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sievert
Display of background radiation in a hotel at Naraha, Japan, showing dose rate in microsieverts per hour, five years after the Fukushima disaster
General information
Unit systemSI
Unit ofStochastic health effect of ionizing radiation (equivalent dose)
SymbolSv
Named afterRolf Maximilian Sievert
Conversions
1 Sv in ...... is equal to ...
   SI base units   m2⋅s−2
   Sv indicates absorbed dose modified by weighting factors.   J⋅kg−1
   CGS units (non-SI)   100 rem

The sievert (symbol: Sv[note 1]) is a derived unit in the International System of Units (SI) intended to represent the stochastic health risk of ionizing radiation, which is defined as the probability of causing radiation-induced cancer and genetic damage. The sievert is important in dosimetry and radiation protection. It is named after Rolf Maximilian Sievert, a Swedish medical physicist renowned for work on radiation dose measurement and research into the biological effects of radiation.

The sievert unit is used for radiation dose quantities such as equivalent dose and effective dose, which represent the risk of external radiation from sources outside the body, and committed dose, which represents the risk of internal irradiation due to inhaled or ingested radioactive substances. According to the International Commission on Radiological Protection (ICRP), one sievert results in a 5.5% probability of eventually developing fatal cancer based on the disputed linear no-threshold model of ionizing radiation exposure.[1][2]

To calculate the value of stochastic health risk in sieverts, the physical quantity absorbed dose is converted into equivalent dose and effective dose by applying factors for radiation type and biological context, published by the ICRP and the International Commission on Radiation Units and Measurements (ICRU). One sievert equals 100 rem, which is an older, CGS radiation unit.

Conventionally, deterministic health effects due to acute tissue damage that is certain to happen, produced by high dose rates of radiation, are compared to the physical quantity absorbed dose measured by the unit gray (Gy).[3]

Definition

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CIPM definition of the sievert

[edit]

The SI definition given by the International Committee for Weights and Measures (CIPM) says:

"The quantity dose equivalent H is the product of the absorbed dose D of ionizing radiation and the dimensionless factor Q (quality factor) defined as a function of linear energy transfer by the ICRU"

H = Q × D[4]

The value of Q is not defined further by CIPM, but it requires the use of the relevant ICRU recommendations to provide this value.

The CIPM also says that "in order to avoid any risk of confusion between the absorbed dose D and the dose equivalent H, the special names for the respective units should be used, that is, the name gray should be used instead of joules per kilogram for the unit of absorbed dose D and the name sievert instead of joules per kilogram for the unit of dose equivalent H".[4]

In summary:

gray: quantity D—absorbed dose
1 Gy = 1 joule/kilogram—a physical quantity. 1 Gy is the deposit of a joule of radiation energy per kilogram of matter or tissue.
sievert: quantity H—equivalent dose
1 Sv = 1 joule/kilogram—a biological effect. The sievert represents the equivalent biological effect of the deposit of a joule of radiation energy in a kilogram of human tissue. The ratio to absorbed dose is denoted by Q.

ICRP definition of the sievert

[edit]

The ICRP definition of the sievert is:[5]

"The sievert is the special name for the SI unit of equivalent dose, effective dose, and operational dose quantities. The unit is joule per kilogram."

The sievert is used for a number of dose quantities which are described in this article and are part of the international radiological protection system devised and defined by the ICRP and ICRU.

External dose quantities

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External radiation dose quantities used in radiological protection

When the sievert is used to represent the stochastic effects of external ionizing radiation on human tissue, the radiation doses received are measured in practice by radiometric instruments and dosimeters and are called operational quantities. To relate these actual received doses to likely health effects, protection quantities have been developed to predict the likely health effects using the results of large epidemiological studies. Consequently, this has required the creation of a number of different dose quantities within a coherent system developed by the ICRU working with the ICRP.

The external dose quantities and their relationships are shown in the accompanying diagram. The ICRU is primarily responsible for the operational dose quantities, based upon the application of ionising radiation metrology, and the ICRP is primarily responsible for the protection quantities, based upon modelling of dose uptake and biological sensitivity of the human body.

Naming conventions

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The ICRU/ICRP dose quantities have specific purposes and meanings, but some use common words in a different order. There can be confusion between, for instance, equivalent dose and dose equivalent.

Although the CIPM definition states that the linear energy transfer function (Q) of the ICRU is used in calculating the biological effect, the ICRP in 1990[6] developed the "protection" dose quantities effective and equivalent dose which are calculated from more complex computational models and are distinguished by not having the phrase dose equivalent in their name. Only the operational dose quantities which still use Q for calculation retain the phrase dose equivalent. However, there are joint ICRU/ICRP proposals to simplify this system by changes to the operational dose definitions to harmonise with those of protection quantities. These were outlined at the 3rd International Symposium on Radiological Protection in October 2015, and if implemented would make the naming of operational quantities more logical by introducing "dose to lens of eye" and "dose to local skin" as equivalent doses.[7]

In the USA there are differently named dose quantities which are not part of the ICRP nomenclature.[8]

Physical quantities

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These are directly measurable physical quantities in which no allowance has been made for biological effects. Radiation fluence is the number of radiation particles impinging per unit area per unit time, kerma is the ionising effect on air of gamma rays and X-rays and is used for instrument calibration, and absorbed dose is the amount of radiation energy deposited per unit mass in the matter or tissue under consideration.

Operational quantities

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Operational quantities are measured in practice, and are the means of directly measuring dose uptake due to exposure, or predicting dose uptake in a measured environment. In this way they are used for practical dose control, by providing an estimate or upper limit for the value of the protection quantities related to an exposure. They are also used in practical regulations and guidance.[9]

The calibration of individual and area dosimeters in photon fields is performed by measuring the collision "air kerma free in air" under conditions of secondary electron equilibrium. Then the appropriate operational quantity is derived applying a conversion coefficient that relates the air kerma to the appropriate operational quantity. The conversion coefficients for photon radiation are published by the ICRU.[10]

Simple (non-anthropomorphic) "phantoms" are used to relate operational quantities to measured free-air irradiation. The ICRU sphere phantom is based on the definition of an ICRU 4-element tissue-equivalent material which does not really exist and cannot be fabricated.[11] The ICRU sphere is a theoretical 30 cm diameter "tissue equivalent" sphere consisting of a material with a density of 1 g·cm−3 and a mass composition of 76.2% oxygen, 11.1% carbon, 10.1% hydrogen and 2.6% nitrogen. This material is specified to most closely approximate human tissue in its absorption properties. According to the ICRP, the ICRU "sphere phantom" in most cases adequately approximates the human body as regards the scattering and attenuation of penetrating radiation fields under consideration.[12] Thus radiation of a particular energy fluence will have roughly the same energy deposition within the sphere as it would in the equivalent mass of human tissue.[13]

To allow for back-scattering and absorption of the human body, the "slab phantom" is used to represent the human torso for practical calibration of whole body dosimeters. The slab phantom is 300 mm × 300 mm × 150 mm depth to represent the human torso.[13]

The joint ICRU/ICRP proposals outlined at the 3rd International Symposium on Radiological Protection in October 2015 to change the definition of operational quantities would not change the present use of calibration phantoms or reference radiation fields.[7]

Protection quantities

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Protection quantities are calculated models, and are used as "limiting quantities" to specify exposure limits to ensure, in the words of ICRP, "that the occurrence of stochastic health effects is kept below unacceptable levels and that tissue reactions are avoided".[14][15][13] These quantities cannot be measured in practice but their values are derived using models of external dose to internal organs of the human body, using anthropomorphic phantoms. These are 3D computational models of the body which take into account a number of complex effects such as body self-shielding and internal scattering of radiation. The calculation starts with organ absorbed dose, and then applies radiation and tissue weighting factors.[16]

As protection quantities cannot practically be measured, operational quantities must be used to relate them to practical radiation instrument and dosimeter responses.[17]

Instrument and dosimetry response

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This is an actual reading obtained from such as an ambient dose gamma monitor, or a personal dosimeter. Such instruments are calibrated using radiation metrology techniques which will trace them to a national radiation standard, and thereby relate them to an operational quantity. The readings of instruments and dosimeters are used to prevent the uptake of excessive dose and to provide records of dose uptake to satisfy radiation safety legislation; such as in the UK, the Ionising Radiations Regulations 1999.

Calculating protection dose quantities

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Graphic showing relationship of "protection dose" quantities in SI units

The sievert is used in external radiation protection for equivalent dose (the external-source, whole-body exposure effects, in a uniform field), and effective dose (which depends on the body parts irradiated).

These dose quantities are weighted averages of absorbed dose designed to be representative of the stochastic health effects of radiation, and use of the sievert implies that appropriate weighting factors have been applied to the absorbed dose measurement or calculation (expressed in grays).[1]

The ICRP calculation provides two weighting factors to enable the calculation of protection quantities.

 1. The radiation factor WR, which is specific for radiation type R – This is used in calculating the equivalent dose HT which can be for the whole body or for individual organs.
 2. The tissue weighting factor WT, which is specific for tissue type T being irradiated. This is used with WR to calculate the contributory organ doses to arrive at an effective dose E for non-uniform irradiation.

When a whole body is irradiated uniformly only the radiation weighting factor WR is used, and the effective dose equals the whole body equivalent dose. But if the irradiation of a body is partial or non-uniform the tissue factor WT is used to calculate dose to each organ or tissue. These are then summed to obtain the effective dose. In the case of uniform irradiation of the human body, these summate to 1, but in the case of partial or non-uniform irradiation, they will summate to a lower value depending on the organs concerned; reflecting the lower overall health effect. The calculation process is shown on the accompanying diagram. This approach calculates the biological risk contribution to the whole body, taking into account complete or partial irradiation, and the radiation type or types.

The values of these weighting factors are conservatively chosen to be greater than the bulk of experimental values observed for the most sensitive cell types, based on averages of those obtained for the human population.

Radiation type weighting factor WR

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Main article: Equivalent dose

Since different radiation types have different biological effects for the same deposited energy, a corrective radiation weighting factor WR, which is dependent on the radiation type and on the target tissue, is applied to convert the absorbed dose measured in the unit gray to determine the equivalent dose. The result is given the unit sievert.

Radiation weighting factors WR
used to represent relative biological effectiveness
according to ICRP report 103[1]
Radiation Energy (E) WR (formerly Q)
x-rays, gamma rays,
beta particles, muons		 1
neutrons < 1 MeV 2.5 + 18.2e−[ln(E)]2/6
1 – 50 MeV 5.0 + 17.0e−[ln(2E)]2/6
> 50 MeV 2.5 + 3.25e−[ln(0.04E)]2/6
protons, charged pions 2
alpha particles,
nuclear fission products,
heavy nuclei		 20

The equivalent dose is calculated by multiplying the absorbed energy, averaged by mass over an organ or tissue of interest, by a radiation weighting factor appropriate to the type and energy of radiation. To obtain the equivalent dose for a mix of radiation types and energies, a sum is taken over all types of radiation energy dose.[1]

where

HT is the equivalent dose absorbed by tissue T,
DT,R is the absorbed dose in tissue T by radiation type R and
WR is the radiation weighting factor defined by regulation.

Thus for example, an absorbed dose of 1 Gy by alpha particles will lead to an equivalent dose of 20 Sv.

The radiation weighting factor for neutrons has been revised over time and remains controversial.

This may seem to be a paradox. It implies that the energy of the incident radiation field in joules has increased by a factor of 20, thereby violating the laws of conservation of energy. However, this is not the case. The sievert is used only to convey the fact that a gray of absorbed alpha particles would cause twenty times the biological effect of a gray of absorbed x-rays. It is this biological component that is being expressed when using sieverts rather than the actual energy delivered by the incident absorbed radiation.

Tissue type weighting factor WT

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Main article: effective dose (radiation)

The second weighting factor is the tissue factor WT, but it is used only if there has been non-uniform irradiation of a body. If the body has been subject to uniform irradiation, the effective dose equals the whole body equivalent dose, and only the radiation weighting factor WR is used. But if there is partial or non-uniform body irradiation the calculation must take account of the individual organ doses received, because the sensitivity of each organ to irradiation depends on their tissue type. This summed dose from only those organs concerned gives the effective dose for the whole body. The tissue weighting factor is used to calculate those individual organ dose contributions.

The ICRP values for WT are given in the table shown here.

Weighting factors for different organs[18]
Organs Tissue weighting factors
ICRP26
1977		 ICRP60
1990[19]		 ICRP103
2007[1]
Gonads 0.25 0.20 0.08
Red bone marrow 0.12 0.12 0.12
Colon — 0.12 0.12
Lung 0.12 0.12 0.12
Stomach — 0.12 0.12
Breasts 0.15 0.05 0.12
Bladder — 0.05 0.04
Liver — 0.05 0.04
Oesophagus — 0.05 0.04
Thyroid 0.03 0.05 0.04
Skin — 0.01 0.01
Bone surface 0.03 0.01 0.01
Salivary glands — — 0.01
Brain — — 0.01
Remainder of body 0.30 0.05 0.12
Total 1.00 1.00 1.00

The article on effective dose gives the method of calculation. The absorbed dose is first corrected for the radiation type to give the equivalent dose, and then corrected for the tissue receiving the radiation. Some tissues like bone marrow are particularly sensitive to radiation, so they are given a weighting factor that is disproportionally large relative to the fraction of body mass they represent. Other tissues like the hard bone surface are particularly insensitive to radiation and are assigned a disproportionally low weighting factor.

In summary, the sum of tissue-weighted doses to each irradiated organ or tissue of the body adds up to the effective dose for the body. The use of effective dose enables comparisons of overall dose received regardless of the extent of body irradiation.

Relation between some ionizing radiation units[20]

Operational quantities

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The operational quantities are used in practical applications for monitoring and investigating external exposure situations. They are defined for practical operational measurements and assessment of doses in the body.[5] Three external operational dose quantities were devised to relate operational dosimeter and instrument measurements to the calculated protection quantities. Also devised were two phantoms, The ICRU "slab" and "sphere" phantoms which relate these quantities to incident radiation quantities using the Q(L) calculation.

Ambient dose equivalent

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This is used for area monitoring of penetrating radiation and is usually expressed as the quantity H*(10). This means the radiation is equivalent to that found 10 mm within the ICRU sphere phantom in the direction of origin of the field.[21] An example of penetrating radiation is gamma rays.

Directional dose equivalent

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This is used for monitoring of low penetrating radiation and is usually expressed as the quantity H'(0.07). This means the radiation is equivalent to that found at a depth of 0.07 mm in the ICRU sphere phantom.[22] Examples of low penetrating radiation are alpha particles, beta particles and low-energy photons. This dose quantity is used for the determination of equivalent dose to such as the skin, lens of the eye.[23] In radiological protection practice value of omega is usually not specified as the dose is usually at a maximum at the point of interest.

Personal dose equivalent

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This is used for individual dose monitoring, such as with a personal dosimeter worn on the body. The recommended depth for assessment is 10 mm which gives the quantity Hp(10).[24]

Proposals for changing the definition of protection dose quantities

[edit]

In order to simplify the means of calculating operational quantities and assist in the comprehension of radiation dose protection quantities, ICRP Committee 2 & ICRU Report Committee 26 started in 2010 an examination of different means of achieving this by dose coefficients related to Effective Dose or Absorbed Dose.

Specifically;

1. For area monitoring of effective dose of whole body it would be:

H = Φ × conversion coefficient

The driver for this is that H∗(10) is not a reasonable estimate of effective dose due to high energy photons, as a result of the extension of particle types and energy ranges to be considered in ICRP report 116. This change would remove the need for the ICRU sphere and introduce a new quantity called Emax.

2. For individual monitoring, to measure deterministic effects on eye lens and skin, it would be:

D = Φ × conversion coefficient for absorbed dose.

The driver for this is the need to measure the deterministic effect, which it is suggested, is more appropriate than stochastic effect. This would calculate equivalent dose quantities Hlens and Hskin.

This would remove the need for the ICRU Sphere and the Q-L function. Any changes would replace ICRU report 51, and part of report 57.[7]

A final draft report was issued in July 2017 by ICRU/ICRP for consultation.[25]

Internal dose quantities

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Main article: Committed dose

The sievert is used for human internal dose quantities in calculating committed dose. This is dose from radionuclides which have been ingested or inhaled into the human body, and thereby "committed" to irradiate the body for a period of time. The concepts of calculating protection quantities as described for external radiation applies, but as the source of radiation is within the tissue of the body, the calculation of absorbed organ dose uses different coefficients and irradiation mechanisms.

The ICRP defines Committed effective dose, as the sum of the products of the committed organ or tissue equivalent doses and the appropriate tissue weighting factors , where is the integration time in years following the intake. The commitment period is taken to be 50 years for adults, and to age 70 years for children.[5]

The ICRP further states "For internal exposure, committed effective doses are generally determined from an assessment of the intakes of radionuclides from bioassay measurements or other quantities (e.g., activity retained in the body or in daily excreta). The radiation dose is determined from the intake using recommended dose coefficients".[26]

A committed dose from an internal source is intended to carry the same effective risk as the same amount of equivalent dose applied uniformly to the whole body from an external source, or the same amount of effective dose applied to part of the body.

Health effects

[edit]
Further information: Radiobiology

Ionizing radiation has deterministic and stochastic effects on human health. Deterministic (acute tissue effect) events happen with certainty, with the resulting health conditions occurring in every individual who received the same high dose. Stochastic (cancer induction and genetic) events are inherently random, with most individuals in a group failing to ever exhibit any causal negative health effects after exposure, while an indeterministic random minority do, often with the resulting subtle negative health effects being observable only after large detailed epidemiology studies.

The use of the sievert implies that only stochastic effects are being considered, and to avoid confusion deterministic effects are conventionally compared to values of absorbed dose expressed by the SI unit gray (Gy).

Stochastic effects

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Stochastic effects are those that occur randomly, such as radiation-induced cancer. The consensus of nuclear regulators, governments and the UNSCEAR is that the incidence of cancers due to ionizing radiation can be modeled as increasing linearly with effective dose at a rate of 5.5% per sievert.[1] This is known as the linear no-threshold model (LNT model). Some argue that this LNT model is now outdated and should be replaced with a threshold below which the body's natural cell processes repair damage and/or replace damaged cells.[27][28] There is general agreement that the risk is much higher for infants and fetuses than adults, higher for the middle-aged than for seniors, and higher for women than for men, though there is no quantitative consensus about this.[29][30]

Deterministic effects

[edit]
This is a graph depicting the effect of dose fractionation on the ability of gamma rays to cause cell death. The blue line is for cells which were not given a chance to recover; the radiation was delivered in one session. The red line is for cells which were allowed to stand for a time and recover with the pause in delivery conferring radioresistance.

The deterministic (acute tissue damage) effects that can lead to acute radiation syndrome only occur in the case of acute high doses (≳ 0.1 Gy) and high dose rates (≳ 0.1 Gy/h) and are conventionally not measured using the unit sievert, but use the unit gray (Gy). A model of deterministic risk would require different weighting factors (not yet established) than are used in the calculation of equivalent and effective dose.

ICRP dose limits

[edit]

The ICRP recommends a number of limits for dose uptake in table 8 of report 103. These limits are "situational", for planned, emergency and existing situations. Within these situations, limits are given for the following groups:[31]

  • Planned exposure – limits given for occupational, medical and public
  • Emergency exposure – limits given for occupational and public exposure
  • Existing exposure – All persons exposed

For occupational exposure, the limit is 50 mSv in a single year with a maximum of 100 mSv in a consecutive five-year period, and for the public to an average of 1 mSv (0.001 Sv) of effective dose per year, not including medical and occupational exposures.[1]

For comparison, natural radiation levels inside the United States Capitol are such that a human body would receive an additional dose rate of 0.85 mSv/a, close to the regulatory limit, because of the uranium content of the granite structure.[32] According to the conservative ICRP model, someone who spent 20 years inside the capitol building would have an extra one in a thousand chance of getting cancer, over and above any other existing risk (calculated as: 20 a·0.85 mSv/a·0.001 Sv/mSv·5.5%/Sv ≈ 0.1%). However, that "existing risk" is much higher; an average American would have a 10% chance of getting cancer during this same 20-year period, even without any exposure to artificial radiation (see natural Epidemiology of cancer and cancer rates).

Dose examples

[edit]
Main article: Orders of magnitude (radiation)
US Department of Energy 2010 dose chart in sieverts for a variety of situations and applications[33]
Various doses of radiation in sieverts, ranging from trivial to lethal, expressed as comparative areas
Comparison of radiation doses – includes the amount detected on the trip from Earth to Mars by the RAD on the MSL (2011–2013).[34][35][36][37]

Significant radiation doses are not frequently encountered in everyday life. The following examples can help illustrate relative magnitudes; these are meant to be examples only, not a comprehensive list of possible radiation doses. An "acute dose" is one that occurs over a short and finite period of time, while a "chronic dose" is a dose that continues for an extended period of time so that it is better described by a dose rate.

Dose examples

[edit]
98 nSv: Banana equivalent dose, an illustrative unit of radiation dose representing the measure of radiation from a typical 150 g banana[38][a]
250 nSv: U.S. limit on effective dose for general-use x-ray security screening systems such as those previously used in airport security screening[39]
5–10 μSv: One set of dental radiographs[40]
80 μSv: Average (one time) dose to people living within 10 mi (16 km) of the plant during the Three Mile Island accident[41]
400–600 μSv: Two-view mammogram, using weighting factors updated in 2007[42]
1 mSv: U.S. 10 CFR § 20.1301(a)(1) dose limit for individual members of the public, total effective dose equivalent, per annum[43]
1.5–1.7 mSv: Annual occupational dose for flight attendants[44]
2–7 mSv: Barium fluoroscopy, e.g. Barium meal, up to 2 minutes, 4–24 spot images[45]
10–30 mSv: Single full-body CT scan[46][47]
50 mSv: U.S. 10 C.F.R. § 20.1201(a)(1)(i) occupational dose limit, total effective dose equivalent, per annum[48]
68 mSv: Estimated maximum dose to evacuees who lived closest to the Fukushima I nuclear accidents[49]
80 mSv: 6-month stay on the International Space Station
160 mSv: Chronic dose to lungs over one year smoking 1.5 packs of cigarettes per day, mostly due to inhalation of Polonium-210 and Lead-210[50][51]
250 mSv: 6-month trip to Mars—radiation due to cosmic rays, which are very difficult to shield against
400 mSv: Average accumulated exposure of residents over a period of 9–20 years, who suffered no ill effects, in apartments in Taiwan constructed with rebar containing Cobalt-60[52]
500 mSv: The U.S. 10 C.F.R. § 20.1201(a)(2)(ii) occupational dose limit, shallow-dose equivalent to skin, per annum[48]
670 mSv: Highest dose received by a worker responding to the Fukushima emergency[53][a]
1 Sv: Maximum allowed radiation exposure for NASA astronauts over their career[34]
4–5 Sv: Dose required to kill a human with a 50% risk within 30 days (LD50/30), if the dose is received over a very short duration[54][33]
5 Sv: Calculated dose from the neutron and gamma ray flash, 1.2 km from ground zero of the Little Boy fission bomb, air burst at 600 m.[55][56]
4.5–6 Sv: Fatal acute doses during Goiânia accident
5.1 Sv: Fatal acute dose to Harry Daghlian in 1945 criticality accident[57]
10 to 17 Sv: Fatal acute doses during Tokaimura nuclear accident. Hisashi Ouchi who received 17 Sv lived for 83 days after the accident.[58]
21 Sv: Fatal acute dose to Louis Slotin in 1946 criticality accident[57]
36 Sv: Fatal acute dose to Cecil Kelley in 1958, death occurred within 35 hours.[59]
54 Sv: Fatal acute dose to Boris Korchilov in 1961 after a reactor cooling system failed on the Soviet submarine K-19 which required work in the reactor with no shielding[60]
64 Sv: Nonfatal dose to Albert Stevens spread over ≈21 years, due to a 1945 plutonium injection experiment by doctors working on the secret Manhattan Project.[61][a]

Dose rate examples

[edit]

All conversions between hours and years have assumed continuous presence in a steady field, disregarding known fluctuations, intermittent exposure and radioactive decay. Converted values are shown in parentheses. "/a" is "per annum", which means per year. "/h" means "per hour".

<1 mSv/a <100 nSv/h Steady dose rates below 100 nSv/h are difficult to measure.[citation needed]
1 mSv/a (100 nSv/h avg) ICRP recommended maximum for external irradiation of the human body, excluding medical and occupational exposures.
2.4 mSv/a (270 nSv/h avg) Human exposure to natural background radiation, global average[a]
(8 mSv/a) 810 nSv/h avg Next to the Chernobyl New Safe Confinement (May 2019)[62]
~8 mSv/a (~900 nSv/h avg) Average natural background radiation in Finland[63]
24 mSv/a (2.7 μSv/h avg) Natural background radiation at airline cruise altitude[64][b]
(46 mSv/a) 5.19 μSv/h avg Next to Chernobyl Nuclear Power Plant, before installing the New Sarcophagus in November 2016[65]
130 mSv/a (15 μSv/h avg) Ambient field inside most radioactive house in Ramsar, Iran[66][c]
(350 mSv/a) 39.8 μSv/h avg Inside "The Claw" of Chernobyl[67]
(800 mSv/a) 90 μSv/h Natural radiation on a monazite beach near Guarapari, Brazil.[68]
(9 Sv/a) 1 mSv/h NRC definition of a high radiation area in a nuclear power plant, warranting a chain-link fence[69]
(17–173 Sv/a) 2–20 mSv/h Typical dose rate for activated reactor wall in possible future fusion reactors after 100 years.[70] After approximately 300 years of decay the fusion waste would produce the same dose rate as exposure to coal ash, with the volume of fusion waste naturally being orders of magnitude less than from coal ash.[71] Immediate predicted activation is 90 MGy/a.[citation needed]
(1.7 kSv/a) 190 mSv/h Highest reading from fallout of the Trinity bomb, 20 mi (32 km) away, 3 hours after detonation.[72][c]
(2.3 MSv/a) 270 Sv/h Typical PWR spent fuel waste, after 10-year cooldown, no shielding and no distance.[73]
(4.6–5.6 MSv/a) 530–650 Sv/h The radiation level inside the primary containment vessel of the second BWR-reactor of the Fukushima power station, in February 2017, six years after a suspected meltdown.[74][75][76][77][78] In this environment, it takes between 22 and 34 seconds to accumulate a median lethal dose (LD50/30).

Notes on examples:

  1. ^ a b c d Noted figures are dominated by a committed dose which gradually turned into effective dose over an extended period of time. Therefore the true acute dose must be lower, but standard dosimetry practice is to account committed doses as acute in the year the radioisotopes are taken into the body.
  2. ^ The dose rate received by air crews is highly dependent on the radiation weighting factors chosen for protons and neutrons, which have changed over time and remain controversial.
  3. ^ a b Noted figures exclude any committed dose from radioisotopes taken into the body. Therefore the total radiation dose would be higher unless respiratory protection was used.

History

[edit]

The sievert has its origin in the röntgen equivalent man (rem) which was derived from CGS units. The International Commission on Radiation Units and Measurements (ICRU) promoted a switch to coherent SI units in the 1970s,[79] and announced in 1976 that it planned to formulate a suitable unit for equivalent dose.[80] The ICRP pre-empted the ICRU by introducing the sievert in 1977.[81]

The sievert was adopted by the International Committee for Weights and Measures (CIPM) in 1980, five years after adopting the gray. The CIPM then issued an explanation in 1984, recommending when the sievert should be used as opposed to the gray. That explanation was updated in 2002 to bring it closer to the ICRP's definition of equivalent dose, which had changed in 1990. Specifically, the ICRP had introduced equivalent dose, renamed the quality factor (Q) to radiation weighting factor (WR), and dropped another weighting factor "N" in 1990. In 2002, the CIPM similarly dropped the weighting factor "N" from their explanation but otherwise kept other old terminology and symbols. This explanation only appears in the appendix to the SI brochure and is not part of the definition of the sievert.[82]

Common SI usage

[edit]

The sievert is named after Rolf Maximilian Sievert. As with every SI unit named after a person, its symbol starts with an upper case letter (Sv), but when written in full, it follows the rules for capitalisation of a common noun; i.e., sievert becomes capitalised at the beginning of a sentence and in titles but is otherwise in lower case.

Frequently used SI prefixes are the millisievert (1 mSv = 0.001 Sv) and microsievert (1 μSv = 0.000 001 Sv) and commonly used units for time derivative or "dose rate" indications on instruments and warnings for radiological protection are μSv/h and mSv/h. Regulatory limits and chronic doses are often given in units of mSv/a or Sv/a, where they are understood to represent an average over the entire year. In many occupational scenarios, the hourly dose rate might fluctuate to levels thousands of times higher for a brief period of time, without infringing on the annual limits. The conversion from hours to years varies because of leap years and exposure schedules, but approximate conversions are:

1 mSv/h = 8.766 Sv/a
114.1 μSv/h = 1 Sv/a

Conversion from hourly rates to annual rates is further complicated by seasonal fluctuations in natural radiation, decay of artificial sources, and intermittent proximity between humans and sources. The ICRP once adopted fixed conversion for occupational exposure, although these have not appeared in recent documents:[83]

8 h = 1 day
40 h = 1 week
50 weeks = 1 year

Therefore, for occupation exposures of that time period,

1 mSv/h = 2 Sv/a
500 μSv/h = 1 Sv/a

Ionizing radiation quantities

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Graphic showing relationships between radioactivity and detected ionizing radiation

The following table shows radiation quantities in SI and non-SI units:

Ionizing radiation related quantities
Quantity Unit Symbol Derivation Year SI equivalent
Activity (A) becquerel Bq s−1 1974 SI unit
curie Ci 3.7×1010 s−1 1953 3.7×1010 Bq
rutherford Rd 106 s−1 1946 1000000 Bq
Exposure (X) coulomb per kilogram C/kg C⋅kg−1 of air 1974 SI unit
röntgen R esu / 0.001293 g of air 1928 2.58×10−4 C/kg
Absorbed dose (D) gray Gy J⋅kg−1 1974 SI unit
erg per gram erg/g erg⋅g−1 1950 1.0×10−4 Gy
rad rad 100 erg⋅g−1 1953 0.010 Gy
Equivalent dose (H) sievert Sv J⋅kg−1 × WR 1977 SI unit
röntgen equivalent man rem 100 erg⋅g−1 × WR 1971 0.010 Sv
Effective dose (E) sievert Sv J⋅kg−1 × WR × WT 1977 SI unit
röntgen equivalent man rem 100 erg⋅g−1 × WR × WT 1971 0.010 Sv

Although the United States Nuclear Regulatory Commission permits the use of the units curie, rad, and rem alongside SI units,[84] the European Union European units of measurement directives required that their use for "public health ... purposes" be phased out by 31 December 1985.[85]

Rem equivalence

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An older unit for the dose equivalent is the rem,[86] still often used in the United States. One sievert is equal to 100 rem:

100.0000 rem = 100,000.0 mrem = 1 Sv = 1.000000 Sv = 1000.000 mSv = 1,000,000 μSv
1.0000 rem = 1000.0 mrem = 1 rem = 0.010000 Sv = 10.000 mSv = 10000 μSv
0.1000 rem = 100.0 mrem = 1 mSv = 0.001000 Sv = 1.000 mSv = 1000 μSv
0.0010 rem = 1.0 mrem = 1 mrem = 0.000010 Sv = 0.010 mSv = 10 μSv
0.0001 rem = 0.1 mrem = 1 μSv = 0.000001 Sv = 0.001 mSv = 1 μSv

See also

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Explanatory notes

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References

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Revisions and contributorsEdit on WikipediaRead on Wikipedia

Sievert

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The sievert (symbol: Sv) is the for measuring the and effective dose of to humans, quantifying the biological health effects rather than the mere energy absorbed by tissues. It is used in to assess risks from sources such as , occupational exposure, and environmental , where 1 Sv represents a dose that could produce significant effects like increased cancer risk. Unlike the gray (Gy), which measures as energy per unit mass (1 Gy = 1 J/kg), the sievert accounts for the varying biological damage caused by different radiation types, such as alpha particles versus gamma rays. The dose equivalent in sieverts is calculated as the in grays multiplied by a dimensionless weighting factor (formerly called the quality factor), which ranges from 1 for photons and electrons to 20 for alpha particles. For effective dose, an additional step weights the by tissue sensitivity factors to estimate whole-body risk. The sievert replaced the older rem unit in the international system in , with 1 Sv equivalent to 100 rem, facilitating global standardization in . The unit is named after Swedish medical physicist Rolf Maximilian Sievert (1896–1966), a pioneer in radiation protection who developed early measurement techniques and served as chairman of the International X-ray and Radium Protection Committee from 1928. Sievert's work on phantom dosimetry and exposure limits laid foundational principles for modern standards, influencing organizations like the International Commission on Radiological Protection (ICRP), where he was president from 1956 to 1962. Typical everyday exposures are far below 1 Sv, with natural background radiation averaging about 3 millisieverts (mSv) annually and a single chest X-ray around 0.2 mSv. Regulatory limits, such as 1 mSv per year for the public and 20 mSv annually for radiation workers, are set in sieverts to ensure safety.

Definition

Formal Definitions

The sievert (symbol: Sv) is the special name for the of dose equivalent, defined as equal to one joule per kilogram (1 Sv = 1 J kg⁻¹). This unit incorporates a dimensionless quality factor to account for the varying biological effectiveness of different types of relative to . The International Committee for Weights and Measures (CIPM) clarified in 2002 that the dose equivalent HH is given by H=Q×DH = Q \times D, where DD is the in gray (Gy) and QQ is the quality factor determined by the of the radiation, ensuring the sievert distinguishes biological risk from mere energy deposition. The (ICRP) defines the sievert as the special name for the SI unit of , effective dose, and operational dose quantities, each expressed in joules per (J kg⁻¹). This emphasizes the sievert's in radiological by integrating type (via radiation weighting factors) and tissue sensitivity (via tissue weighting factors) to estimate stochastic health risks, as outlined in ICRP Publication 60 (1990) and reaffirmed without substantive changes in Publication 103 (2007). As of 2025, the ICRP's 2007 recommendations remain the current standard, with no major revisions to the sievert's foundational . The sievert has its origins in mid-1970s efforts by the International Commission on Radiation Units and Measurements (ICRU) to adopt SI units for quantities. It was formally introduced by the ICRP in 1977 (Publication 26) to unify dose concepts in the SI system, replacing earlier units like the rem and providing a coherent measure for dose equivalent that factors in biological effects. This was recognized by the 16th General Conference on Weights and Measures (CGPM) in 1979 via Resolution 5, establishing it as an SI unit specifically for purposes.

Relation to Gray

The sievert (Sv) builds upon the gray (Gy), the International System of Units (SI) base unit for absorbed dose, which measures the amount of energy deposited by in a material. The gray is defined as an absorbed dose of 1 joule of energy per kilogram of mass, or 1 Gy = 1 J/kg. This physical quantity, denoted as D, provides a measure of energy deposition without regard to the type or biological effects of the . To incorporate the differing biological impacts of various radiation types, the dose equivalent H is calculated by multiplying the absorbed dose D in grays by a quality factor Q—a legacy term from earlier dosimetry systems—or, in modern practice, by the radiation weighting factor w_R as recommended by the International Commission on Radiological Protection (ICRP). Thus, the sievert serves as the unit for dose equivalent, where 1 Sv = 1 Gy × w_R (or Q), enabling the assessment of stochastic health risks from ionizing radiation. This relation distinguishes the sievert from the gray by adjusting for the of particles: photons and electrons have w_R = 1, while heavier particles like alpha particles have higher values, such as w_R = , reflecting their greater potential for cellular damage per unit energy absorbed. For instance, an of 1 Gy from alpha particles equates to an of Sv, highlighting how the sievert facilitates comparisons of biological harm across types.

Unit Symbol and Prefixes

The sievert is represented by the symbol Sv, consisting of a capital "S" followed by a lowercase "v" with no period, except when the symbol concludes a sentence. This notation was formally adopted by the 16th General Conference on Weights and Measures (CGPM) in 1979 as the special name for the SI unit of dose equivalent in radioprotection. The symbol is never abbreviated as "sie," adhering to standard SI conventions that prohibit informal shortenings of unit names. SI prefixes are applied to the sievert for practical scaling in measurements, particularly in low-dose scenarios common to environmental and occupational monitoring. The most frequently used prefixes include the millisievert (mSv) and microsievert (μSv), with conversion factors as follows:
PrefixSymbolFactorConversion to Sv
Milli-mSv10−310^{-3}1 mSv = 10−310^{-3} Sv
Micro-μSv10−610^{-6}1 μSv = 10−610^{-6} Sv
These prefixes form inseparable symbols without spaces (e.g., mSv, μSv), as specified in the SI Brochure. Larger prefixes like kilisievert (kSv) are rare due to the high doses they imply, which exceed typical regulatory limits. The Bureau International des Poids et Mesures (BIPM) and the (IAEA) provide guidelines for sievert usage in scientific reports, labels, and safety documentation to ensure clarity and consistency. According to BIPM, a space must separate the numerical value from symbol (e.g., 2.5 mSv), unit symbols are printed in upright type without modification for plurals, and no period follows the symbol internally. IAEA publications emphasize SI compliance, recommending the sievert over legacy units like the rem (with 1 Sv = 100 rem noted for conversions) and using en dashes for dose ranges (e.g., 1–5 mSv). A common pitfall in notation arises from potential confusion between millisievert (mSv) and the velocity unit meters per second (m/s), though this is mitigated by the distinct contextual use in versus .

Dose Quantities

Physical Quantities

The physical quantities in provide the foundational measures of energy transfer and deposition from to matter, serving as the basis for deriving biologically weighted quantities like the sievert. These include , , and fluence, which quantify interactions without incorporating radiation type or tissue sensitivity factors. Kerma, or kinetic energy released per unit , represents the initial transfer of kinetic energy from indirectly (such as photons or neutrons) to directly ionizing charged particles (like electrons) in a . It is defined as the of the sum of the initial kinetic energies of all charged particles liberated by uncharged particles in a small element divided by that :
K=dEtrdmK = \frac{dE_\text{tr}}{dm}
where dEtrdE_\text{tr} is the transferred energy and dmdm is the of the volume element. For monoenergetic photons, kerma relates to energy fluence Ψ\Psi (product of particle fluence and photon ) via the mass energy transfer coefficient μtr/ρ\mu_\text{tr}/\rho:
K=Ψ(μtrρ).K = \Psi \left( \frac{\mu_\text{tr}}{\rho} \right).
This quantity is particularly useful for describing energy deposition at the onset of interactions, before charged particles lose through subsequent collisions. Absorbed dose measures the actual imparted to matter by after interactions, defined as the mean deposited per unit :
D=dεˉdmD = \frac{d\bar{\varepsilon}}{dm}
where dεˉd\bar{\varepsilon} is the average transferred to the dmdm. Under conditions of equilibrium—where the number of s entering a volume equals those leaving— approximates collision (kerma excluding radiative losses): D≈KcolD \approx K_\text{col}. can be specified as a point value, representing the local deposition at a specific location, or as an organ-averaged value, which integrates the dose over the or volume of a tissue or organ to assess overall exposure: DT,RD_{T,R}, the in tissue TT from type RR, averaged over the organ volume. Point doses highlight localized effects, such as in radiotherapy hotspots, while organ-averaged doses provide a mean for broader evaluation. The unit for both kerma and is the gray (Gy), equivalent to 1 joule per (J/kg). Fluence quantifies the incident radiation field as the number of particles passing through a unit area, typically an infinitesimal : Φ=dNda\Phi = \frac{dN}{da}, where dNdN is the number of particles and dada is the cross-sectional area (unit: m⁻²). Energy fluence Ψ=Φ⋅E\Psi = \Phi \cdot E (with EE as average particle ) links directly to dose quantities; for example, absorbed dose in a medium relates to energy fluence via the mass energy absorption μen/ρ\mu_\text{en}/\rho: D=Ψ(μenρ)D = \Psi \left( \frac{\mu_\text{en}}{\rho} \right). This connection allows fluence measurements to estimate dose deposition, especially in uniform fields, though actual dose varies with material properties and geometry. These physical quantities in grays underpin sievert calculations by providing the unweighted metrics that are later modified for biological effectiveness.

Operational Quantities

Operational quantities in are defined by the International Commission on Units and Measurements (ICRU) as practical, measurable proxies for the protection quantities established by the (ICRP), enabling assessments of external through and calculations. These quantities, expressed in sieverts (Sv), approximate the biological effects of radiation by incorporating quality factors or radiation weighting factors into measurements at specified depths in idealized phantoms, without requiring full anatomical modeling of human tissues. Their primary role is to support the of dosimeters and survey meters, ensuring that instrument readings provide conservative estimates of potential health risks in diverse radiation fields. Central to the definition of many operational quantities is the ICRU sphere, a standardized phantom consisting of a 30 constructed from tissue-equivalent material with a of 1 g/³ and elemental composition of 76.2% oxygen, 11.1% carbon, 10.1% , and 2.6% . This simulates for area monitoring purposes, where dose equivalents are computed at depths such as 10 for deeper-penetrating (relevant to ambient dose equivalents) or shallower depths like 0.07 mm for exposure in directional fields. By expanding and aligning fields within or around this phantom, the quantities account for scattered , providing a basis for environmental and workplace assessments that correlate reasonably with protection quantities like effective dose. Conversion coefficients link measurable physical quantities, such as particle fluence (particles per unit area) or air for photons, to operational dose equivalents, allowing estimation across various radiation types including photons, , electrons, protons, and heavier ions. These coefficients, calculated via simulations of radiation transport in the ICRU sphere or updated phantoms, vary with energy and field geometry; for example, neutron coefficients incorporate fluence-to-dose conversions that peak around 1 MeV due to tissue interactions. Tabulated in seminal reports like ICRU Report 57 (1998) and extensively revised in the joint ICRU/ICRP Report 95 (2020), they enable instruments to display readings directly in sieverts for photons from diagnostic X-rays (e.g., coefficients around 1.2 pSv for 100 keV) to high-energy (up to 10 pSv at energies). This approach ensures practical application in without exhaustive biological computations, prioritizing overestimation for safety.

Protection Quantities

Protection quantities in radiological protection are sievert-based measures designed to estimate the stochastic health risks, such as cancer induction and heritable effects, from ionizing radiation exposure to humans. These quantities account for the varying biological effectiveness of different radiation types and the differing sensitivities of body tissues, providing a framework for assessing overall risk rather than physical energy deposition alone. Unlike absorbed dose, which is a fundamental physical quantity in grays, protection quantities incorporate weighting factors to better represent health detriments. The , denoted HTH_T, to a specific tissue or organ T is calculated as the sum over all types R of the product of the radiation weighting factor wRw_R and the mean DT,RD_{T,R} in that tissue: HT=∑RwRDT,RH_T = \sum_R w_R D_{T,R} This quantity expresses the dose in sieverts (Sv) and adjusts for the of the on effects in the targeted tissue, enabling organ-specific evaluation. For instance, it is used to assess potential harm to radiosensitive organs like the from mixed fields. The unit of equivalent dose is the sievert, the same as for effective dose, emphasizing its role in protection contexts. The effective dose, denoted EE, extends this by providing a whole-body metric through the tissue-weighted sum of equivalent doses across all specified organs and tissues: E=∑TwTHTE = \sum_T w_T H_T Here, wTw_T represents the tissue weighting factor, which reflects the relative contribution of each tissue to total . Expressed in sieverts, effective dose allows comparison of risks from uniform or non-uniform exposures, equating them to the stochastic detriment from a whole-body uniform exposure of the same magnitude. This makes it particularly valuable for scenarios involving partial-body , where direct whole-body would underestimate or misrepresent the health impact. A key distinction between organ equivalent dose and effective dose lies in their scope: HTH_T focuses on the risk to individual tissues or organs, useful for targeted assessments like deterministic effects thresholds, whereas EE integrates these into a single value representing the total body risk, facilitating broad protection strategies. In practice, effective dose serves as the primary quantity for regulatory limits and , such as annual limits of 20 mSv for radiation workers and 1 mSv for the public, ensuring compliance and optimization in planned exposure situations like medical diagnostics or occupational settings. These applications, as outlined in ICRP Publication 103, support prospective dose planning and verification against international standards.

Calculation of Protection Quantities

Radiation Weighting Factor

The radiation weighting factor, denoted as wRw_R, is a dimensionless multiplier applied to the absorbed dose from a specific radiation type to derive the equivalent dose, accounting for the relative biological effectiveness (RBE) of different ionizing radiations in inducing stochastic health effects. It adjusts the physical absorbed dose, measured in grays (Gy), to reflect variations in biological damage potential due to differences in linear energy transfer (LET), where high-LET radiations like alpha particles cause denser ionization tracks and greater cellular harm compared to low-LET radiations such as photons. The rationale for wRw_R centers on RBE values derived from radiobiological studies, emphasizing endpoints like cancer induction and hereditary effects at low doses, rather than deterministic effects. These factors are established through a combination of and data, epidemiological observations, and biophysical modeling, averaged over human tissues to provide a conservative estimate suitable for radiological protection. In the equivalent dose calculation for protection quantities, wRw_R scales the to yield results in sieverts (Sv). The (ICRP) Publication 103 specifies fixed wRw_R values for most radiation types, with neutrons requiring energy-dependent adjustment. These values represent refinements from prior recommendations, incorporating updated RBE data without altering the core framework for photons, electrons, or heavy ions.
Radiation TypewRw_R Value
Photons, all energies1
Electrons and muons, all energies1
Protons and charged pions, >2 MeV2
Alpha particles, fission fragments, and heavy ions20
For neutrons, wRw_R is defined as a continuous function of incident neutron energy EnE_n (in MeV) to better capture spectral variations in biological effectiveness, replacing the discrete steps of earlier guidelines. The function is piecewise: If En<1:wR=2.5+18.2exp⁡(−(ln⁡En)26)If 1≤En≤50:wR=5.0+17.0exp⁡(−(ln⁡(2En))26)If En>50:wR=2.5+3.25exp⁡(−(ln⁡(0.04En))26)\begin{align*} & \text{If } E_n < 1: \quad w_R = 2.5 + 18.2 \exp\left( -\frac{(\ln E_n)^2}{6} \right) \\ & \text{If } 1 \leq E_n \leq 50: \quad w_R = 5.0 + 17.0 \exp\left( -\frac{(\ln (2 E_n))^2}{6} \right) \\ & \text{If } E_n > 50: \quad w_R = 2.5 + 3.25 \exp\left( -\frac{(\ln (0.04 E_n))^2}{6} \right) \end{align*} This formulation peaks at approximately 21 near 1 MeV, reflecting enhanced damage from nuclear recoils and secondary particles, and approaches 2.5 at very low (<10 keV) or very high (>1 GeV) energies. As of 2025, no revisions to these wRw_R values have been adopted by the ICRP, though a system-wide review is underway for future recommendations.

Tissue Weighting Factor

The tissue weighting factor, denoted as wTw_T, represents the fraction of the total stochastic detriment (primarily cancer induction and heritable effects) attributable to the irradiation of a specific tissue or organ TT, assuming uniform whole-body exposure. These factors are dimensionless and sum to 1 across all tissues, enabling the calculation of effective dose by weighting the equivalent dose to each organ according to its relative radiosensitivity. In the 2007 recommendations (ICRP Publication 103), the tissue weighting factors were revised based on updated epidemiological data from atomic bomb survivors and other cohorts, emphasizing sex-averaged values derived from reference male and female computational phantoms. The values are applied uniformly for both sexes in general protection scenarios, though sex-specific factors can be used for targeted assessments; no major revisions to these factors have occurred since 2007. Key examples include bone marrow (red blood cells) at 0.12, lungs at 0.12, and the remainder tissues (a group of 13 organs including adrenals, extrathoracic region, gall bladder, heart, kidneys, lymphatic nodes, muscle, oral mucosa, pancreas, prostate, small intestine, spleen, thymus, and uterus/cervix) at 0.12 collectively.
Tissue or Organ GroupTissue Weighting Factor wTw_T
Bone marrow (red), colon, lung, stomach, breast0.12 each
Remainder tissues (13 specified organs)0.12 (total)
Gonads0.08
Bladder, oesophagus, liver, thyroid0.04 each
Bone surface, brain, salivary glands, skin0.01 each
These factors integrate with radiation weighting factors wRw_R from the previous subsection to form equivalent doses for each tissue, which are then summed to yield the effective dose. Compared to the 1990 recommendations (ICRP Publication 60), notable changes include an increase in the factor from 0.05 to 0.12, reflecting higher estimated cancer risks, and a decrease in the gonads factor from 0.20 to 0.08 due to revised heritable effects estimates. The remainder tissues' collective weight also rose from 0.05 (for 12 organs) to 0.12 (for 13 organs), better accounting for distributed risks.

Effective Dose Formula

The effective dose EE, a protection quantity used to quantify stochastic radiation risks to the whole body, is computed as the double summation over specified tissues TT and radiation types RR: E=∑TwT∑RwRDT,RE = \sum_T w_T \sum_R w_R D_{T,R}, where wTw_T is the tissue weighting factor, wRw_R is the radiation weighting factor, and DT,RD_{T,R} is the to tissue TT from radiation RR. This formula integrates the of different radiations and the varying sensitivities of body tissues to produce a single risk-related value in sieverts (Sv). The derivation proceeds in steps from fundamental physical quantities. First, the DT,RD_{T,R}, measured in grays (Gy) as energy deposited per unit mass, quantifies energy absorption but does not account for type differences. Second, the HTH_T to tissue TT adjusts for biological impact by applying wRw_R: HT=∑RwRDT,RH_T = \sum_R w_R D_{T,R}, expressed in Sv. Third, the effective dose EE then weights these equivalent doses by wTw_T to reflect overall detriment: E=∑TwTHTE = \sum_T w_T H_T, yielding the composite formula above. These steps enable comparison of diverse exposures on a common scale for radiological protection. This framework rests on key assumptions, including the linear no-threshold (LNT) model, which posits that effects like cancer induction are proportional to dose across all levels without a safe threshold, allowing summation for mixed exposures. Additionally, calculations average over a reference population, typically sex-averaged values, to represent rather than individual-specific doses. For , consider a hypothetical uniform external exposure delivering 0.10 Gy from gamma rays (photons, wR=1w_R = 1) to the lungs (wT=0.12w_T = 0.12) and 0.05 Gy from alpha particles (wR=20w_R = 20) to red (wT=0.12w_T = 0.12), with negligible doses elsewhere. The to lungs is Hlungs=1×0.10=0.10H_{\text{lungs}} = 1 \times 0.10 = 0.10 Sv, and to marrow is Hmarrow=20×0.05=1.00H_{\text{marrow}} = 20 \times 0.05 = 1.00 Sv. The effective dose is then E=(0.12×0.10)+(0.12×1.00)=0.132E = (0.12 \times 0.10) + (0.12 \times 1.00) = 0.132 Sv, demonstrating how high-LET radiation amplifies overall despite lower absorbed dose. This example uses wRw_R and wTw_T values from established standards but simplifies by ignoring other tissues and sex-averaging.

External Dose Measurement

Ambient Dose Equivalent

The ambient dose equivalent, denoted as H∗(10)H^*(10), is an operational quantity defined as the dose equivalent at a depth of 10 mm in the ICRU sphere resulting from the corresponding expanded and aligned field at a specified point in the actual field. The ICRU sphere is a 30 cm diameter sphere composed of tissue-equivalent material with 1 g/cm³ and elemental composition approximating . This quantity is specifically intended for strongly penetrating and serves as a conservative estimate of the effective dose for external whole-body exposures, particularly from photons, where it approximates the protection quantity by accounting for depth dose in a simplified phantom. In practical applications, H∗(10)H^*(10) is widely used for area monitoring in radiation-controlled workplaces, such as nuclear facilities and environments, to assess potential exposure risks to personnel. It also forms the basis for survey meters and other area dosimeters, ensuring instruments respond appropriately to ambient fields by relating their readings to established conversion coefficients. For instance, factors for survey monitors are determined as NH∗=H∗(10)/MN_{H^*} = H^*(10) / M, where MM is the instrument reading, facilitating accurate environmental dose assessments. The energy response of H∗(10)H^*(10) is engineered for near-uniformity across relevant spectra: for photons, conversion coefficients from air to H∗(10)H^*(10) remain approximately flat, with a close to 1.20 from 20 keV to 10 MeV, enabling reliable measurements without significant energy dependence in this range. For s, fluence-to-H∗(10)H^*(10) conversion coefficients h∗(10)h^*(10) vary with energy, increasing from low values below 1 keV to a peak around 1 MeV (reaching about 80 pSv·cm² at 1 MeV) before decreasing at higher energies, reflecting the quality factor's modulation by neutron interaction characteristics. Similar overestimations occur for high-energy protons and muons; updated coefficients in ICRU Report 95 address energies up to 10 GeV for better accuracy in such fields. Despite its utility, H∗(10)H^*(10) has limitations, particularly overestimating the effective dose for neutrons in high-energy ranges above 10 MeV or in fields dominated by high-energy charged particles like protons or muons, where the operational definition based on expanded fields does not fully capture anisotropic or secondary particle contributions. This can lead to conservative but potentially excessive assessments in accelerator or environments.

Directional Dose Equivalent

The directional dose equivalent, denoted H′(d,α)H'(d, \alpha), is an operational quantity in that quantifies the dose equivalent at a specified depth dd in tissue along a given direction of incidence α\alpha. It is defined as the dose equivalent produced at a point within the ICRU (a 30 cm filled with tissue-equivalent material of 1 g/cm³) by the corresponding expanded and aligned field from the actual anisotropic field. The unit is the sievert (Sv). Common depths include d=0.07d = 0.07 mm for shallow dose assessment, corresponding to H′(0.07,α)H'(0.07, \alpha), and d=10d = 10 mm for deep dose, corresponding to H′(10,α)H'(10, \alpha). The shallow version approximates the equivalent dose to the skin or lens of the eye in oriented fields, while the deep version serves as a conservative surrogate for the effective dose from external exposure. This directionality distinguishes it from isotropic quantities like the ambient dose equivalent, making it suitable for scenarios with known radiation direction. The quantity is applied in monitoring anisotropic external radiation fields, such as scattered from accelerators or nuclear facilities, where the incident direction can be specified. Conversion coefficients from fluence to H′(d,α)H'(d, \alpha) are provided in ICRP Publication 116 for various radiation types and energies to facilitate practical measurements. For calibration in directional fields, the ICRU sphere is used; the slab phantom is employed for personal dose equivalents.

Personal Dose Equivalent

The personal dose equivalent, denoted as Hp(d)H_p(d), is an operational quantity defined by the (ICRP) as the dose equivalent in ICRU four-element at a depth dd below a specified point on the , calculated using a slab phantom that simulates the human trunk. This phantom is a rectangular prism measuring 30 cm × 30 cm × 15 cm, composed of ICRU tissue with a of 1 g cm⁻³, to account for the and scatter from the body during external . The most commonly used variants are Hp(10)H_p(10), which estimates the dose to tissues at a 10 mm depth for penetrating radiation such as photons and neutrons, and Hp(0.07)H_p(0.07), which measures the dose at a 0.07 mm depth for superficial effects like skin dose from beta particles or low-energy photons. A key feature of the personal dose equivalent is its inclusion of the backscatter factor, which represents the increase in dose due to radiation reflected from the body surface back toward the dosimeter. For photon radiation in the energy range above 100 keV, this factor typically increases the measured dose by approximately 30% compared to measurements in free air, as the body's tissues reflect a portion of the incident radiation, enhancing the local dose at the point of measurement. This correction is embedded in the conversion coefficients provided by ICRP Publication 74, ensuring that Hp(d)H_p(d) more accurately reflects the dose to the wearer than field quantities alone. The personal dose equivalent thus builds on the directional dose equivalent by incorporating body scatter effects for individual monitoring scenarios. In practice, the personal dose equivalent is primarily applied in personal dosimetry systems, such as thermoluminescent dosimeter (TLD) badges or optically stimulated luminescence (OSL) dosimeters worn by radiation workers to track cumulative exposure. These devices are calibrated to read Hp(10)H_p(10) and Hp(0.07)H_p(0.07), enabling the estimation of individual doses in occupational settings like nuclear facilities, medical radiology, and . Annual monitoring is standard for workers likely to exceed 10% of regulatory dose limits, with badges exchanged periodically to integrate exposure over time and ensure compliance with protection standards. For relating personal dose equivalent to protection quantities, ICRP provides approximate conversion factors from Hp(10)H_p(10) to effective dose, particularly for external photon exposures where Hp(10)H_p(10) serves as a conservative overestimate of effective dose in anterior-posterior geometries. These factors, derived from simulations in ICRP Publication 116, vary by radiation type and energy but generally show Hp(10)H_p(10) approximating effective dose within 20-30% for broad-beam fields, allowing dosimetric readings to inform risk assessments without full phantom calculations. For neutrons and other particles, specific coefficients adjust Hp(10)H_p(10) to better align with organ-equivalent doses.

Instrumentation Response

Instrumentation in is calibrated to operational quantities such as the ambient dose equivalent H∗(10)H^*(10) and personal dose equivalent Hp(10)H_p(10), expressed in sieverts (Sv), to approximate protection quantities for monitoring purposes. typically employs standard sources like cesium-137 (Cs-137) for photons and americium-beryllium (Am-Be) for s, ensuring to national or international standards such as those defined in ISO 4037. For instance, Cs-137 sources, emitting gamma rays at 662 keV, are used to irradiate instruments in controlled fields, with the reference dose determined via air measurements converted to H∗(10)H^*(10) using established coefficients (e.g., H∗(10)/Ka=1.20H^*(10)/K_a = 1.20 Sv·Gy⁻¹). Am-Be sources provide a with a mean of about 4.4 MeV, calibrated similarly for neutron fields to match Hp(10)H_p(10) on phantoms like the ICRU slab. Common types of detectors include ionization chambers, thermoluminescent dosimeters (TLDs), and optically stimulated luminescence (OSL) dosimeters. Ionization chambers, often used in survey meters, directly measure ionization current proportional to absorbed dose, suitable for real-time monitoring of H∗(10)H^*(10). TLDs, typically based on lithium fluoride (LiF), accumulate dose over time and are read via thermoluminescence, covering ranges from 0.1 mSv to 10 Sv for Hp(10)H_p(10). OSL dosimeters, using aluminum oxide (Al₂O₃:C), offer similar ranges (10 µSv to 10 Sv) with optical readout, providing advantages in reusability and lower detection limits. These devices are calibrated on phantoms (e.g., PMMA slabs for Hp(10)H_p(10)), where backscatter effects are included via the phantom setup, ensuring response to the defined depths of 10 mm. Response functions of these instruments account for energy and angular dependencies to approximate sievert-based quantities accurately. Energy dependence is critical; for example, electronic personal dosimeters (EPDs) must maintain response within ±20% over 30 keV to 1.3 MeV for photons, while OSL dosimeters exhibit flat response from 5 keV to 40 MeV. Angular response for survey meters is evaluated up to ±60° or ±80° from normal incidence, ensuring isotropic behavior in varied fields, as per ISO standards. Conversion from raw detector signals (e.g., counts or charge) to Sv involves applying calibration factors, such as H=h⋅N⋅MH = h \cdot N \cdot M, where hh is the conversion coefficient, NN the reading, and MM any corrections for environmental factors. Uncertainties in these measurements arise from factors like energy spectrum variations, scatter, and field non-uniformity, with typical values of ±20-30% for operational quantities under conditions using survey meters. For personal dosimeters, uncertainties are around ±10% at 95% confidence, but can reach ±100% in workplace scenarios due to unknown field characteristics. These uncertainties highlight the approximate nature of operational quantities, emphasizing the need for regular and performance testing.

Recent Developments

In 2024, ICRU Report 95 proposed revisions to operational quantities for external to improve alignment with quantities. Key changes include redefining H*(10), H'(d, α), and H_p(d) as products of air or fluence with appropriate factors at a point in air or on a phantom surface, using an updated ICRU computational phantom, and providing conversion coefficients for particles up to 10 GeV. These updates address limitations in high-energy fields and are under consideration by ICRP for adoption in radiological standards as of 2025.

Internal Dose Assessment

Committed Effective Dose

The committed effective dose quantifies the total effective dose resulting from the incorporation of radionuclides into the body, projected over a specified integration period following intake. It represents the sum of the products of the committed equivalent doses to specified tissues or organs, HT(τ)H_T(\tau), and their respective tissue weighting factors, wTw_T, such that E(τ)=∑TwTHT(τ)E(\tau) = \sum_T w_T H_T(\tau). This integration time τ\tau is 50 years for adults and extends to age 70 for children, capturing the long-term risk from internal emitters. Intake of radionuclides occurs primarily through of aerosols or gases, of contaminated or , and to a lesser extent, absorption through the skin or wounds, with the activity intake denoted as II in becquerels (Bq). The committed effective dose is derived by applying biokinetic models to model radionuclide uptake, distribution, retention, and excretion in reference individuals, as established by the (ICRP). These models account for physiological processes specific to each and exposure route. Recent updates in the ICRP Environmental Intakes of Radionuclides series (e.g., Publication 158, 2024) provide revised age-specific coefficients aligned with updated biokinetics and tissue weighting factors from Publication 103 (2007). Dose coefficients, denoted hTh_T for committed equivalent dose to tissue TT per unit intake or e(50)e(50) for committed effective dose per unit intake, are computed from these biokinetic and dosimetric data. For instance, the committed effective dose coefficient for ingestion of by an adult member of the public is approximately 1.6×10−81.6 \times 10^{-8} Sv/Bq (as of 2024), predominantly due to uptake in the thyroid gland. These coefficients enable straightforward calculation of E(τ)=e(50)×IE(\tau) = e(50) \times I, facilitating assessments in occupational and public exposure scenarios. Distinctions between acute and chronic intakes influence assessment but not the core definition of committed effective dose, which applies to each identifiable intake event. For acute intakes, a single E(τ)E(\tau) is computed based on the instantaneous activity incorporated. In chronic exposure scenarios, involving repeated or continuous intakes, the total committed effective dose is the sum of individual E(τ)E(\tau) values for each intake over the relevant period, often using time-integrated intake rates.

Integration Over Time

In internal dosimetry, the effective dose rate Ė(t) represents the time-dependent to the whole body following the of radionuclides, arising from their within organs and tissues as influenced by biokinetic processes such as uptake, translocation, and . This rate varies over time due to the combined effects of physical decay (characterized by radionuclide-specific half-lives) and biological elimination, which determine the amount of activity present in target tissues at any moment post-. The effective dose, which quantifies the total internal dose attributable to a single , is obtained by integrating the effective over a specified period following exposure. For adults, this integration extends from the time of intake to 50 years later, effectively capturing the long-term dose accumulation while truncating at to ensure practicality; for children, it extends to age 70 years to account for longer remaining lifespan. Mathematically, this is expressed as: E(τ)=∫0τE˙(t) dtE(\tau) = \int_0^\tau \dot{E}(t) \, dt where τ\tau is the integration period (50 years for adults), and E˙(t)\dot{E}(t) incorporates tissue-specific contributions weighted by and tissue weighting factors. To compute these quantities, organ retention functions fT(t)f_T(t) describe the of the systemic activity retained in tissue TT at time tt after entry into the , typically modeled as a sum of exponential terms to reflect multi-compartmental biokinetics: fT(t)=∑iaie−λitf_T(t) = \sum_i a_i e^{-\lambda_i t} Here, aia_i are fractional coefficients summing to 1, and λi=λr,i+λb,i\lambda_i = \lambda_{r,i} + \lambda_{b,i} combines the physical decay constant λr\lambda_r with biological removal rates λb\lambda_b for each compartment ii. These functions enable the derivation of time-integrated activity and subsequent dose coefficients used in practice. Updated biokinetic parameters in recent ICRP publications (e.g., Occupational Intakes series, 2016–2017) refine these retention functions for accuracy. The nature of the isotope significantly affects the integration outcome. For short-lived radionuclides, such as (physical half-life of 8 days), the dose rate peaks rapidly post-intake and decays quickly, with nearly all committed dose delivered within weeks due to swift physical decay dominating over biological retention. In contrast, for long-lived isotopes like (physical half-life of 30 years), the dose accumulates gradually over decades, as the integration period captures a substantial portion of the physical decay while biological retention—modeled with components of about 0.25 days and 70 days half-life—prolongs systemic exposure beyond the physical half-life alone. This distinction underscores the importance of the 50-year truncation, which conservatively includes most relevant dose for such nuclides without extending indefinitely.

Biokinetic Models

Biokinetic models in describe the uptake, distribution, retention, and excretion of within the following internal intake, enabling the estimation of time-integrated dose to organs and tissues. These models are physiological representations that account for biological processes such as absorption from entry sites, transport via blood, and accumulation in target organs. Developed primarily by the (ICRP), they form the basis for calculating committed internal doses, integrating radionuclide behavior over periods like 50 years for adults or until age 70 for children. The Occupational Intakes of Radionuclides series (Publications 130–137, 2016–2017) and Environmental Intakes series (e.g., Publication 158, 2024) provide updated models and coefficients. The ICRP Human Respiratory Tract Model (HRTM), introduced in Publication 66, specifically addresses as a primary route by modeling particle deposition, , and absorption into blood across respiratory regions. The tract is divided into the extrathoracic region (ET), comprising ET1 (anterior nasal passages and mouth) and ET2 (posterior nasal passages, , and ), the bronchial region (BB: bronchi), bronchiolar region (bb: terminal bronchioles), and alveolar-interstitial region (AI: alveoli and associated interstitium). Deposition efficiency varies with particle aerodynamic diameter (typically 0.001–20 μm): particles larger than 5 μm predominantly deposit in ET1 and BB via inertial impaction, with up to 50% of ET1 deposits cleared directly to the environment; particles of 1–5 μm settle in BB and bb through sedimentation and impaction, with rapid clearance (e.g., 2 hours from BB); and ultrafine particles below 1 μm favor AI deposition via , where retention can extend to years in slow-cleared compartments (AI2: ~2 years, AI3: ~20 years). This size-dependent deposition ensures accurate prediction of initial burdens for aerosols with activity median aerodynamic diameters of 1 μm (environmental) or 5 μm (occupational). Once absorbed into the systemic circulation, radionuclide behavior is governed by element-specific biokinetic models that quantify transfer rates between (as the central compartment) and organs such as liver, kidneys, , and . These models use fractional transfer coefficients (e.g., in day⁻¹) to represent uptake from to tissues and back to plasma, tailored to chemical form and . For instance, ICRP Publication 128 compiles such models for key elements in , including rapid uptake of into the (transfer coefficient ~0.3 from ) and strontium-89 retention in via surface-seeking mechanisms. Gastrointestinal absorption models, like those in Publication 100, further specify fractional uptake (f₁ values) ranging from 0.001 for to 1 for cesium, influencing systemic entry from . ICRP biokinetic models incorporate age- and sex-dependent parameters to reflect physiological variations, particularly higher uptake and retention in vulnerable populations. Children exhibit elevated gastrointestinal absorption for elements like (f₁ up to 0.3 vs. 0.15 in adults) and faster turnover, leading to greater skeletal doses; for example, lead models in Publication 72 show 30–50% higher blood retention in infants due to immature barriers. Sex differences arise from variances in organ masses and hormonal influences, such as lower iron absorption in adult males compared to females, as detailed in Publication 89's . These adjustments ensure dose coefficients scale appropriately, with pediatric models often derived from adult baselines scaled by body weight and maturity. Software tools implement these ICRP models to automate committed dose computations from data or intake scenarios. IMBA (Integrated Modules for Analysis) supports user-defined parameters for HRTM and systemic kinetics, calculating organ-specific committed effective doses for over 800 radionuclides while allowing customization of transfer coefficients. Similarly, MONDAL (Monitoring to Dose cALculation support system), developed by Japan's National Institute of Radiological Sciences (now QST), integrates biokinetic simulations for intake assessment, generating retention functions and dose coefficients aligned with ICRP recommendations, particularly for occupational monitoring. Both tools facilitate integration of biokinetic outputs with time-dependent exposure data to derive total internal doses and are compatible with updated ICRP data as of 2024.

Health Effects and Limits

Stochastic Effects

Stochastic effects refer to radiation-induced health outcomes, such as cancer and hereditary disorders, where the probability of occurrence is proportional to the absorbed dose in sieverts, but the severity remains independent of dose level. These effects are characterized by their random nature and lack of a dose threshold, meaning even small exposures carry some risk of manifestation years or decades later. The effective dose, expressed in sieverts, serves as the primary quantity for quantifying and comparing these probabilistic risks across different exposure scenarios. The linear no-threshold (LNT) model underpins for stochastic effects, positing a straight-line relationship between dose and risk probability without a safe threshold. Endorsed in the BEIR VII report, this model extrapolates from high-dose observations to predict low-dose risks, estimating an approximate 5% increase in lifetime fatal cancer risk per sievert of low-linear energy transfer (low-LET) for the general population. This extrapolation assumes risks scale linearly, with adjustments for factors like age, sex, and exposure type, though uncertainties increase at doses below 100 millisieverts. Among sensitive endpoints, exhibits elevated susceptibility, with epidemiological models showing risks detectable around 100 millisieverts, aligning with LNT predictions despite statistical challenges at lower doses. Hereditary effects, involving transgenerational genetic mutations, carry an estimated risk of approximately 0.6% per sievert, though direct human evidence remains limited and primarily inferred from animal data and doubling dose concepts. The epidemiological foundation for these models derives mainly from the Life Span Study of over 120,000 atomic bomb survivors in and , which has tracked excess cancers proportional to dose over decades. Supporting data come from cohorts exposed via medical procedures, such as diagnostic imaging and radiotherapy, confirming stochastic patterns in populations receiving 10-500 millisieverts. These studies collectively validate the LNT framework for sievert-based risk estimation.

Deterministic Effects

Deterministic effects, also referred to as tissue reactions, are radiation-induced injuries to normal tissues and organs that exhibit a clear threshold dose below which no observable occurs. Above this threshold, the severity of the injury increases predictably with higher absorbed doses, measured in sieverts (Sv) for equivalent dose to account for radiation type and biological effectiveness. These effects are distinct from processes because they depend on the depletion of functional cells rather than random genetic alterations, allowing for dose-dependent clinical manifestations in contexts. The underlying mechanisms of deterministic effects primarily involve cell killing through processes such as clonogenic cell death or apoptosis, leading to insufficient repopulation and subsequent tissue dysfunction. This contrasts with stochastic effects, which stem from unrepaired DNA damage causing mutations and probabilistic outcomes like cancer. For instance, in highly radiosensitive tissues, radiation depletes parenchymal cells (e.g., epithelial cells in the skin or intestinal crypts) or damages supportive structures like vascular endothelium, resulting in observable harm only when a critical fraction of cells is lost. Biological modifiers, including repair mechanisms and tissue-specific responses, can influence the expression of these effects post-exposure. Prominent examples include skin , where acute exposures of 2-6 Sv cause transient reddening starting at around 2 Sv and more pronounced reactions at 6 Sv due to vascular damage and inflammatory responses. (ARS) emerges in whole-body exposures exceeding 1 Sv, encompassing hematopoietic, gastrointestinal, and neurovascular subsyndromes with increasing lethality above 2-10 Sv from widespread cell depletion in , gut, and . Lens opacification leading to cataracts has a threshold of 0.5-2 Sv for acute doses to the eye, involving damage to epithelial cells and fiber disruption, though individual variability exists. Dose-rate plays a critical role in modulating deterministic effects, as protracted exposures allow time for sublethal damage repair and cell repopulation, thereby raising effective thresholds and reducing severity compared to acute . For example, chronic lens exposures may tolerate up to 5 Sv without cataracts, while skin and hematopoietic tissues show enhanced recovery during fractionated dosing. This sparing effect underscores the importance of exposure timing in assessing risks for protection quantities like .

Regulatory Dose Limits

The (ICRP) establishes fundamental dose limits in sieverts to safeguard workers and the public from exposure in planned situations. For occupational exposure, the effective dose limit is 20 mSv per year, averaged over 5 consecutive years, with no single year exceeding 50 mSv; for members of the public, it is 1 mSv per year. These limits encompass the total effective dose, which sums contributions from both external irradiation (e.g., measured via personal dosimeters) and internal contamination (e.g., from or , assessed using biokinetic models). In addition to effective dose, ICRP specifies separate equivalent dose limits for radiosensitive tissues to prevent deterministic effects. The equivalent dose limit to the lens of the eye is 20 mSv per year, averaged over 5 years, with no single year exceeding 50 mSv for workers, and 15 mSv per year for the public; for the skin, it is 500 mSv per year (averaged over any 1 cm² for any part of the body) for workers and 50 mSv per year for the public. These tissue-specific limits complement the effective dose by addressing localized exposures that could lead to tissue reactions. A core principle underlying these limits is the ALARA (As Low As Reasonably Achievable) optimization process, which requires keeping doses below the limits through , administrative measures, and protective equipment, while balancing economic and social factors. These limits are designed to minimize risks of effects, such as cancer, while ensuring deterministic effects are avoided. Many national and international regulations align with ICRP recommendations; for instance, the (IAEA) endorses these limits in its Basic Safety Standards (GSR Part 3, 2014), with no fundamental changes to the core framework since ICRP Publication 103 (2007), except for the reduced lens of eye limit in 2012.

Practical Examples

Common Dose Levels

The sievert (Sv) quantifies the effective dose of , providing context for health risks when compared to typical exposure levels from natural, , and accidental sources. These doses are expressed in millisieverts (mSv; 1 mSv = 0.001 Sv) for everyday scenarios and sieverts for higher acute exposures, helping to illustrate the scale relative to regulatory limits like the 1 mSv annual public exposure guideline from the . Natural , arising from cosmic rays, terrestrial sources, and internal radionuclides like , delivers a global average annual effective dose of approximately 2.4 mSv, though this varies by location due to factors such as soil composition and altitude. In regions with elevated concentrations, such as certain mining areas or geologically active zones, annual doses can reach up to 10 mSv, primarily from inhalation of radon decay products. Medical procedures contribute variably to individual doses, with a standard chest computed tomography (CT) scan delivering an effective dose of about 7 mSv, equivalent to roughly three years of natural background exposure. Routine dental X-rays, assuming 2-4 intraoral images per year, result in a negligible annual effective dose of approximately 0.01 mSv. Notable accidental exposures highlight higher dose ranges; during the 1986 Chernobyl nuclear accident, acute effective doses to initial responders and cleanup workers (liquidators) ranged from less than 0.1 Sv for most cleanup workers to over 6 Sv for some initial responders, with averages around 0.12 Sv across over 500,000 participants, leading to in cases exceeding 1 Sv. In contrast, public exposures from the 2011 Fukushima Daiichi accident were much lower, with lifetime effective doses for residents in affected prefectures estimated at less than 10 mSv, primarily from external gamma and minor internal contamination. Over a typical lifespan of 70 years, cumulative natural background exposure accumulates to about 100-200 mSv, underscoring that most individuals encounter low-level routinely without exceeding safe thresholds.
SourceTypical Effective DoseNotes
Global natural background (annual)2.4 mSvIncludes cosmic, terrestrial, and internal sources; varies by geography.
High- areas (annual)Up to 10 mSvMainly from in homes or workplaces.
Chest ~7 mSvSingle procedure; diagnostic imaging.
Annual dental X-rays~0.01 mSvRoutine checkups with 2-4 images.
Chernobyl workers (acute)<0.1 to >6 SvInitial responders and liquidators; average 0.12 Sv.
Fukushima public (lifetime)<10 mSvEvacuated and nearby residents.
Lifetime natural background100-200 mSvOver 40-80 years at average rates.

Dose Rate Comparisons

The dose rate, expressed in sieverts per unit time (typically per hour), quantifies the rate at which effective dose is delivered from sources, allowing comparisons of exposure intensity across everyday, occupational, and accidental scenarios. This metric is crucial for assessing relative risks without integrating over exposure duration. Natural , arising from cosmic rays, terrestrial sources, and , exposes individuals to an average of approximately 0.3 μSv per hour in the United States. This baseline level varies by location but provides a reference for negligible chronic exposure. In contrast, cosmic radiation during commercial at typical cruising altitudes (around 10 km) elevates the to 5–10 μSv per hour, primarily due to galactic cosmic rays and solar particles, with higher values at polar routes or during . Medical procedures, such as those in or , can produce significantly higher dose rates to the patient, reaching up to 50 mSv per hour for prolonged or complex imaging, though typical rates for standard procedures are lower, around 8–10 mSv per hour of beam-on time. Extreme dose rates occurred during the 1986 Chernobyl accident, where initial levels near the exposed reactor core were estimated at up to 300 Sv per hour, posing immediate lethal risks to unprotected personnel within minutes.
SourceTypical Dose RateContext
Natural Background~0.3 μSv/hGlobal average exposure
5–10 μSv/hCruising altitude, commercial flights
(Medical)Up to 50 mSv/hPatient during interventional procedures
Chernobyl Core (1986)Up to 300 Sv/hImmediately post-explosion

Occupational and Public Exposure

In occupational settings, workers typically receive an average annual effective dose of around 1 mSv, with the majority below 5 mSv, as reported by the International Atomic Energy Agency's Information System on Occupational Exposure (ISOE) programme based on 2017 data from facilities worldwide. This is well below regulatory limits of 20-50 mSv per year, reflecting optimized protection measures such as shielding and . Similarly, crew members experience elevated cosmic radiation exposure due to high-altitude flights, with annual effective doses ranging from 2-5 mSv for frequent long-haul routes, particularly polar paths, according to IAEA assessments. For the general public, additional exposure near nuclear power plants is minimal, typically less than 0.1 mSv per year from routine operations, as determined by U.S. Nuclear Regulatory Commission monitoring data showing average doses of about 0.0001 mSv annually within 50 miles of a site. Public exposure to cosmic rays also varies geographically and with altitude; at sea level, it averages 0.38 mSv per year globally, increasing to over 1 mSv at higher elevations or latitudes, per IAEA estimates. In emergency scenarios, such as the 2011 Fukushima Daiichi accident, (TEPCO) workers received cumulative effective doses up to 678 mSv for a few individuals, though most stayed below 100 mSv, with monitoring involving electronic personal dosimeters for external exposure and whole-body counters for internal contamination, as detailed in IAEA reports. These monitored personal dose equivalents, Hp(10), provide a conservative estimate of effective dose for external , generally exceeding the true effective dose E to ensure safety margins in , according to IAEA dosimetry guidelines.

History

Origin and Naming

The sievert (Sv), the for and effective dose in , is named after (1896–1966), a pioneering Swedish renowned for his foundational work in and radiological safety. The name was formally adopted in 1979 by the 16th General Conference on Weights and Measures (CGPM), following recommendations from the International Commission on Radiation Units and Measurements (ICRU), to honor Sievert's lifetime contributions to understanding and mitigating the biological effects of . This adoption aligned with broader efforts to standardize quantities within the (SI), replacing non-coherent legacy units. Prior to 1979, radiation dosimetry relied on units such as the roentgen (R) for exposure, the rad for absorbed dose, and the rem for dose equivalent, which were based on the centimeter-gram-second (CGS) system and lacked direct compatibility with SI base units like the joule and . These units, while practical, created inconsistencies in international scientific communication and applications, prompting the ICRU and (ICRP) to advocate for SI-derived alternatives to ensure coherence and precision in measuring risks to human health. The sievert addressed this by defining dose equivalent as absorbed per unit mass (joules per ), weighted for biological effectiveness. The concept of dose equivalent, for which the sievert later served as , was first formalized in the ICRU's Report 19, Radiation Quantities and Units (1971), with further elaboration in its 1973 supplement specifically on dose equivalent. Sievert's own innovations, including the development of the capacitor-type (known as the Sievert chamber) in the 1920s and standardization of the skin erythema dose, laid critical groundwork for accurate and , influencing these ICRU definitions. For context, 1 Sv equals 100 rem, maintaining numerical continuity with the prior unit during the transition.

Evolution of Definitions

The evolution of the sievert as a unit for effective dose began with the (ICRP) Publication 26 in 1977, which introduced the concept of effective dose equivalent to quantify stochastic risks across the body. This incorporated radiation weighting factors (w_R) to account for differences in biological effectiveness among radiation types and tissue weighting factors (w_T) to reflect varying sensitivities of organs and tissues, replacing earlier, less comprehensive approaches to whole-body dose assessment. The sievert was subsequently adopted as the unit for these quantities in 1979. In 1990, ICRP Publication 60 refined these concepts by updating the w_R values, particularly for neutrons, to better align with emerging data on , and it phased out the use of the quality factor () in favor of the more standardized w_R for protection purposes. These changes aimed to simplify while maintaining conservatism in risk estimation for mixed radiation fields. ICRP Publication 103 in 2007 further evolved the framework by refining the w_T values based on updated epidemiological evidence, such as increasing the remainder tissue weighting from 0.30 to 0.12 distributed across additional organs, including the addition of the to the remainder category for more accurate representation of cancer risks. Although some proposals for operational modifications to dose quantities were considered during this revision, they were ultimately rejected to preserve compatibility with existing regulatory systems. As of 2025, no further updates to the core definitions of the sievert or its associated weighting factors have been issued by the ICRP, despite ongoing discussions regarding the linear no-threshold (LNT) model for low-dose risks; the unit and its conceptual basis remain unchanged from the recommendations.

Key Publications and Revisions

The sievert (Sv) was formally adopted as the special name for the SI unit of dose equivalent by the 16th General Conference on Weights and Measures (CGPM) in , equivalent to one joule per kilogram (J/kg), to quantify the biological effects of on human tissue. This adoption followed the International Commission on Radiation Units and Measurements' (ICRU) introduction of the dose equivalent concept in 1971 and the International Commission on Radiological Protection's (ICRP) early use of the quantity in joules per kilogram in its foundational recommendations. ICRP Publication 26 (1977) marked the seminal introduction of protection quantities using the dose equivalent and effective dose equivalent concepts in radiological protection, defining the dose equivalent as the product of absorbed dose and a quality factor to account for radiation type, and the effective dose equivalent as a weighted sum across tissues to estimate stochastic risk. This publication established the framework for these protection quantities in J/kg (later sieverts), shifting from earlier concepts like the rem and emphasizing detriment from cancer and hereditary effects, with a nominal risk coefficient of 1.25 × 10^{-2} Sv^{-1} for fatal cancer (part of total detriment of 1.65 × 10^{-2} Sv^{-1}). (Note: Full text access via ICRP archives.) Subsequent revisions refined these definitions without altering the sievert unit itself. ICRP Publication 60 (1990) replaced effective dose equivalent with effective dose, incorporating updated tissue weighting factors based on revised risk estimates from atomic bomb survivors and other data, increasing the overall detriment coefficient to 5.0 × 10^{-2} Sv^{-1} for the whole population. This update prioritized sex-averaged risks and separated deterministic effects, maintaining the sievert for summation across exposure scenarios. ICRP Publication 103 (2007) further evolved the framework by revising tissue weighting factors—e.g., elevating those for lungs (0.12) and (0.12) while reducing for gonads (0.08)—to better reflect epidemiological , with the effective dose now serving optimization in planned exposures. It reaffirmed the sievert's role in limiting risks, with total detriment of approximately 5% per Sv. More recent guidance, such as ICRP Publication 147 (2021), clarifies the application of equivalent and effective doses in sieverts for , addressing ambiguities in operational quantities and emphasizing their non-use for individual in contexts. These publications collectively underscore the sievert's enduring utility in balancing protection against varied sources.

Equivalence to Rem

The rem (roentgen equivalent man) is a legacy non-SI unit for measuring dose equivalent, developed to quantify the biological impact of on human tissue in a manner parallel to the modern sievert. Introduced by the (ICRP) in 1954, the rem accounted for the varying effectiveness of different radiation types by weighting , addressing limitations in earlier units like the roentgen that focused solely on exposure. This unit emerged during a period of post-World War II advancements in , where the need for a biologically weighted measure became evident amid growing concerns over nuclear activities. The precise equivalence between the rem and sievert was established with the adoption of the (SI) in the 1970s, defining 1 rem = 0.01 Sv exactly, or conversely, 1 Sv = 100 rem. This direct scalar relationship simplifies conversions across scales; for instance, 100 millirem (mrem) equals 1 millisievert (mSv), and 5 rem equals 0.05 Sv. The sievert, named after physicist Rolf Sievert and based on joules per (J/kg), supersedes the rem as the SI-derived unit for equivalent and effective dose, emphasizing absorbed energy weighted by radiation type and tissue sensitivity. Despite international efforts to standardize on SI units, the rem persists in U.S. regulatory frameworks, such as those from the (NRC), where dose limits for workers and the public are often expressed in rem alongside Sv equivalents. Organizations like the (IAEA) and the International Bureau of Weights and Measures (BIPM) advocate for the exclusive use of the sievert to align global practices, reduce conversion errors, and facilitate international collaboration in radiation safety. This promotion reflects broader SI adoption policies, though the rem's entrenched role in American standards delays a full phase-out.

SI Usage Guidelines

The sievert (Sv) is a derived unit in the (SI), defined as the special name for the unit of dose equivalent, equivalent to one joule per kilogram (J/kg), making it coherent with the SI base units of mass (, kg) and energy (joule, J). This definition ensures consistency in measurements, where the sievert quantifies the biological effectiveness of absorbed doses. The International Bureau of Weights and Measures (BIPM) recommends using the sievert exclusively for equivalent dose, effective dose, and operational dose quantities in SI-compliant reporting to maintain uniformity across scientific and regulatory contexts. In reporting doses, SI guidelines specify expressing values as an Arabic numeral followed by a space and the unit symbol (e.g., 2.4 mSv), with SI prefixes applied for smaller magnitudes to enhance readability. For low-level exposures, such as natural , doses are typically reported in whole numbers using prefixes like millisievert (mSv, 10^{-3} Sv) or microsievert (μSv, 10^{-6} Sv), for example, annual public exposure around 2400 μSv. Higher precision requires decimal places when necessary, but whole numbers suffice for approximate or rounded values below 1 mSv to avoid implying unwarranted accuracy. Unit symbols must be upright (Sv, not italicized) and not abbreviated further, with plural forms identical to singular (e.g., 1 Sv, 10 Sv). International standards, such as ISO 8529, provide protocols for calibrating measuring devices in terms of the sievert, particularly for fields and dosimeters used in protection-level monitoring. Part 3 of ISO 8529 outlines procedures for calibrating area and personal dosimeters with reference radiations, ensuring to SI units and specifying ambient dose equivalent in Sv for operational quantities. These standards emphasize avoiding mixed units, such as combining sievert with non-SI equivalents like rem in the same report, to promote SI coherence; conversions should be provided separately if non-SI units are referenced for legacy compatibility. Common errors in sievert usage include misapplying SI prefixes, such as using them inconsistently (e.g., "mSv" for 10^{-3} but omitting for larger scales), or formatting issues like attaching units directly to numbers without a space (e.g., 10Sv instead of 10 Sv). Another frequent mistake is confusing the sievert symbol (Sv) with non-radiation notations, such as velocity units (m/s), leading to erroneous interpretations in multidisciplinary documents; always contextualize as the radiation dose unit. To prevent such issues, adhere strictly to BIPM and NIST conventions for symbol rendering and unit coherence.

Broader Ionizing Radiation Quantities

The sievert (Sv) is a unit within a broader framework of quantities used to quantify ionizing radiation effects, ranging from initial exposure in the environment to biological impacts on the human body. This hierarchy begins with measures of radiation interaction with air and matter, progressing to quantities that account for radiation type and tissue sensitivity, ultimately informing radiological protection standards. Exposure quantifies the ionization produced by photons (such as X-rays or gamma rays) in air, serving as an early step in assessing fields. The traditional unit for exposure is the roentgen (R), defined as the amount of that produces ions carrying 0.000258 coulombs of charge per of dry air under standard conditions. This quantity does not directly measure energy absorption but provides a basis for estimating subsequent effects in tissue. Following exposure, kerma (kinetic energy released per unit mass) describes the energy transferred from indirectly ionizing radiation (like photons) to charged particles in a medium, such as air or tissue. Air kerma, often used in operational dosimetry, links exposure to potential energy deposition and is expressed in grays (Gy), helping to bridge environmental measurements to absorbed doses in materials. Absorbed dose, measured in grays (Gy), represents the energy deposited by any ionizing radiation per unit mass in a specified material, such as human tissue, and is fundamental to understanding local energy transfer. One gray equals one joule per kilogram. From absorbed dose, the chain advances to equivalent dose, calculated by multiplying absorbed dose by a radiation weighting factor (w_R) to account for the relative biological effectiveness of different radiation types; this yields the sievert as the unit for equivalent dose, focusing on stochastic risks to specific organs. Effective dose, also in sieverts, extends equivalent dose by applying tissue weighting factors (w_T) to sum risks across the whole body, enabling comparisons of overall health impacts from varied exposure scenarios. Thus, the progression—exposure (R) to (Gy) to (Gy) to (Sv) to effective dose (Sv)—provides a comprehensive pathway from source interactions to protective dose limits. Upstream in this framework, activity, measured in , quantifies the source strength as the number of nuclear disintegrations per second, serving as the origin for exposures that lead to the aforementioned dose quantities. One equals one disintegration per second.

References

  1. https://www.nrc.gov/reading-rm/doc-collections/cfr/part020/part020-1004
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