencies, e.g. B. Scenarios like the one in Goiânia (Brazil) in 1987 , where massive, but locally limited radioactive contamination was caused by a highly radioactive radiation source that was scrapped, are outside of such scenarios and are not considered here.
Dose benchmarks
Reviews in radiation protection based on planned applications of ionizing radiation at a justification bid to dose limits and a requirement to minimize . In the case of radiological emergencies that are always unplanned and in which a distinction is made between emergency exposure situations and “existing” exposure situations instead, according to StrlSchG, these principles can be applied accordingly. Reference values and emergency dose values take the place of dose limit values .
Reference values correspond to dose values that are related to a year and are determined on the assumption that they come about under realistic conditions, in particular the use of protective measures. They are set in advance as the upper limit for the effective dose, but can be reduced and additionally specified in the event of an emergency (e.g. with regard to organ equivalent doses). Once established, reference values must not be exceeded during an emergency, but should be undercut as far as possible by adapting the protective measures taken.
Emergency dose values are planning values based on the fictitious scenario of an imaginary, unprotected and permanently exposed person in the open air. They serve as radiological criteria for the appropriateness of certain protective measures. If the simplifying estimates made according to this model at the beginning of an emergency reveal that emergency dose values are being exceeded, the associated protective measures must generally be initiated.
The following compilation of basic dose values, which was made against the background of planned, existing and emergency exposures as well as natural radiation exposure, is intended to provide orientation (see also Fig. 1).
Benchmarks 1 and 2: Natural radiation exposure is omnipresent, both spatially and temporally. In Germany, health risks have so far only been proven for the contribution made by inhaling radon (increase in the risk of lung cancer). The effective dose resulting from natural radiation exposure as the sum of external and internal radiation exposure, i.e. H. is ingested by an individual of the population per year through external irradiation, through the incorporation of natural radionuclides with the breath (inhalation) and through food (ingestion), averages a little over 2 mSv / a.
Not only the level of natural radiation exposure is decisive, but also its range of fluctuation. There are significant differences in the radiation dose absorbed depending on where a person lives and where they live. The range of variation of the mean dose mentioned, i.e. H. the difference between the lowest and highest observed values is more than twice the stated mean. In relation to the lifetime , this fluctuation range is approx. 300 mSv.
The natural external radiation exposure is effectively around 0.7 mSv / a and has mainly two causes, terrestrial radiation and cosmic radiation. The natural terrestrial radiation comes from natural radioactive substances in the soil, in the rock and in building materials. In Germany, a person receives an average effective dose of 0.4 mSv per year from terrestrial radiation. The cosmic radiation leads to an average effective dose of 0.3 mSv per year.
The natural internal radiation exposure is effectively around 1.4 mSv / a. It arises mainly due to the inhalation of radon and its decay products, especially when staying in buildings. In Germany, this gives a person a mean effective dose of 1.1 mSv per year. The incorporation of natural radionuclides through ingestion leads to an average annual mean effective dose of 0.3 mSv per year.
Benchmark 3: The radiation exposure of the population that is caused by a planned occupational activity is effectively limited to 1 mSv / a. Radiation exposure due to the discharge of radioactive substances must also be included. If there is a possibility that this limit value could be exceeded, which is only permitted in operational areas, the person responsible must monitor these (" monitoring areas ") and access restrictions apply. Radiation exposures to which members of the public are exposed by being medically examined or treated are not included in this limitation. In Germany, people receive an average effective dose of 1.7 mSv per year through X-ray diagnostics . For benchmark 3, see § 80 StrlSchG.
Benchmark 4: By discharging radioactive substances, individuals in Germany may not receive an effective dose of more than 0.3 mSv / a. Compliance with the limit value must be proven for the most unfavorable exposure point for an intended reference person with unfavorable lifestyle habits. For benchmark 4, see Section 99 of the German Radiation Protection Ordinance (StrlSchV).
Benchmark 5: If the possible exposure of persons is limited to an effective 6 mSv / a in the occupational area, the conditions for a monitoring area are initially present, unless it is reliably below 1 mSv / a. If the required monitoring measures for these areas are transferred to radiological emergencies, measures taken by the state in the form of extensive measurements of environmental radioactivity (e.g. IMIS) can be viewed as analogous monitoring measures and as appropriate. An exposed person who exceeds this value is considered an "occupationally exposed person" and is subject to preventive occupational medical examinations. Areas in which people can exceed this dose value are controlled areas and require a series of radiation protection measures (labeling, access restrictions, instructions, personal dosimetry). When measuring the dose values, which define the monitoring and control areas, a residence time of 40 hours per week is taken as a basis. For benchmark 5, see Sections 52, 53, 55, 56, 63, 64 StrlSchV.
Benchmark 6: The radiation exposure of occupationally exposed persons is limited to an effective 20 mSv / a. The limit value of 20 mSv may be exceeded in a calendar year up to a value of a maximum of 50 mSv, provided that it is adhered to on average together with the dose values for the following four years. Organ equivalent doses are also limited, e.g. B. for hands to 500 mSv per calendar year. For benchmark 6 see Section 78 StrlSchG.
An effective dose of 20 mSv / a is also the upper reference value for unplanned "existing" exposures. He could e.g. This can be used, for example, if after an evacuation or resettlement it is checked whether it is possible to return to an abandoned contaminated area (transition from an emergency exposure situation to an existing exposure situation). Regarding benchmark 6, see also § 118 Paragraph 1 No. 1 and Paragraph 4 StrlSchG.
Benchmark 7: In order to limit the exposure of individuals in the population to major accidents, the relevant designs in nuclear power plants must ensure that the effective dose remains limited to 50 mSv. The subsequent organ equivalent dose of the thyroid is also limited to 150 mSv. For benchmark 7, see § 104 StrlSchV.
Benchmark 8: The limit for a particularly approved radiation exposure is effectively 100 mSv. This is intended to enable "necessary specific" work processes after approval. Such work processes may only be carried out by volunteers who, as occupationally exposed persons, are subject to preventive occupational medical examinations. For benchmark 8 see § 74 StrlSchV.
Benchmark 9: The emergency dose value for staying in buildings is effectively 10 mSv in one week. For benchmark 9 see § 2 NDWV.
Benchmark 10: The emergency dose value for an evacuation is effectively 100 mSv in one week. For benchmark 10, see § 4 NDWV.
Benchmark 11: The reference value for the effective dose absorbed in the first year after the occurrence of an emergency is set at 100 mSv. In order to comply with it, the affected population must be relocated if necessary. If it were exceeded, deterministic radiation damage could also occur. For benchmark 11 see Section 93 (1) StrlSchG.
Key values 12, 13: The emergency dose value for taking iodine tablets relates to the inhalation of radioiodine and is 50 mSv in 7 days for children and adolescents under the age of 18 and for pregnant women for the thyroid dose (subsequent organ equivalent dose). For people aged 18 to 45, the corresponding emergency dose value is 250 mSv in 7 days. Older people are not advised to take iodine tablets. In general, because of the possible side effects, iodine tablets should only be taken if specifically requested by the responsible authorities. For key values 12 and 13, see § 3 NDWV.
Benchmark 14: The EU regulation on the limitation of the activity concentration of radionuclides in food is based on an upper reference value for the effective dose of 1 mSv per year, whereby it is assumed that 10% of the food consumed annually is contaminated. For benchmark 14 cf. Section Ingestion and see Section 94 (2) No. 3 StrlSchG (authorization to issue ordinances).
Dose estimation
The following estimates are limited to the three main radiation exposures in the form of external radiation , in particular from radionuclides deposited on the ground or from submersion, inhalation of radionuclides and ingestion of radionuclides. In principle, the effects of all three forms of radiation exposure mentioned must be added together. Depending on the development of the situation, however, one or the other exposure path may temporarily dominate and can be considered on its own, while the other exposure paths no longer have to be considered or may only have to be considered later. For the effective dose, the assessment is based on the sum of the individual types of exposure. This also applies to contributions from individual radionuclides in a radionuclide mixture. Each radionuclide can be viewed on its own. The timing is important. Due to the different half-lives, some radionuclides may only play a role in certain time windows and can otherwise be disregarded.
In the following, estimates of a selection of radionuclides are presented. The selection was made primarily with a view to which radionuclides could be relevant as fission products. The radionuclides that are routinely monitored within the framework of IMIS and the radionuclides that were spread as a result of the reactor accidents at Chernobyl and Fukushima were used as a guide. Co-60 was chosen as the radionuclide that could be typical of “dirty bombs”. K-40 was chosen as a typical representative and reference nuclide for natural radiation exposure. Important radionuclides are the fission products I-131 and Cs-137. I-131 is also an example of a relatively short-lived radionuclide, Cs-137 for a relatively long-lived one. I-131 is also an example of a radionuclide that causes relevant organ doses (here the thyroid dose (subsequent organ equivalent dose)).
It is impossible to reliably detect in advance all the radionuclides that may arise in a radiological emergency. For the radionuclides not listed here, required data, e.g. B. the relevant dose coefficients and the dose rate coefficients can be looked up at any time. The dose and dose rate coefficients are given in the published tables of the ICRP, taking into account the age of the exposed person, both for organs and for the effective dose. In the following estimates, mostly only the effective value is considered (with the exception of iodine, where the thyroid dose (subsequent organ equivalent dose) is also given in some representations with regard to inhalation and ingestion).
External exposure
Without first having to go into more detail about radionuclides, the dose estimate for external radiation focuses on the dose rate. In general, a body dose results as the product of the dose rate with the exposure time.
- Body dose (microsievert) = dose rate (microsievert per hour) • exposure time (hours)
In the case of longer periods of time and a changing dose rate, this is to be added up over the relevant "integration period". An integration period of one year is usually used to evaluate a dose, and an integration period of seven days for emergency dose values.
At the beginning of a radiological emergency, measurement results for the dose measurement variable local dose rate ( dose equivalent rate ) are usually available first . Your measurement is simple and quickest. Strictly speaking, only the local dose , the product of the local dose rate over time, can initially be inferred from the local dose rate. Nevertheless, this also results in a conservative value for the body dose .
In the case of a large-scale, temporally and spatially largely homogeneous local dose rate, the following applies with regard to the usually decisive effective dose:
- Effective dose (microsievert) ≤ local dose rate (microsievert per hour) • exposure time (hours)
If one equates the effective dose to the local dose in such a simplistic way, the following rule of thumb is available:
- 1 µSv / h local dose rate → slightly less than 10 mSv effective dose per year.
Whether a measured value of the local dose rate is meaningful depends in particular on
- where the measurements were taken and whether it is a representative, and in particular accessible, location,
- The spatial extent over which the measured value extends and whether there are local fluctuations,
- when it was measured and whether there is a trend over time,
- whether it is a maximum or average value,
- with which measuring method the measured value was recorded.
Only the presence of such additional information allows a useful interpretation of a measured value and determines the quality of information. Notwithstanding this, missing information can sometimes be inferred from others. As far as external irradiation is concerned due to radionuclides deposited on the ground, see the next section in this regard. On the other hand, serious measurement results are always published in connection with the additional information required.
In the attached figure (Fig. 2) the dose rate acts as a parameter and forms the initial value (DL 0 ) of an exposure. The associated function graphs are projections of the dose for the case of a constant dose rate over the period under consideration. This is the case, for example, if the source of the dose rate does not change significantly during this time. If it is a long-lived radionuclide such as Cs-137, this can be the case for the entire year if the general conditions otherwise remain the same. With shorter-lived radionuclides like I-131, however, the radioactive decay must be taken into account when interpreting the graph. The maximum achievable dose in this case can be read off at around 1.5 times the half-life; for I-131 this would be twelve days.
The extrapolated doses are assessed by comparing them with the basic dose values. Taking the mentioned restrictions into account, Fig. 2 quickly shows which key values can be achieved by extrapolating a local dose rate. Conversely, on the basis of a benchmark value, it is possible to estimate whether this can be achieved at the present dose rate and what the temporal relevance is for it.
Because the local dose rate can easily be measured, it is the preferred metric to assess radiological emergencies based on their value and to decide immediately whether certain protective measures are necessary. As a decision-making aid , the Radiation Protection Commission (SSK) derived and recommended guide values for the local dose rate for the relevant dose values and relevant scenarios. Radiologically justified exposure conditions corresponding to the respective scenario as well as pragmatic considerations are included in these deductions. In addition to the measured variable local dose rate, the SSK also considered other suitable measured variables (in particular surface contamination) and recommended derived guide values for this. See the Assessment section for details .
A map of the local dose rate in Germany from the IMIS can be viewed at any time on the Internet via the website of the Federal Office for Radiation Protection. For each of the 1800 measuring points operated in IMIS, the current measured value as well as the chronological sequence of the hourly measured values over the past seven days and the daily mean values over the past year can be read.
Ground radiation
Typical of external radiation exposure is gamma radiation, which emanates from a radionuclide or a mixture of radionuclides deposited over a large area in the area (ground radiation).
The local dose rate resulting from the deposited activity is decisive for an assessment of the radiation exposure. The measurand from which this can be derived is the activity of the radionuclide deposited per square meter of soil, i.e. H. surface contamination or, more specifically, “soil contamination”.
In the case of an imaginary, infinitely extensive area on which the radionuclide is evenly distributed, the soil contamination from all areas of the area contributes to the local dose rate at any point over the area. Because of the greater distance and because of the weakening of the radiation in the atmosphere, the contributions from areas further away are of course correspondingly smaller.
The knowledge of the radionuclide in connection with its activity distribution can be used to derive the other variables that are decisive for estimating the risk.
In particular, the type of radiation and half-life, which are characteristic of each radionuclide, are decisive. The half-life determines how quickly the radiation decays unless activity is continuously replenished, e.g. B. caused by the weather with the wind or rain washed out of the atmosphere. The composition of the activity can also change, e.g. B. by other radionuclides follow and be deposited. The better the information about the root causes of the contamination, the better the predictions.
The easiest way to calculate the dose rate under the conditions of ground radiation is with the help of dose rate coefficients, which are given by the ICRP (see Fig. 4 and Tab. 2). These nuclide-related coefficients do not describe the local dose rate (equivalent dose), but rather the increase in body doses (organ equivalent doses and effective doses). This means that these body doses can be calculated directly for an imaginary person standing on a contaminated surface:
- Dose rate (Sv / s) = activity per area (Bq / m²) • Dose rate coefficient ((Sv / s) / (Bq / m²))
The extensive tables of the ICRP consider all known radionuclides and differentiate between age groups of people. The dose rate coefficients for children are higher than for adults because their body organs are closer to the radiating surface due to their smaller body size. Possibly. Existing short-lived daughter nuclides are taken into account in the dose rate coefficients.
For example, the dose rate coefficient for Cs-137 in adults (age over 17 years) has the value 1.9 (nSv / h) / (kBq / m²) for the effective dose, as can be seen from the attached figures (Figs. 4, 5 and Tab. 2) can be read out. With a soil contamination of 40,000 Bq / m² (40 kBq / m²), multiplication with this dose rate coefficient results in a dose rate of around 80 nSv / h for the effective dose. Soil contamination of this magnitude occurred in areas of southern Germany due to the Chernobyl reactor accident. With a dose rate meter that measures the local dose rate, a value of this order of magnitude would be measured in a real, comparable scenario (equivalent dose rate). The exact value actually measured would still depend on the actual natural terrain conditions present and on fluctuations in the deposited activity.
Without a calculation, even if this were simple, the dose rate value for various radionuclides can be read off directly from Fig. 5 if the activity per m² deposited is known. The maximum value results for Te-132 (with daughter I-132), which enables a "conservative" estimate (cf. the "rule of thumb" given in Fig. 5).
Fig. 5 also shows two dose rate values that correspond to the key values in the “Basic dose values” section and with which the dose rate can be evaluated. The benchmark value 1 of the mean natural external radiation exposure (effectively 0.7 mSv per year) corresponds to a dose rate of 80 nSv / h. The figure shows that deposits of at least 10 kBq / m² would be necessary to cause such dose rates. That would be the case, for example, with the radionuclide Co-60, which has one of the highest dose rate coefficients. With Cs-137, for example, the figure confirms that only a deposit of approx. 40 kBq / m² would correspond to the natural external radiation exposure. In the case of I-131, the natural dose rate value would only be reached with even higher deposits, whereby, due to the short half-life of only 8 days (see Table 1), an initial activity with a corresponding dose rate would quickly have subsided.
The benchmark value 5 (effectively 6 mSv per year), where personal protective and monitoring measures have to be taken in the event of occupational exposure, would be achieved with a continuous dose rate of almost 700 nSv / h (0.7 µSv / h). This requires deposits of at least 100 kBq / m², for Cs-137 for example more than 300 kBq / m².
| Z | element | Order code drawing |
half-life time |
Type of decay | DL coefficient effective dose in (Sv / s) / (Bq / m²) |
DL coefficient effective dose in (nSv / h) / (kBq / m²) |
Hints |
|---|---|---|---|---|---|---|---|
| 19th | potassium | K-40 | 1.3 billion a | Beta, K, Beta + | 1.40E-16 | 5.04E-01 | - |
| 27 | cobalt | Co-60 | 5.3 a | beta | 2.20E-15 | 7.92E + 00 | - |
| 40 | zirconium | Zr-95 | 64 d | beta | 1.40E-15 | 5.04E + 00 | with daughter Nb-95 |
| 42 | molybdenum | Mon-99 | 2.7 d | beta | 2.40E-16 | 8.64E-01 | with daughter Tc-99m |
| 44 | Ruthenium | Ru-103 | 39.3 d | beta | 4.40E-16 | 1.58E + 00 | with daughter Rh-103m |
| 44 | Ruthenium | Ru-106 | 1 a | beta | 1.90E-16 | 6.84E-01 | with daughter Rh-106 |
| 52 | Tellurium | Te-129m | 33.6 d | gamma | 7.00E-17 | 2.52E-01 | - |
| 52 | Tellurium | Te-132 | 3.2 d | beta | 2.30E-15 | 8.28E + 00 | with daughter I-132 |
| 53 | iodine | I-131 | 8 d | beta | 3.60E-16 | 1.3E + 00 | - |
| 53 | iodine | I-133 | 20.8 h | beta | 6.10E-16 | 2.20E + 00 | - |
| 55 | Cesium | Cs-134 | 2 a | beta | 1.50E-15 | 5.40E + 00 | - |
| 55 | Cesium | Cs-136 | 13.1 d | beta | 2.00E-15 | 7.20E + 00 | - |
| 55 | Cesium | Cs-137 | 30.2 a | beta | 5.30E-16 | 1.91E + 00 | with daughter Ba-137m |
| 56 | barium | Ba-140 | 12.8 d | beta | 2.20E-15 | 7.92E + 00 | with daughter La-140 |
| 58 | cerium | Ce-141 | 32.5 d | beta | 6.90E-17 | 2.48E-01 | - |
| 58 | cerium | Ce-144 | 284.7 d | beta | 5.60E-17 | 2.02E-01 | - |
| 88 | radium | Ra-226 | 1602 a | alpha | 1.60E-15 | 5.76E + 00 | - |
| 92 | uranium | U-234 | 250,000 a | alpha | 6.50E-19 | 2.34E-03 | - |
| 92 | uranium | U-235 | 700 mill. A | alpha | 1.50E-16 | 5.40E-01 | - |
| 92 | uranium | U-238 | 4.4 billion a | alpha | 2.30E-17 | 8.28E-02 | - |
| 93 | neptunium | Np-239 | 2.4 d | beta | 1.50E-16 | 5.40E-01 | - |
| 94 | plutonium | Pu-238 | 87.7 a | alpha | 6.90E-19 | 2.48E-03 | - |
| 94 | plutonium | Pu-239 | 24110 a | alpha | 3.20E-19 | 1.15E-03 | - |
| 94 | plutonium | Pu-241 | 14.4 a | beta | 1.20E-18 | 4.32E-03 | breaks up into Am-241 |
| 95 | Americium | Am-241 | 432.2 a | alpha | 2.30E-17 | 8.28E-02 | - |
| 96 | Curium | Cm-242 | 162.8 d | alpha | 9.70E-19 | 3.49E-03 | breaks down into Pu-238 |
Internal radiation
inhalation
A radionuclide that is in the air you breathe can be breathed in (inhaled). Particles below 10 micrometers are alveolar and are thus fed into the body. The supply is at the beginning of a chain of effects of radiation exposure through inhalation. The measurable variables that determine the intake of activity by inhalation are the activity concentration in the air and the breathing rate, which is known for an adult with 8100 m³ per year (approx. 22 m³ per day or 0.9 m³ per hour).
It can be assumed that the time during which a radionuclide is present in the breath and can be inhaled is limited. In the event of a radiological emergency, the concentration of activity in the breathing air (see Fig. 6) will quickly decrease or disappear as the radioactive substances precipitate on the ground (e.g. due to gravity or washed out of the atmosphere by rain) or air currents carry them on. A radionuclide that has been knocked down can be whirled up again ("resuspension") or radionuclides can be carried in the air again, even different than before. Nevertheless, the radiation exposure due to the inhalation of radionuclides is limited in time and is therefore of less importance compared to the long-term effective exposure paths of external radiation and ingestion. The supply of a radionuclide does not mean that it is also taken up (incorporated) by the body. It may be exhaled again with the next breath, e.g. B. Radionuclides, which are chemically noble gases such as argon, krypton or xenon. The material form in which the radionuclide is present determines the further course of incorporation after it has been supplied by inhalation.
The relevant material, chemical and physical forms that determine the incorporation of radionuclides supplied are z. B. elementary forms (e.g. iodine isotopes as iodine molecule I 2 or, in the case of metals, the metallic form), inorganic chemical compounds (e.g. oxides, hydroxides, salts), organic chemical compounds (e.g. iodomethane) or Aerosols of different sizes.
Due to its material form, a radionuclide can be more or less respirable, differently soluble in body fluids and different chemically reactive. As a result, the material form determines the further distribution of the supplied activity in the body in terms of time and space ( pharmacokinetics ). Incorporation behavior is categorized with the help of lung absorption classes , to which radionuclides are assigned depending on their chemical properties. If no individual specific data are known in individual cases, a more specific assessment is made possible by choosing an absorption class that appears suitable. A distinction is made between absorption classes F (“fast”), M (“moderate”) and S (“slow”). If a radionuclide occurs as an aerosol , it is also assigned to an activity-related averaged aerodynamic diameter (AMAD - “Activity Median Aerodynamic Diameter”). With regard to the radiation exposure of the population, an AMAD of 1 µm is usually assumed.
A radionuclide that is in a non-respiratory form can instead also get from the respiratory tract into the mouth and throat and finally into the digestive tract, which is classified as ingestion. During this time, it contributes to the dose absorption of the surrounding tissue. The further behavior of an incorporated radionuclide in the body, its transport to different body regions, its incorporation and whereabouts in certain body organs or tissues is determined by its chemical and pharmacokinetic properties and the metabolic processes . A radionuclide behaves chemically just like its stable isotopes. Quantitatively, however, it is numerically in a relatively small minority with its atoms compared to stable isotopes or elements that are chemically similar in the body. Atoms of a radionuclide are not particularly noticeable as such and “swim with” until individual atomic nuclei of the radionuclide randomly transform according to the laws of radioactive decay and the decay products go their own way.
From the point of view of the metabolism, an ingested radionuclide may be relatively insignificant because of its usually negligible mass. In absolute terms, however, there can be an enormous number of radioactive atomic nuclei. The number of decays can be correspondingly large and the emitted radiation has a destructive effect on the biological tissue. On the other hand, we are not even aware of the much larger number of stable atoms, which is inconceivable for us at all. In summary, the fate of a radionuclide in the body is described by its biological half-life . This is the time in which the initially incorporated activity has decreased by half, both through radioactive decay (due to the physical half-life) and through biological excretions .
The decisive variables for an assessment of the inhalation of radioactive substances from a radiation protection point of view are the body doses, more precisely the subsequent organ equivalent doses and the effective subsequent dose. The body doses depend on the one hand on the radiation specific for the radionuclide, in particular the particle radiation which is now becoming effective, and on the other hand on its accumulation in organs. They are assigned as subsequent organ equivalent doses or as effective subsequent dose to the time of incorporation, even if the actual dose intake may be spread over a fairly long later period. The entire physical and biokinetic behavior of the radionuclide in the future must be considered for an assessment. The ICRP provides the solution for how a dose can be estimated in practice with manageable effort under such complex framework conditions. It provides a comprehensive set of tables with calculated dose coefficients for inhalation (see Fig. 7 and Tab. 3).
For each radionuclide, the dose coefficients express the body doses received per delivered (inhaled) activity. With their help, the dose is calculated from the supplied activity as follows:
- Body dose (Sv) = Inhaled activity (Bq) • Dose coefficient (Sv / Bq)
The specific secondary conditions with regard to the form in which radionuclides are present, usually the absorption classes and the aerosol sizes, are taken into account when selecting the appropriate dose coefficients. Such secondary conditions must therefore be known at the outset. If no information is available, standard conditions are to be assumed. If necessary, “worst-case” conditions must be applied so that the selected dose coefficient is on the “safe side” and the dose estimate is considered to be “conservative”. However, this can lead to a significant overestimation of the dose.
The ICRP provides dose coefficients for subsequent organ equivalent doses and for the effective subsequent dose. They are provided for different age groups. The values shown in Tab. 3 and Fig. 7 and used in Figs. 8-11 relate to body doses for selected radionuclides (see Table 1) and apply to people over 17 years of age. Daughter nuclides are taken into account, if applicable. There are considerable differences between the radionuclides. These differences are based on the different energies of the particle radiation characteristic of the radionuclides, their different physical half-lives and their different biokinetic and metabolic behavior. For radionuclides in the form of noble gases, such as For example, Kr-85 or Xe-133, which are important breakdown products, have no dose coefficients because they are not absorbed by the body when inhaled. The range of the values is much larger than that of the dose rate coefficients for gamma radiation from ground radiation.
As a rule of thumb, inhalation of 100 Bq of the radionuclides shown in Fig. 8 leads to an effective subsequent dose of a maximum of four microsieverts. For Cs-137, the use of this rule of thumb leads to an overestimation by a factor of almost 10, which makes its conservatism clear. The prerequisite for this conservatism, however, is the existence of beta emitters only, which is the case with fission products. Alpha emitters must be able to be excluded, because for them there are much higher dose values, sometimes by a factor of 1000 and more.
Based on a known activity concentration in the atmosphere, after selecting the appropriate or, under certain circumstances, the most unfavorable dose coefficient, the expected or "worst-case" dose can be extrapolated as follows:
- Body dose (Sv) = activity concentration (Bq / m³) • dose coefficient (Sv / Bq) • breathing rate (m 3 / h) • time (h)
As a rule of thumb, staying in an atmosphere contaminated with fission products with a maximum of 1 Bq / m 3 leads to an effective dose of one microsievert per day (see Fig. 9). This enables a first and quick assessment by comparison with measured values of the so-called “total beta activity” in the air without precise knowledge of the specific radionuclides present.
Before fission products through inhalation, in addition to external radiation through soil contamination, contribute significantly to emergency dose values, very high activity concentrations must be present. In order to remain in buildings (10 mSv in seven days - basic dose value 9), for example, they must be in the order of a few hundred Bq / m 3 (see e.g. the graphs for I-131 and Cs-137 in Fig . 10).
The graphs shown for the thyroid dose (subsequent organ equivalent dose) by inhalation of I-131 (Fig. 11) apply to the elementary form (I 2 ). For the aerosol form, the dose coefficients are significantly lower. The figure shows that the intake of iodine tablets would only be appropriate (basic dose value 13) if several 1000 Bq / m 3 I-131 had to be expected in the atmosphere. Regardless of any emergency measures that may have been declared, this applies to staying in buildings.
| Z | element | Short bez. |
half-life time |
Type of decay | shape | Effective dose in Sv / Bq |
largest organ dose (organ) in Sv / Bq |
Hints |
|---|---|---|---|---|---|---|---|---|
| 1 | hydrogen | H-3 | 12.3 a | beta | organic elemental CH3T HTO aerosol (F) aerosol (M) aerosol (S) |
4.10E-11 1.80E-15 1.80E-13 1.80E-11 6.20E-12 4.50E-11 2.60E-10 |
- - - - - 3.20E-10 (lungs) 2.10E-09 (lungs) |
- |
| 19th | potassium | K-40 | 1.3 billion a | Beta, K, Beta + | Aerosol (F) | 2.10E-09 | 4.60E-09 | - |
| 27 | cobalt | Co-60 | 5.3 a | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
5.20E-09 1.00E-08 3.10E-08 |
1.30E-08 (airways) 5.20E-08 (lungs) 1.80E-07 (lungs) |
- |
| 36 | krypton | Kr-85 | 10.7 a | beta | elementary | 0 | 0 | Noble gas |
| 38 | strontium | Sr-89 | 50.5 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
1.00E-09 6.10E-09 7.90E-09 |
5.4E-09 (bone surface) 4.50E-08 (lungs) 6.20E-08 (lungs) |
- |
| 38 | strontium | Sr-90 | 28.8 a | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
2.40E-08 3.60E-08 1.60E-07 |
3.70E-07 (bone surface 2.10E-07 (lungs) 1.30E-06 (lungs) |
with daughter Y-90 |
| 39 | yttrium | Y-90 | 2.7 d | beta | - | - | - | Decay product of Sr-90 |
| 40 | zirconium | Zr-95 | 64 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
2.50E-09 4.80E-09 5.90E-09 |
5.30E-08 (bone surface) 3.10E-08 (lungs) 4.20E-08 (lungs) |
with daughter Nb-95 |
| 41 | niobium | Nb-95 | 35 d | beta | - | - | - | Decay product of Zr-95 |
| 42 | molybdenum | Mon-99 | 2.7 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
2.20E-10 8.90E-10 9.90E-10 |
1.60E-09 (airways) 6.30E-09 (lungs) 6.00E-09 (lungs) |
with daughter Tc-99m |
| 43 | Technetium | Ts-99m | 6 h | gamma | - | - | - | Decay product of Mo-99 |
| 44 | Ruthenium | Ru-103 | 39.9 d | beta | Tetroxid Aerosol (F) Aerosol (M) Aerosol (S) |
1.10E-09 4.80E-10 2.40E-09 3.00E-09 |
3.10E-09 (large intestine) 2.50E-09 (airways) 1.80E-08 (lungs) 2.20E-08 (lungs) |
with daughter Rh-103m |
| 44 | Ruthenium | Ru-106 | 1 a | beta | Tetroxid Aerosol (F) Aerosol (M) Aerosol (S) |
1.80E-08 7.90E-09 2.80E-08 6.60E-08 |
4.10E-08 (large intestine) 1.30E-08 (large intestine) 2.00E-07 (lungs) 5.30E-07 (lungs) |
with daughter Rh-106 |
| 45 | Rhodium | Rh-103m | 56 min | gamma | - | - | - | Decay product of Ru-103 |
| 45 | Rhodium | Rh-106 | 30 s | beta | - | - | - | Decay product of Ru-106 |
| 52 | Tellurium | Te-129m | 33.6 d | gamma | elementary aerosol (F) aerosol (M) aerosol (S) |
3.70E-09 1.30E-09 6.60E-09 7.90E-09 |
3.40E-08 (bone surface) 1.10E-08 (bone surface) 4.80E-08 (lungs) 6.10E-08 (lungs) |
with Te-129 |
| 52 | Tellurium | Te-129 | 1.2 h | beta | - | - | - | breaks up into I-129 |
| 52 | Tellurium | Te-132 | 3.2 d | beta | elementary aerosol (F) aerosol (M) aerosol (S) |
5.10E-09 1.80E-09 2.00E-09 2.00E-09 |
7.60E-08 (thyroid) 2.50E-08 (thyroid) 1.00E-08 (lungs) 1.10E-08 (lungs) |
with daughter I-132 |
| 53 | iodine | I-132 | 2.3 h | beta | - | - | - | Decay product of Te-132 |
| 53 | iodine | I-131 | 8 d | beta | CH3I elemental aerosol (F) aerosol (M) aerosol (S) |
1.50E-08 2.00E-08 7.40E-09 2.40E-09 1.60E-09 |
3.10E-07 (thyroid) 3.90E-07 (thyroid) 1.50E-07 (thyroid) 2.20E-08 (thyroid) 1.10E-08 (lungs) |
- |
| 53 | iodine | I-133 | 20.8 h | beta | CH3I elemental aerosol (F) aerosol (M) aerosol (S) |
3.10E-09 4.00E-09 1.50E-09 5.50E-10 4.30E-10 |
6.00E-08 (thyroid gland) 7.60E-08 (thyroid gland) 2.80E-08 (thyroid gland) 3.60E-09 (thyroid gland) 2.00E-09 (lungs) |
- |
| 54 | xenon | Xe-133 | 5.2 d | beta | elementary | 0 | 0 | Noble gas |
| 55 | Cesium | Cs-134 | 2 a | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
6.60E-09 9.10E-09 2.00E-08 |
1.20E-08 (airways) 5.00E-08 (lungs) 1.40E-07 (lungs) |
- |
| 55 | Cesium | Cs-136 | 13.1 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
1.20E-09 2.50E-09 2.80E-09 |
8.90E-09 (airways) 1.60E-08 (lungs) 1.80E-08 (lungs) |
- |
| 55 | Cesium | Cs-137 | 30.2 a | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
4.60E-09 9.70E-09 3.90E-08 |
7.40E-09 (airways) 6.30E-08 (lungs) 3.00E-07 (lungs) |
with daughter Ba-137m |
| 56 | barium | Ba-137m | 2.6 min | gamma | - | - | - | Decay product of Cs-137m |
| 56 | barium | Ba-140 | 12.8 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
1.00E-09 5.10E-09 5.80E-09 |
5.40E-09 (colon) 3.50E-08 (lungs) 4.20E-08 (lungs) |
with daughter La-140 |
| 57 | Lanthanum | La-140 | 1.7 d | beta | - | - | - | Decay product of Ba-140 |
| 58 | cerium | Ce-141 | 32.5 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
9.30E-10 3.20E-09 3.80E-09 |
1.70E-08 (bone surface) 2.40E-08 (lungs) 3.00E-08 (lungs) |
- |
| 58 | cerium | Ce-144 | 284.7 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
4.00E-08 3.60E-08 5.30E-08 |
4.06E-07 (liver) 1.90E-07 (lungs) 4.20E-07 (lungs) |
with daughter Pr-144 and derivatives |
| 59 | Praseodymium | Pr-144 | 17.3 min | beta | - | - | - | Decay product of Ce-144 |
| 86 | radon | Rn-222 | 3.8 d | alpha | - | - | - | Decay product of Ra-226 |
| 88 | radium | Ra-226 | 1602 a | Apha | Aerosol (F) Aerosol (M) Aerosol (S) |
3.60E-07 3.50E-06 9.50E-06 |
1.60E-05 (bone surface) 2.80E-05 (lungs) 7.90E-05 (lungs) |
with daughter Rn-222 and derived products |
| 92 | uranium | U-234 | 250,000 a | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
5.60E-07 3.50E-06 9.40E-06 |
9.50E-06 (bone surface) 2.70E-05 (lungs) 7.80E-05 (lungs) |
with secondary products |
| 92 | uranium | U-235 | 700 mill. A | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
5.20E-07 3.10E-06 8.50E-06 |
9.00E-06 (bone surface) 2.40E-05 (lungs) 7.00E-05 (lungs) |
with secondary products |
| 92 | uranium | U-238 | 4.5 billion a | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
5.00E-07 2.90E-06 8.00E-06 |
8.70E-06 (bone surface) 2.20E-05 (lungs) 6.70E-05 (lungs) |
with secondary products |
| 93 | neptunium | Np-239 | 2.4 d | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
1.70E-10 9.30E-10 1.00E-09 |
2.70E-09 6.30E-09 (lung) 7.10E-09 (lung) |
with secondary products |
| 94 | plutonium | Pu-238 | 87.7 a | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
1.10E-04 4.60E-05 1.60E-05 |
3.60E-03 (bone surface) 1.40E-03 (bone surface) 1.60E-04 (bone surface) |
Decay product of Cm-242 |
| 94 | plutonium | Pu-239 | 24,110 a | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
1.20E-04 5.00E-05 1.60E-05 |
4.00E-03 (bone surface) 1.50E-03 (bone surface) 1.80E-04 (bone surface) |
with secondary products |
| 94 | plutonium | Pu-241 | 14.4 a | beta | Aerosol (F) Aerosol (M) Aerosol (S) |
2.30E-06 9.00E-07 1.70E-07 |
7.90E-05 (bone surface) 3.10E-05 (bone surface) 4.10E-06 (bone surface) |
with subsidiary Am-241 and follow-on products |
| 95 | Americium | Am-241 | 432.2 a | alpha | Aerosol (F) Aerosol (M) Aerosol (S) |
9.60E-05 4.20E-05 1.60E-05 |
4.40E-03 (bone surface) 1.70E-03 (bone surface) 2.10E-04 (bone surface) |
with secondary products |
| 96 | Curium | Cm-242 | 162.8 d | Apha | Aerosol (F) Aerosol (M) Aerosol (S) |
3.30E-06 5.20E-06 5.90E-06 |
9.00E-05 (bone surface) 3.50E-05 (lungs) 4.90E-05 (lungs) |
with daughter Pu-238 and derivatives |
Ingestion
Ingestion of radionuclides is their intake with food. The incorporation takes place via the digestive tract . In the short term, in a radiological emergency, radionuclides that have deposited on the surface of plants, e.g. B. on lettuce, leafy vegetables and fruit, and are directly supplied to humans through their consumption. The deposition on the plants is largely independent of the form of the radionuclides. A wide variety of chemical compounds and also relatively short-lived radionuclides can occur. Washing, storage for an appropriate length of time before consumption, restrictions on consumption, etc. are simple and effective protective measures. In the same way, radionuclides can be ingested by cows on pasture and supplied to humans through milk and dairy products.
Long-lived radionuclides can also lead to long-term radioactive contamination of food of plant and animal origin. Radionuclides can e.g. B. can be enriched via the plant metabolism from the soil in plants or from water in fish and concentrated via food chains . Accordingly, mushrooms or the meat of wild boars in certain regions of southern Germany still contain Cs-137, which comes from the Chernobyl reactor accident in 1986.
Radionuclides in the body with food are not always fully incorporated. The non-incorporated portion is excreted again. The chemical form in which the radionuclide is present and its solubility in body fluids play a decisive role. Built into the tissue of plant and animal organisms, however, radionuclides are already in a form that is much easier to incorporate when consumed than radionuclides that are merely deposited on the surface.
The conditions for ingestion of radionuclides are not quite as complex as for inhalation. Ingestion is called a fractionated absorption and this is defined by a factor ( f1-factor ), which expresses the ratio between the activity that is taken in and the activity that is incorporated via the digestive tract. In the case of complete (100%) absorption, the f1 factor has the value 1. An f1 factor of 0.3 means that only 30% of the added activity is incorporated.
The above section on inhalation already contains the essential functional aspects of the incorporation of radionuclides. Analogous to inhalation, the subsequent organ equivalent doses and the effective subsequent dose are again the determining parameters for an assessment of the incorporation by ingestion of radionuclides. The dose values can be derived from the added activity using dose coefficients ( dose coefficients for ingestion ) provided by the ICRP (see Fig. 12 and Tab. 4). Analogous to inhalation, these again relate to one radionuclide, including possibly daughter nuclides, and apply to different age groups. The f1 factor is also specific to the nuclide. The ICRP also indicates intermediate values of the subsequent doses for periods after the administration.
| Z | element | Short bez. |
half-life time |
Type of decay | f1 value |
Effective dose in Sv / Bq |
largest organ dose (organ) in Sv / Bq |
Hints |
|---|---|---|---|---|---|---|---|---|
| 1 | hydrogen | H-3 | 12.3 a | beta | 1 1 |
4.20E-11 1.80E-11 |
- | organic - |
| 19th | potassium | K-40 | 1.3 billion a | Beta, K, Beta + | 1 | 6.20E-09 | 1.40E-08 (large intestine) | - |
| 27 | cobalt | Co-60 | 5.3 a | beta | 0.1 | 3.40E-09 | 8.70E-09 (large intestine) | - |
| 36 | krypton | Kr-85 | 10.7 a | beta | - | - | - | Noble gas |
| 38 | strontium | Sr-89 | 50.5 d | beta | 0.3 | 2.60E-09 | 1.40E-08 (large intestine) | - |
| 38 | strontium | Sr-90 | 28.8 a | beta | 0.3 | 2.80E-08 | 4.10E-07 (bone surface) | with daughter Y-90 |
| 39 | yttrium | Y-90 | 2.7 d | beta | - | - | - | Decay product of Sr-90 |
| 40 | zirconium | Zr-95 | 64 d | beta | 0.01 | 9.50E-10 | 5.10E-09 (large intestine) | with daughter Nb-95 |
| 41 | niobium | Nb-95 | 35 d | beta | 0.01 | 5.80E-10 | 2.80E-09 (large intestine) | Decay product of Zr-95 |
| 42 | molybdenum | Mon-99 | 2.7 d | beta | 1 | 6.00E-10 | 3.10E-09 (kidneys) | with daughter Tc-99m |
| 43 | Technetium | Tc-99m | 6 h | gamma | 0.5 | 2.20E-11 | 6.70E-11 (large intestine) | Decay product of Mo-99 |
| 44 | Ruthenium | Ru-103 | 39.3 d | beta | 0.05 | 7.30E-10 | 4.30E-09 (large intestine) | with daughter Rh-103m |
| 44 | Ruthenium | Ru-106 | 1 a | beta | 0.05 | 7.00E-09 | 4,50E-08 (large intestine) | with daughter Rh-106 |
| 45 | Rhodium | Rh-103m | 56 min | gamma | - | - | - | Decay product of Ru-103 |
| 45 | Rhodium | Rh-106 | 30 s | beta | - | - | - | Decay product of Ru-106 |
| 52 | Tellurium | Te-129m | 33.6 d | gamma | 0.3 | 3.00E-09 | 1.40E-08 (large intestine) | - |
| 52 | Tellurium | Te-129 | 1.2 h | beta | - | - | - | Ground state of Te-129m (breaks up into I-129) |
| 52 | Tellurium | Te-132 | 3.2 d | beta | 0.3 | 3.80E-09 | 3,10E-08 (thyroid) | with daughter I-132 |
| 53 | iodine | I-132 | 2.3 h | beta | 1 | 2.90E-10 | 3.40E-09 (thyroid) | Decay product of Te-132 |
| 53 | iodine | I-131 | 8 d | beta | 1 | 2.20E-08 | 4.30E-07 (thyroid) | - |
| 53 | iodine | I-133 | 20.8 h | beta | 1 | 4.30E-09 | 8.20E-08 (thyroid) | with daughter Xe-133 |
| 54 | xenon | Xe-133 | 5.2 d | beta | - | - | - | Noble gas |
| 55 | Cesium | Cs-134 | 2 a | beta | 1 | 1.90E-08 | - | - |
| 55 | Cesium | Cs-136 | 13.1 d | beta | 1 | 3.00E-09 | - | - |
| 55 | Cesium | Cs-137 | 30.2 a | beta | 1 | 1.30E-08 | 1.50E-08 (large intestine) | with daughter Ba-137m |
| 56 | barium | Ba-137m | 2.6 min | gamma | - | - | - | Decay product of Cs-137 |
| 56 | barium | Ba-140 | 12.8 d | beta | 0.2 | 2.60E-09 | 1,70E-08 (large intestine) | with daughter La-140 |
| 57 | Lanthanum | La-140 | 1.7 d | beta | 0.0005 | 2.00E-09 | 1.30E-08 (large intestine) | Decay product of Ba-140 |
| 58 | cerium | Ce-141 | 32.5 d | beta | 0.0005 | 7.10E-10 | 5.50E-09 (large intestine) | |
| 58 | cerium | Ce-144 | 284.7 d | beta | 0.0005 | 5.20E-09 | 4.20E-08 (large intestine) | with secondary products |
| 59 | Praseodymium | Pr-144 | 17.3 min | beta | - | - | - | Decay product of Ce-144 |
| 86 | radon | Rn-222 | 3.8 d | alpha | - | - | - | Decay product of Ra-226 |
| 88 | radium | Ra-226 | 1602 a | alpha | 0.2 | 2.80E-07 | 1.20E-05 (bone surface) | with secondary products |
| 92 | uranium | U-234 | 250,000 a | alpha | 0.02 | 4.90E-08 | 7.80E-07 (bone surface) | with secondary products |
| 92 | uranium | U-235 | 700 mill. A | alpha | 0.02 | 4.70E-08 | 7.40E-07 (bone surface) | with secondary products |
| 92 | uranium | U-238 | 4.5 billion a | alpha | 0.02 | 4.50E-08 | 7.10E-07 (bone surface) | with secondary products |
| 93 | neptunium | Np-239 | 2.4 d | beta | 0.0005 | 8.00E-10 | 6.00E-09 (large intestine) | with secondary products |
| 94 | plutonium | Pu-238 | 87.7 a | alpha | 0.0005 | 2.30E-07 | 7.40E-06 (bone surface) | Decay product of Cm-242 |
| 94 | plutonium | Pu-239 | 24,110 a | alpha | 0.0005 | 2.50E-07 | 8.20E-06 (bone surface) | with secondary products |
| 94 | plutonium | Pu-241 | 14.4 a | beta | 0.0005 | 4.80E-09 | 1.60E-07 (bone surface) | breaks up into Am-241 |
| 95 | Americium | Am-241 | 432.2 a | alpha | 0.0005 | 2.00E-07 | 9.00E-06 (bone surface) | Decay product of Pu-241 |
| 96 | Curium | Cm-242 | 162.8 d | alpha | 0.0005 | 1.20E-08 | 1.90E-07 (bone surface) | with secondary products |
With the help of the dose coefficients, the dose is calculated from the added activity, analogous to inhalation, as follows:
- Body dose (Sv) = added activity (Bq) • dose coefficient (Sv / Bq)
As a rule of thumb, which is based on the radionuclide Sr-90, an intake of 100 Bq leads to an effective subsequent dose of a maximum of 3 µSv (see Fig. 13). Without precise knowledge of the specific radionuclides present, a first quick and conservative assessment can be made. However, the rule of thumb only applies to beta emitters. Alpha emitters must be able to be excluded, because dose values that are up to a factor of 10 are higher. The differences are considerable, if not quite as great as with inhalation.
For the dose uptake by the particularly relevant radionuclide Cs-137 see Fig. 14. The Fig. Applies to constant daily activity intake.
In addition to the contamination levels of the food, which are known as the result of measurement campaigns in a radiological emergency, the activity consumed depends heavily on the respective consumption behavior of a person. The values for the activity intake can be calculated from the individual consumption quantities and the determined activity concentrations (specific activity in Bq / kg or Bq / l). The dose derived from this results in:
- Body dose (Sv) = dose coefficient (Sv / Bq) • activity concentration (Bq / kg) • amount consumed (kg)
| Food | mean consumption rate (kg / a) |
|---|---|
| Drinking water | 350 |
| Milk, dairy products | 130 |
| fish | 7.5 |
| Meat, sausage, eggs | 90 |
| Cereals, cereal products | 110 |
| Local fresh fruit, fruit products, juices | 35 |
| Potatoes, root vegetables, juices | 55 |
| Leafy vegetables | 13 |
| Vegetables, vegetable products, juices | 40 |
Based on mean values for consumption (see Table 5), the European Union has set maximum values for radionuclides in food (Table 6). It is based on the assumption that 10% of the food consumed annually is contaminated. Under this condition, a reference value of 1 mSv per year for the individual effective dose from ingestion is maintained. In the case of Cs-137, this would correspond to a limitation of the supply to an average value in the order of magnitude of 210 Bq per day (see Fig. 14). The European Union maximum levels for baby food are based on different assumptions and are lower.
| Dairy products and liquid foods |
Other foods (unless of minor importance) |
|
|---|---|---|
| Strontium isotopes, especially Sr-90 | 125 | 750 |
| Iodine isotopes, especially I-131 | 500 | 2000 |
| Alpha particle emitting nuclides such as Pu-239 | 20th | 80 |
| All other nuclides with a half-life of over 10 days, especially Cs-134 and Cs-137 |
1000 | 1250 |
The entry into force of limit values within the framework of these maximum values requires the adoption of implementing regulations that relate to a specific emergency. The same applies to import restrictions. Accordingly, in the aftermath of the Chernobyl accident, limit values are still applied in the EU for the placing on the market of radioactively contaminated products (e.g. 600 Bq radiocesium per kg wild boar meat), which go back to import restrictions issued at the time and a recommendation by the European Commission.
rating
For the purposes of this article, the assessment of a radiological emergency consists in the first step of an estimate of the dose values it has caused and, in the second step, of comparing them with the benchmark values defined in the Dose benchmark values section.
As described in the section on dose estimation, the dose values caused are derived from measured values of measured variables such as the
- Local dose rate (in µSv / h),
- Contamination of the soil (in Bq / m²),
- Contamination of the air (in Bq / m³),
- Contamination of food (in Bq / kg, Bq / l) or activity input in Bq / d.
In addition to such measured values, a dose estimation basically requires knowledge of the radionuclides present and thus their properties (dose coefficients, half-lives).
However, even if only activity information is available without precise knowledge of the radionuclides, conservative, i.e. H. evaluations lying on the safe side possible. This is done by assuming maximum dose coefficients (see the rule of thumb in the section on dose estimation) and, at the same time, longevity of the radionuclides.
An assessment is also possible without your own dose estimation, if derived guide values are available. These are derived for the aforementioned measured variables in relation to the relevant basic dose values and certain scenarios. If these scenarios occur, an assessment is possible through a direct comparison of the measured values with the derived guide values. In the evaluation, these take on the role of the actually decisive basic dose values. The SSK has developed such derived guide values as a decision-making aid for certain protective measures.
The attached Fig. 15 illustrates such a rough evaluation of measured variables. It applies to the scenario of a widespread spread of fission products whose composition is unknown. Incorporation of alpha emitters can be excluded.
The value scales of the relevant measured variables are marked with the following basic dose values:
- the natural radiation exposure, here the benchmarks 1, 2a, 2b,
- the occupational limit values, represented by benchmarks 5 and 6,
- the emergency dose values for early emergency management measures with the key values 9 and 10 in conjunction with the SSK guideline values derived for this for local dose rates and soil contamination,
- a mean value derived from benchmark 14 for the daily activity intake through food,
- the reference value for emergency exposures (benchmark 11).
The evaluation of such a scenario is clearly structured. It allows a quick orientation as to whether a radiological situation is comparable with the natural environment, whether it is still within the scope of occupational radiation exposure or whether protective measures would be appropriate depending on the situation. In the context of the conservatism or the relevance of the derived guide values, the orders of magnitude become clear in this way. A classification of measured values based on this model must, however, be checked and updated when new findings become available, in particular with regard to the nuclide composition.
The derived guideline values of the SSK used in Fig. 15 refer to the emergency dose values for the request to stay in buildings (basic dose value 9) and the arrangement of an evacuation (basic dose value 10). In addition, the SSK has derived guide values for protective measures such as contamination controls of people and objects, decontamination, the demarcation of danger areas and a package of agricultural measures.
Protective measures
Protective measures in radiological emergencies range from adapted individual behavior (e.g. restriction of leisure activities), remaining in buildings, precautionary measures in agriculture (e.g. restriction of grazing), restrictions on placing radioactively contaminated food on the market, and taking iodine tablets , Decontamination measures through to evacuation and resettlement. Their aim is to ensure that the dose reference values are observed for individuals in the population and that radiation exposure is further minimized.
Some of these protective measures represent considerable cuts in the lives of citizens and the economy. Their application must be proportionate, i. H. their benefits must be weighed against the costs, including the risks that these measures in turn entail.
From the point of view of the political decision-makers, it is absolutely essential in this area of tension to have a reliable determination and assessment of the situation as a basis for decision-making. For this purpose, an emergency management system was set up in connection with the IMIS. Radiological fundamentals were developed and procedures were created to be able to draw conclusions about body doses from measurement results. The model calculations of the ICRP, numerous publications by the German Radiation Protection Commission, including their analyzes of the effectiveness of individual protective measures, as well as the specialist information systems belonging to IMIS enable a situation assessment and the determination of suitable and appropriate protective measures in a radiological emergency. Furthermore, international agreements have been made to immediately exchange comprehensive information on the cause of a hazard. This is intended to largely ensure forecasts of how the situation is likely to develop further, even if the cause is far away from the home country. However, all of these preparations in our free society are only effective if the citizens are "taken along". Specifically, this means that they have to understand why measures are triggered if necessary. However, you must also be able to understand that a measure may not be necessary under certain circumstances . In this case, they should not be able to be unsettled by scare tactics, improper reporting or unfounded accusations that the authorities are inactive.
With increased expertise, as many citizens as possible
- neither naive nor calculated trivializations nor scaremongering have a chance,
- factually incompetent opinions are recognized,
- appropriate decisions are accepted,
- fear can be reduced.
This also reduces the pressure on decision-makers. You will get back the decision-making leeway that was lost in practice in order to initiate protective measures as a precaution.
literature
Laws and regulations
- Law on the precautionary protection of the population against radiation exposure (Radiation Protection Provision Act - StrVG ) of December 19, 1986 ( Federal Law Gazette I p. 2610 ) (replaced by the StrlSchG with effect from October 1, 2017)
- Law on protection against the harmful effects of ionizing radiation (Radiation Protection Act - StrlSchG ) of June 27, 2017 ( Federal Law Gazette I, p. 1966 )
- Ordinance on protection against the harmful effects of ionizing radiation (Radiation Protection Ordinance - StrlSchV ) of November 29, 2018 ( BGBl. I p. 2034, 2036 )
- Ordinance establishing dose values for early emergency protection measures (Emergency Dose Values Ordinance - NDWV ) of November 29, 2018 ( Federal Law Gazette I p. 2034, 2172 )
- Council Directive 2013/59 / Euratom of December 5, 2013 laying down basic safety standards for protection against the dangers of exposure to ionizing radiation, Official Journal No. L 13 of January 17, 2014 p. 1 ( online )
- Regulation (EURATOM) No. 2016/52 of the Council of 15 January 2016 laying down maximum levels for radioactivity in food and feed in the event of a nuclear accident or other radiological emergency and repealing Regulation (EURATOM) No. 3954/87 of the Council and Regulations (EURATOM) No. 944/89 and (EURATOM) No. 770/90 of the Commission, Official Journal No. L 13 of 20 January 2016 p. 2 ( online )
- General administrative regulation for the integrated measurement and information system for monitoring radioactivity in the environment (IMIS) according to the Radiation Protection Precautionary Act (AVV-IMIS) of December 13, 2006, BAnz. 2006, No. 244a ( PDF ; 1.15 MB)
Technical / radiological basics
- List of isotopes in: Wikipedia, The Free Encyclopedia
- International Commission on Radiological Protection (ICRP): The 2007 Recommendations of the International Commission on Radiological Protection , ICRP Publication 103, Ann. ICRP 37 (2-4), 2007, online , German edition published by the Federal Office for Radiation Protection ( PDF ; 2.2 MB)
- International Commission on Radiological Protection (ICRP): Application of the Commission's Recommendations to the Protection of People Living in Long-term Contaminated Areas after a Nuclear Accident or a Radiation Emergency , ICRP Publication 111, Ann. ICRP 39 (3), 2009, ( PDF ; 658 kB)
- Radiation Protection Commission: Framework recommendations for disaster control in the vicinity of nuclear facilities , recommendation adopted at the 274th SSK meeting on 19/20 February 2015 ( PDF ; 115 kB)
- Dose coefficients for calculating radiation exposure , Federal Gazette 160 a and b, August 28, 2001 ( online )
- ICRP: Data Viewer for dose coefficients for internal radiation exposure at the workplace, electronic annex for ICRP publications 134, 137 and 141, Occupational Intakes of Radionuclides - Part 2, 3 and 4, zip file for download , (executable file, installed 85.4 MB)
- National Nuclear Data Center (NNDC): NuDat, NSR, XUNDL, ENSDF, MIRD, ENDF, CSISRS, Sigma, Chart of Nuclides , etc., collection of online databases
- Radiation Protection Commission: Radiological basis for decisions on measures to protect the population in the event of incidents with the release of radionuclides , recommendation adopted in the 268th meeting of the SSK on 13./14. February 2014, BAnz AT November 18 , 2014 B5 , ( PDF ; 722 kB)
- Radiation Protection Commission (SSK): Derived reference values for measures to protect the population in the event of incidents with the release of radionuclides , recommendation adopted in the 303rd meeting of the SSK on 24/25. October 2019, BAnz AT 04/22/2020 B3 , ( PDF ; 1.82 MB)
- Radiation Protection Commission: Overview of measures to reduce radiation exposure after events with not inconsiderable radiological effects , catalog of measures, recommendation of the Radiation Protection Commission from 5./6. December 2007, Reports of the Radiation Protection Commission, Volume 60 (print medium)
- Radiation Protection Commission: Use of iodine tablets to block the thyroid gland in an emergency with the release of radioactive iodine , recommendation adopted at the 294th meeting of the SSK on April 26, 2018, changed at the 298th meeting of the SSK on February 6, 2019, BAnz AT 07.05 .2019 B4 , ( PDF ; 509 kB)
- Radiation Protection Commission: Use of particle-filtering half masks in emergency protection , recommendation adopted in the 300th session of the Radiation Protection Commission on 27./28. June 2019, BAnz AT 29.01.2020 B4 , ( PDF ; 959 kB)
Reports
- United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR): Health effects due to radiation from the Chernobyl accident , in: UNSCEAR Report, 2008, Annex D, Key Chernobyl links .
- Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety (BMBU): Environmental radioactivity and radiation exposure: Annual report 2012 (general report) , June 2014, URN: nbn: de: 0221-2013090511044 ( PDF ; 6.22 MB).
- Society for Reactor Safety (GRS): The accident in the Chernobyl nuclear power plant , GRS-S-039, June 1986 ( PDF ; 4.12 MB).
- Society for Reactor Safety (GRS): Recent findings on the accident at the Chernobyl nuclear power plant , GRS-S-40, February 1987 ( PDF ; 39.49 MB).
- Federal Office for Radiation Protection (BfS): The 1986 reactor accident in Chernobyl , 2011 ( PDF ; 2.79 MB).
- German Atomic Forum V .: The reactor accident in Chernobyl , April 2011, Unchanged reprint April 2015 ( PDF ; 1.7 MB).
- United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR): UNSCEAR 2013 Report to the General Assembly , VOLUME I, Scientific Annex A, Levels and effects of radiation exposure due to the nuclear accident after the 2011 great east-Japan earthquake and tsunami , United Nations, New York 2014 ( PDF ; 5.8 MB); Fact sheet for this report ( PDF ; 278 kB); First white paper on this report, Vienna 2015 ( PDF ; 1.25 MB); Second white paper for this report, New York 2016 ( PDF ; 917 kB).
- International Commission on Radiological Protection (ICRP): Experience and current issues with recovery management from the Fukushima accident , Michiaki KAI, presentation at the 2nd International Symposium on the System of Radiological Protection , Abu Dabi 22.-24. October 2013 ( PDF ; 604 kB).
- International Atomic Energy Agency (IAEA): The Fukushima Daiichi Accident , Technical Volume 4/5, Radiological Consequences , Vienna 2015 ( PDF ; 19.4 MB).
Individual evidence
- ↑ BfS: What is an emergency?
- ↑ a b c BMU: Environmental radioactivity and radiation exposure in 2011: Information from the Federal Government , Parliamentary Report 2011, June 4, 2013, URN: nbn: de: 0221-2013060410695
- ↑ a b Radiation Protection Commission (SSK): Radiological basis for decisions on measures to protect the population in the event of incidents with the release of radionuclides , recommendation adopted at the 268th meeting of the SSK on 13/14 February 2014
- ↑ a b Radiation Protection Commission (SSK): Derived reference values for measures to protect the population in the event of incidents with the release of radionuclides , recommendation adopted in the 303rd meeting of the SSK on 24/25. October 2019
- ↑ Map of the local dose rate
- ^ GRS: The accident in the Chernobyl nuclear power plant , GRS-S-039, June 1986
- ^ BfS: The 1986 reactor accident in Chernobyl , 2011
- ↑ Regulation (EURATOM) No. 2016/52
- ↑ Recommendation of the Commission of April 14, 2003 on the protection and information of the population with regard to exposure from the continued contamination of certain wild foods with radioactive cesium as a result of the accident at the Chernobyl nuclear power plant (2003/274 / EC)
Web links
- Federal Office for Radiation Protection on ionizing radiation and nuclear emergency protection
- Nuclear emergency protection , subpage of the information portal "Nuclear Safety" of the federal and state governments
- Radiation Protection Commission with numerous other web links
- International Commission on Radiological Protection (ICRP)
- International Atomic Energy Agency (IAEA)
- National Nuclear Data Center (NNDC) with Nuclear Structure & Decay Data
- United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR)