Roger Clarke became Chairman of ICRP in 1993 and is presently (as of January 2005) serving his third term of office. He retired as National Radiological Protection Board Director in June 2003 after 16 years as Director (Source: Health Protection Agency, NRPB Press Releases, Archive for 2005, New Year Honour for Former NRPB Director, 17 January 2005

• Roger H. Clarke, Controllable Dose: A Discussion on the Control of Individual Doses from Single Sources, 1 October 1998
• Controllable Dose: A Discussion on the Control of Individual Doses from Single Sources, Editor's Note, Health Physics Society, December 2009.

Roger H. Clarke, "Controllable Dose: A Discussion on the Control of Individual Doses from Single Sources", 1 October 1998
Quotes:

A Controllable Dose is the dose or the sum of the doses to an individual from a particular source that can reasonably be controlled by whatever means.

THE PRINCIPLE
The protection philosophy for controllable dose is based on the individual. If the individual is sufficiently protected from a single source, then that is a sufficient criterion for the control of the source. In the past ICRP has used societal criteria, using collective dose summed over all populations and all times in cost-benefit analysis to determine the optimum spend on the control of a source. [Contrary to that] This new approach is a totally individual source-related criterion. The principle is

If the risk of harm to the health of the most exposed individual is trivial, then the total risk is trivial - irrespective of how many people are exposed.

The principle of individual source-related protection was recognised by ICRP in ICRP Publication 60 with the introduction of the concept of a constraint on the optimisation process.

CONTROLLABLE DOSE

Fatal Risk (deaths/(persons year))	Dose	Proposed System	Current Criteria
10-3	30 mSv	Dose should not exceed this level - approach only if there is benefit to individual or dose is difficult to reduce or prevent	Occupational Dose Limit Upper Radon Action Level Relocation Intervention Level CT Scan
10-4	3 mSv in a year	There may be a need to reduce or prevent doses, particularly if no benefit to individual	Lower Intervention Level for Shelter, KI-tablets Lower Radon Action Level Average Background Diagnostic Levels
10-5	0.3 mSv in a year	Maximum dose to individual who receives no direct benefit from one source of radiation	Maximum constraint from a single source Background variation - excluding Radon
10-6	0.03 mSv in a year	Trivial risk to the individual	Exemption levels (in cache, 24 Nov. 2011) Clearance Levels (in cache, 24 Nov. 2011)

1. fatal cancer risk which will be of the order of 10^-3 (1 in 1000) per year
   Thus, the highest dose that will normally be tolerated before control is definitely instituted is in the range of a few tens of millisieverts (a few rems) although this may be tolerated in successive years.

   This covers:

   • The permanent relocation of people following an accident is recommended to avert a dose of 1 Sv (100 rem) in a lifetime, which corresponds to some 10s of mSv (several rem) in the first year,
   • The occupational dose limit of 20 mSv (2 rem) in a year,
   • The upper (justified) action level for radon in homes (10 mSv (1 rem) per year),
   • The lower level of averted dose above which evacuation is recommended after an accident (50 mSv = 5 rem), and
   • A CT scan (~few 10s of mSv (some few rems).

   While these levels of dose to the individual can hardly be called unacceptable, they are levels at which questions should be asked as to whether the dose and associated fatal cancer risk which will be of the order of 10^-3 (1 in 1000) per year, can be avoided by some sort of action. That action may be disruptive, or, as in the case of a CT scan, be simply to question whether the required information can be obtained by another means involving lower dose.

   Controllable doses should not generally exceed this level and actual or potential doses approaching this level would only be allowed if the individual receives a benefit or the doses cannot be reduced or prevented without significant disruption to lifestyle.

2. fatal cancer risk which will be of the order of 10^-4 (1 in 10 000) per year
   At levels of controllable dose of the order of a few millisieverts (a few 100 mrem) per year, the exposures should not be of great concern from the point of view of an individual's health. Natural background radiation is about 2-3 mSv (200 - 300 mrem) in a year, and even if radon exposures are excluded, the figure is 1-2 mSv (100 - 200 mrem).

   The exposures covered would be:

   • the lower level of optimized range for radon intervention (3 mSv),
   • the lower level for simple countermeasures (sheltering, KI) in an accident (5 mSv),
   • the existing dose limit for members of the pubic (1 mSv), and
   • simple diagnostic x-ray examinations (1 mSv)

   Steps may be taken to reduce these exposures, or to prevent them, particularly if the individual receives no benefit. Thus from a controllable dose of a few millisieverts upwards it becomes increasingly desirable to reduce or prevent the dose depending both on the practicability of doing so and whether the individual is deriving any tangible benefit from the exposure, e.g. annual occupational exposures or medical examination doses. The associated levels of fatal cancer risk would be 10^-4, 1 in 10,000 (per year).

3. fatal cancer risk which will be of the order of 10^-5 (1 in 100 000) per year
   Doses that are below the millisievert (100 mrem) level are also relevant in the control of exposures. In connection with manmade uses of radiation, the Commission has set the maximum dose from a single new source to a member of the public at 0.3 mSv a year. The associated level of fatal cancer risk is about 10^-5 per year. This level of dose is about 10% of total natural background dose and is also of the same order as to variation in background radiation (excluding the radon contribution) over much of the world.
4. fatal cancer risk which will be of the order of 10^-6 (1 in 1000 000) per year
   A level of risk of death of 10^-6 per year is commonly regarded as trivial and the corresponding annual dose of about 0.01 - 0.02 mSv (1 - 2 mrem) has been used to set exemption criteria for the Basic Safety Standards. At this level of dose there should be no need to consider protection of the individual.

EPA: MCL for man-made beta and photon emitters

In setting the Max. Conc. Limits for man-made beta and photon emitters, EPA used cancer risk estimates from the BEIR I report for the U.S. population in the year 1967 (NAS, 1972). For an exposed group having the same age distribution as the U.S. 1967 population, the BEIR I report indicated that the individual risk of a fatal cancer from a lifetime total body dose rate of 4 mrem per year ranged from about 0.4 to 2 x 10^-6 per year depending on whether an absolute or relative risk model was used. Using best estimates from both models for fatal cancer, EPA believed that an individual risk of

• 0.8 x 10^-6 per year resulting from a 4 mrem annual total body dose

TERMS

Somatic effect means a health effect on an exposed body. With regard to ionizing radiation,
somatic effects mainly refer to cancers and leukemias (USEPA, 1981).

Stochastic effects means effects for which the probability of occurrence is proportional to dose,
but not the severity of effect, and it is assumed that there is no threshold below which they do not
occur (WHO, 1993).

Non-stochastic describes effects whose severity is a function of dose; for these, a threshold may
occur. Examples of non-stochastic somatic effects are cataract induction, nonmalignant damage
to the skin, hematological deficiencies, and impairment of fertility (NIH, 1994). Non-stochastic
may also be referred to as deterministic effects.

Stochastic means random events leading to effects whose probability of occurrence in a exposed
population (rather than severity in an affected individual) is a direct function of dose; these effects
are commonly regarded as having no threshold; heredity effects are regarded as being stochastic;
some somatic effects, especially carcinogens, are regarded as being stochastic (NIH, 1994).

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Ten years after removal of spent fuel from a reactor, the radiation dose 1 meter away from a typical spent fuel assembly exceeds 20,000 rems per hour. A dose of 5,000 rems would be expected to cause immediate incapacitation and death within one weekSource: Radioctive Waster: Production, Storage, Disposal, U.S. Nuclear Regulatory Commission, NUREG/BR-0216, Revision 2, May 2002

DOSES TO INDIVIDUALS

You give three different sets of measurements that might be used for estimating dose to an individual in the measured location from airborne radon and thoron and their daughter products:

1. A set of radon and thoron concentrations (in Bq m^-3),
2. The potential alpha energy concentration (PAEC) of the airborne daughter products, (in mWL) Ð presumably the combined value for radon and thoron, and
3. A pair of separate radon-daughter and thoron-daughter PAECs (in mWL) from air sampling -- presumably for the same location as for measurement sets 1 and 2.

The dose is primarily from the radon and decay products and, thus, the PAEC measurements are the ones most directly related to dose; however, it is often more expeditious to make measurements of the precursor radon gas, and thus, with the application of additional factors or assumptions, dose can also be estimated from airborne radon and thoron concentrations.

The relationship between airborne concentration of radon, thoron, and/or radon and thoron decay products and the associated radiation dose is affected by a number of factors, the dosimetry continues to be reviewed, and as yet there is not complete unanimity as to the appropriate conversion factors. For the purpose of this response we will use some published values as examples.

Alternative 1; Calculating Effective Dose from radon and thoron gas concentrations.
Given: Radon = 75 Bq m^-3 Thoron = 150 Bq m^-3

The 1993 report of UNSCEAR (the United Nations Scientific Committee on the Effects of Atomic Radiation), "Sources and Effects of Ionizing Radiation," uses a radon conversion factor of 25 µSv y^-1 per Bq m^-3 for an indoor occupancy factor of 0.8 or about 7,000 h y^-1). From the same report a thoron conversion factor of 25 µSv y^-1 per Bq m^-3 it can be inferred for the same occupancy conditions.

Effective dose = (75 x 25) +(150 x 22) = 5,175 µSv y^-1 or 5.2 mSv y^-1.

Alternative 2; Calculating Effective Dose from radon plus thoron PAEC.
Given: PAEC =20 mWL.

For this case we need a PAEC-to-dose conversion factor. Factors have been published for estimating effective dose from the cumulative exposure in WLM (working level months), where a WLM is exposure to one WL for 170 h. Usually an occupancy factor of 0.7 for the home is assumed, and thus for 1 year in a home with 1 WL, the exposure in WLM is (365 x 24 x 0.7/170) = 36 WLM.

We are not aware of a published single factor for radon daughters and thoron daughters combined. The ICRP (International Commission on Radiological Protection), in its Publication 65, "Protection Against Radon-222 at Home and at Work" (1993), presents a radon daughter dose coefficient for members of the public of 3.9 mSv per WLM. ICRP 65 does not give a dose coefficient for thoron, but ICRP Publication 32, "Limits for Inhalation of Radon Daughters by Workers" (1981), gives an effective dose equivalent of 3.4 mSv per WLM of thoron daughters. These two reports are based on different dosimetry models and the two conversion factors are not necessarily directly comparable. However, in the absence of an alternative, let us use the ICRP thoron values for the purpose of making an estimate. The values for radon daughters and thoron daughters are not greatly different, and thus for an estimate for the radon-thoron aggregate we will use an average value of 3.6 mSv per WLM.

This gives a dose estimate for radon daughters +thoron daughters at 20 mWL:
0.020 WL x 36 WLM/WL-y x 3.6 mSv WLM^-1 = 2.6 mSv y^-1.

Alternative 3; Calculating Effective Dose from radon PAEC and thoron PAEC.
Given: Radon daughter PAEC = 5 mWL; Thoron daughter PAEC = 10 mWL.

Effective dose equivalent estimates then are:

• Radon daughters at 5 mWL:
  0.005 WL x 36 WLM/WL-y x 3.9 mSv WLM^-1 = 0.70 mSv y^-1.
• Thoron daughters at 10 mWL:
  0.010 WL x 36 WLM/WL-y x 3.4 mSv/WLM = 1.22 mSv y^-1.
• Radon daughters + Thoron daughters: 1.9 mSv y^-1.

Comment on the Three Alternative Estimates. The three estimates, 5.2, 2.6, and 1.9 mSv y^-1 are different but of the same order of magnitude. Differences probably reflect differences in precision and accuracy of the three measurement methods, the uncertainties in the three difference calculational methods, and differences in dose conversion conventions developed and published at different times by different groups.

POPULATION DOSE

If all the individuals in the population of interest are exposed to the same concentrations, then the estimates of individual dose represent the average dose rate (mSv y^-1) for members of the population and the population dose (person-Sv) is the average dose times the number of persons in the population.

If different subgroups of the population are exposed to different concentrations (and hence receive different average doses), then a calculation should be made of the typical individual dose for each population group. The average dose for the whole population is the person-weighted average across all groups and total population dose (person-Sv) is the sum of the population doses for the individual subgroups. ===================

Table 1—Release Limits for Containment Requirements

[Cumulative releases to the accessible environment for 10,000 years after disposal]

Radionuclide	Release limit per 1,000 MTHM or other unit of waste (see notes) (curies)
Americium-241 or -243	100
Carbon-14	100
Cesium-135 or -137	1,000
Iodine-129	100
Neptunium-237	100
Plutonium-238, -239, -240, or -242	100
Radium-226	100
Strontium-90	1,000
Technetium-99	10,000
Thorium-230 or -232	10
Tin-126	1,000
Uranium-233, -234, -235, -236, or -238	100
Any other alpha-emitting radionuclide with a half-life greater than 20 years	100
Any other radionuclide with a half-life greater than 20 years that does not emit alpha particles	1,000

2.55

Application of Table 1

Note 1: Units of Waste. The Release Limits in Table 1 apply to the amount of wastes in any one of the following:

(a) An amount of spent nuclear fuel containing 1,000 metric tons of heavy metal (MTHM) exposed to a burnup between 25,000 megawatt-days per metric ton of heavy metal (MWd/MTHM) and 40,000 MWd/MTHM;

(b) The high-level radioactive wastes generated from reprocessing each 1,000 MTHM exposed to a burnup between 25,000 MWd/MTHM and 40,000 MWd/MTHM;

(c) Each 100,000,000 curies of gamma or beta-emitting radionuclides with half-lives greater than 20 years but less than 100 years (for use as discussed in Note 5 or with materials that are identified by the Commission as high-level radioactive waste in accordance with part B of the definition of high-level waste in the NWPA);

(d) Each 1,000,000 curies of other radionuclides (i.e., gamma or beta-emitters with half-lives greater than 100 years or any alpha-emitters with half-lives greater than 20 years) (for use as discussed in Note 5 or with materials that are identified by the Commission as high-level radioactive waste in accordance with part B of the definition of high-level waste in the NWPA); or

(e) An amount of transuranic (TRU) wastes containing one million curies of alpha-emitting transuranic radionuclides with half-lives greater than 20 years.

Note 2: Release Limits for Specific Disposal Systems. To develop Release Limits for a particular disposal system, the quantities in Table 1 shall be adjusted for the amount of waste included in the disposal system compared to the various units of waste defined in Note 1. For example:

(a) If a particular disposal system contained the high-level wastes from 50,000 MTHM, the Release Limits for that system would be the quantities in Table 1 multiplied by 50 (50,000 MTHM divided by 1,000 MTHM).

(b) If a particular disposal system contained three million curies of alpha-emitting transuranic wastes, the Release Limits for that system would be the quantities in Table 1 multiplied by three (three million curies divided by one million curies).

(c) If a particular disposal system contained both the high-level wastes from 50,000 MTHM and 5 million curies of alpha-emitting transuranic wastes, the Release Limits for that system would be the quantities in Table 1 multiplied by 55:

$$

Note 3: Adjustments for Reactor Fuels with Different Burnup. For disposal systems containing reactor fuels (or the high-level wastes from reactor fuels) exposed to an average burnup of less than 25,000 MWd/MTHM or greater than 40,000 MWd/MTHM, the units of waste defined in (a) and (b) of Note 1 shall be adjusted. The unit shall be multiplied by the ratio of 30,000 MWd/MTHM divided by the fuel's actual average burnup, except that a value of 5,000 MWd/MTHM may be used when the average fuel burnup is below 5,000 MWd/MTHM and a value of 100,000 MWd/MTHM shall be used when the average fuel burnup is above 100,000 MWd/MTHM. This adjusted unit of waste shall then be used in determining the Release Limits for the disposal system.

For example, if a particular disposal system contained only high-level wastes with an average burnup of 3,000 MWd/MTHM, the unit of waste for that disposal system would be:

$$

If that disposal system contained the high-level wastes from 60,000 MTHM (with an average burnup of 3,000 MWd/MTHM), then the Release Limits for that system would be the quantities in Table 1 multiplied by ten:

$$

which is the same as:

$$

Note 4: Treatment of Fractionated High-Level Wastes. In some cases, a high-level waste stream from reprocessing spent nuclear fuel may have been (or will be) separated into two or more high-level waste components destined for different disposal systems. In such cases, the implementing agency may allocate the Release Limit multiplier (based upon the original MTHM and the average fuel burnup of the high-level waste stream) among the various disposal systems as it chooses, provided that the total Release Limit multiplier used for that waste stream at all of its disposal systems may not exceed the Release Limit multiplier that would be used if the entire waste stream were disposed of in one disposal system.

Note 5: Treatment of Wastes with Poorly Known Burnups or Original MTHM. In some cases, the records associated with particular high-level waste streams may not be adequate to accurately determine the original metric tons of heavy metal in the reactor fuel that created the waste, or to determine the average burnup that the fuel was exposed to. If the uncertainties are such that the original amount of heavy metal or the average fuel burnup for particular high-level waste streams cannot be quantified, the units of waste derived from (a) and (b) of Note 1 shall no longer be used. Instead, the units of waste defined in (c) and (d) of Note 1 shall be used for such high-level waste streams. If the uncertainties in such information allow a range of values to be associated with the original amount of heavy metal or the average fuel burnup, then the calculations described in previous Notes will be conducted using the values that result in the smallest Release Limits, except that the Release Limits need not be smaller than those that would be calculated using the units of waste defined in (c) and (d) of Note 1.

Note 6: Uses of Release Limits to Determine Compliance with §191.13 Once release limits for a particular disposal system have been determined in accordance with Notes 1 through 5, these release limits shall be used to determine compliance with the requirements of §191.13 as follows. In cases where a mixture of radionuclides is projected to be released to the accessible environment, the limiting values shall be determined as follows: For each radionuclide in the mixture, determine the ratio between the cumulative release quantity projected over 10,000 years and the limit for that radionuclide as determined from Table 1 and Notes 1 through 5. The sum of such ratios for all the radionuclides in the mixture may not exceed one with regard to §191.13(a)(1) and may not exceed ten with regard to §191.13(a)(2).

For example, if radionuclides A, B, and C are projected to be released in amounts Q_a, Q_b, and Q_c, and if the applicable Release Limits are RL_a, RL_b, and RL_c, then the cumulative releases over 10,000 years shall be limited so that the following relationship exists:

$$

[50 FR 38084, Sept. 19, 1985, as amended at 58 FR 66415, Dec. 20, 1993]

Appendix B to Part 191—Calculation of Annual Committed Effective Dose

I. Equivalent Dose

The calculation of the committed effective dose (CED) begins with the determination of the equivalent dose, H_T, to a tissue or organ, T, listed in Table B.2 below by using the equation:

$$

where D_T,Ris the absorbed dose in rads (one gray, an SI unit, equals 100 rads) averaged over the tissue or organ, T, due to radiation type, R, and w_Ris the radiation weighting factor which is given in Table B.1 below. The unit of equivalent dose is the rem (sievert, in SI units).

Table B.1—Radiation Weighting Factors, w_R¹

Radiation type and energy range2	wRvalue
Photons, all energies	1
Electrons and muons, all energies	1
Neutrons, energy < 10 keV	5
10 keV to 100 keV	10
>100 keV to 2 MeV	20
>2 MeV to 20 MeV	10
>20 MeV	5
Protons, other than recoil protons, >2 MeV	5
Alpha particles, fission fragments, heavy nuclei	20

¹All values relate to the radiation incident on the body or, for internal sources, emitted from the source.

²See paragraph A14 in ICRP Publication 60 for the choice of values for other radiation types and energies not in the table.

II. Effective Dose

The next step is the calculation of the effective dose, E. The probability of occurrence of a stochastic effect in a tissue or organ is assumed to be proportional to the equivalent dose in the tissue or organ. The constant of proportionality differs for the various tissues of the body, but in assessing health detriment the total risk is required. This is taken into account using the tissue weighting factors, w_Tin Table B.2, which represent the proportion of the stochastic risk resulting from irradiation of the tissue or organ to the total risk when the whole body is irradiated uniformly and H_Tis the equivalent dose in the tissue or organ, T, in the equation:

$$

Table B.2—Tissue Weighting Factors, w_T¹

Tissue or organ	wTvalue
Gonads	0.25
Breast	0.15
Red bone marrow	0.12
Lung	0.12
Thyroid	0.03
Bone surfaces	0.03
Remainder	20.30

¹The values are considered to be appropriate for protection for individuals of both sexes and all ages.

²For purposes of calculation, the remainder is comprised of the five tissues or organs not specifically listed in Table B.2 that receive the highest dose equivalents; a weighting factor of 0.06 is applied to each of them, including the various sections of the gastrointestinal tract which are treated as separate organs. This covers all tissues and organs except the hands and forearms, the feet and ankles, the skin and the lens of the eye. The excepted tissues and organs should be excluded from the computation of H_E.

III. Annual Committed Tissue or Organ Equivalent Dose

For internal irradiation from incorporated radionuclides, the total absorbed dose will be spread out in time, being gradually delivered as the radionuclide decays. The time distribution of the absorbed dose rate will vary with the radionuclide, its form, the mode of intake and the tissue within which it is incorporated. To take account of this distribution the quantity committed equivalent dose, HΤ(τ) where is the integration time in years following an intake over any particular year, is used and is the integral over time of the equivalent dose rate in a particular tissue or organ that will be received by an individual following an intake of radioactive material into the body. The time period, τ, is taken as 50 years as an average time of exposure following intake:

$$

for a single intake of activity at time t₀where H_T(t) is the relevant equivalent-dose rate in a tissue or organ at time t. For the purposes of this part, the previously mentioned single intake may be considered to be an annual intake.

IV. Annual Committed Effective Dose

If the committed equivalent doses to the individual tissues or organs resulting from an annual intake are multiplied by the appropriate weighting factors, w_T, and then summed, the result will be the annual committed effective dose, E(τ):

$$

[58 FR 66415, Dec. 20, 1993]

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Subpart C--Occupational Dose Limits

Source: 56 FR 23396, May 21, 1991, unless otherwise noted.

§ 20.1201 Occupational dose limits for adults.

(a) The licensee shall control the occupational dose to individual adults, except for planned special exposures under § 20.1206, to the following dose limits.

(1) An annual limit, which is the more limiting of--

(i) The total effective dose equivalent being equal to 5 rems (0.05 Sv); or

(ii) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 Sv).

(2) The annual limits to the lens of the eye, to the skin of the whole body, and to the skin of the extremities, which are:

(i) A lens dose equivalent of 15 rems (0.15 Sv), and

(ii) A shallow-dose equivalent of 50 rem (0.5 Sv) to the skin of the whole body or to the skin of any extremity.

(b) Doses received in excess of the annual limits, including doses received during accidents, emergencies, and planned special exposures, must be subtracted from the limits for planned special exposures that the individual may receive during the current year (see § 20.1206(e)(1)) and during the individual's lifetime (see § 20.1206(e)(2)).

(c) When the external exposure is determined by measurement with an external personal monitoring device, the deep-dose equivalent must be used in place of the effective dose equivalent, unless the effective dose equivalent is determined by a dosimetry method approved by the NRC. The assigned deep-dose equivalent must be for the part of the body receiving the highest exposure. The assigned shallow-dose equivalent must be the dose averaged over the contiguous 10 square centimeters of skin receiving the highest exposure. The deep-dose equivalent, lens-dose equivalent, and shallow-dose equivalent may be assessed from surveys or other radiation measurements for the purpose of demonstrating compliance with the occupational dose limits, if the individual monitoring device was not in the region of highest potential exposure, or the results of individual monitoring are unavailable.

(d) Derived air concentration (DAC) and annual limit on intake (ALI) values are presented in table 1 of appendix B to part 20 and may be used to determine the individual's dose (see § 20.2106) and to demonstrate compliance with the occupational dose limits.

(e) In addition to the annual dose limits, the licensee shall limit the soluble uranium intake by an individual to 10 milligrams in a week in consideration of chemical toxicity (see footnote 3 of appendix B to part 20).

(f) The licensee shall reduce the dose that an individual may be allowed to receive in the current year by the amount of occupational dose received while employed by any other person (see § 20.2104(e)).

Code of Federal Regulations (10 CFR)

Part 20 "Standards for protection against radiation",

• Appendix B - Radionuclide Table - Index of Radioisotopes.

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Canadian Guidelines for the Restriction of Radioactively Contaminated Food and Water Following a Nuclear Emergency

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