Datensammlung zur Strahlenbelastung

Dr. Joachim Gruber

Units

SI-Unit Old Unit (Source) - Comment
Activity Bequerel (Bq) Curie (1 Ci = 3.7 1010Bq)
Radiation Dose Roentgen (R) 1 R is the amount of gamma or x-rays required to produce ions resulting in a charge of 0.000258 coulombs/kilogram of air under standard conditions.

1 R is the amount of radiation that produces 1 coulomb of ions in 1 cm3 of dry air at zero degrees celsius at standard atmospheric pressure.

Energiedosis
Absorbed Dose
Gray (Gy) = 1 Joule/kg

Gy squaremeter = measure of intensity of x-ray device
Quelle: R. Padovani, Training Course on Radiation Protection Training for Interventional Cardiologists and Radiologists, Dose Quantities and Units (in cache)
zum Vergrößern auf Bild klicken
Definition: DAP (Gy cm2) = Dosis x Fläche

absorbed dose, D
Source: IAEA, Nuclear Installation Safety Net - Tutorials - Regulatory Control, Glossary D (in cache)
The fundamental dosimetric quantity D is defined as:
D = dε/dm
where dε is the mean energy imparted by ionizing radiation to matter in a volume element and dm is the mass of matter in the volume element.
The energy can be averaged over any defined volume, the average dose being equal to the total energy imparted in the volume divided by the mass in the volume.
Absorbed dose is defined at a point; for the average dose in a tissue or organ, see organ dose (in cache). Unit: J/kg, termed the gray (Gy) (formerly, the rad was used).

1 Gy = 100 Rad
1 Rad = 0.01 Gy = 1 cGy
10 mGy = 1 cGy (analogous to 10 mm = 1 cm = 0.01 m)

1.15 R of gamma radiation exposure results in 1 Rad of absorbed dose.

Äquivalentdosis
Equivalent Dose

Strahlungswichtungsfaktor - Radiation Weighting Factor - Quality Factor (wR)

Effektive Dosis
Committed Effective Dose

Gewebewichtungsfaktor - Organ Weighting Factor (wT)

Sievert (Sv) = Gy x wR
1 Sv = 100 rem
with 1 year = 365.2422 hours follows:
1 μSv/h = 8766 μSv/a = 876600 μrem/a ≈ 1 rem/a

definitions of dose
Quelle: IAEA, Nuclear Installation Safety Net - Tutorials - Regulatory Control, Glossary D (in cache)
zum Vergrößern auf Bild klicken
Definitionen: Dosis

quality factors
Quelle: Annex 1.1, Umweltradioaktivität und Strahlenbelastung, Jahresbericht 2009, Bundesamt für Strahlenschutz (im Cache)
zum Vergrößern auf Bild klicken
Abbildung: Radiation Weighting Factors wR (deutsch: Strahlungswichtungfaktoren).
wR is also called "Quality Factor" Q

Gewebewichtungsfaktoren
Quelle: Annex 1.2, Umweltradioaktivität und Strahlenbelastung, Jahresbericht 2009, Bundesamt für Strahlenschutz, BfS (im Cache)
zum Vergrößern auf Bild klicken
Abbildung: Organ Weighting Factors wT (deutsch: Gewebewichtungsfaktoren)

"Dadurch, dass die effektive Dosis entsprechend der Strahlenwirkung gewichtet ist und damit einem bestimmten Risiko zugeordnet [worden ist], lassen sich die effektiven Dosen aus verschiedenen [Strahlen]Quellen zu einem Wert für die Gesamtstrahlenbelastung eines Menschen addieren. Auf der Ebene der Dosisbetrachtung [werden auf diese Weise] die Beiträge aus natürlichen, zivilatorisch veränderten und zivilisatorischen Strahlenquellen verglichen." (Quelle BfS)

Dose Concept
Source: Risk Assessment: Guidance for Superfund Volume 1, EPA, 2011
Dose Equivalent (HT) (deutsch: "Organdosis")
The product of the absorbed dose (DT,R, Gray), the radiation weighting factor (wR), and any other modifying factors (omitted henceforth):

HT = wR DT,R

where DT,R is the absorbed dose delivered by radiation type R averaged over a tissue or organ T and wR is the radiation weighting factor for radiation type R. When the radiation field is composed of different radiation types with different values of wR the equivalent dose is:

HT = ΣR wR DT,R

The SI unit of dose equivalent is the sievert (Sv); the conventional unit is the rem (1 rem = 0.01 Sv).

Committed Dose Equivalent (HT,50)
The dose equivalent (HT) deposited over the 50-year period following the intake of a radionuclide.

Effective Dose Equivalent (HE) (deutsch: "Effektive Dosis")
The sum over specified tissues of the products of the dose equivalent in a tissue or organ (T) and the weighting factor wT for that tissue:

HE = ΣT wT HT

ΣT wT = 1

Committed Effective Dose Equivalent (HE,50)
The weighted sum of committed dose equivalents to specified organs and tissues, in analogy to the effective dose equivalent.

HE,50 = ΣT wT HT,50

HT,50, HE,50 as defined in Title 10, Section 20.1003, of the Code of Federal Regulations (10 CFR 20.1003)

Radionuclides Notice of Data Availability Technical Support Document, USEPA,  March 2000 (in cache))

EPA knew that partial body irradiation was common for ingested radionuclides since they are, like radium, largely deposited in a particular organ. 
In such cases, EPA acknowledged that the risk per millirem varies depending on the radiosensitivity of the organs at risk.

For example (thyroid exposure compared with whole body exposure):

EPA estimated that cancers due to the thyroid gland receiving 4 mrem per year continuously ranged from about 0.2 to 0.5 per year per million exposed persons (averaged over all age groups). 

Fatality due to thyroid cancer varies with age, from nearly zero for children and young adults to about 20 percent of the incidence for persons well past middle age. 
EPA noted (therefore) that estimated fatalities from thyroid exposure were at least five times less than that from whole body exposure.

Äquivalentdosis - eine Zusammenfassung
(Quelle: Annex 1.2, Umweltradioaktivität und Strahlenbelastung, Jahresbericht 2009, Bundesamt für Strahlenschutz (im Cache)

Die Wirkung einer Strahlenart auf ein Organ [(Organdosis, HT) wird] durch Multiplikation der Energiedosis DT, R [Gray] mit dem Strahlungswichtungsfaktor wR und dem Gewebewichtungsfaktor wT in Zahlen gefasst.

Soll die effektive Dosis E ("Äquivalentdosis") für eine Strahlenexposition durch mehrere Strahlungsarten berechnet werden, die mehrere Organe betreffen, so müssen die genannten Formeln zusammengefasst und eine Summierung über alle Komponenten durchgeführt werden:

E = ΣT wT HT = ΣT wT ΣR wR DT, R

Dosisfaktor Sv/(Bq incorporated)
 
 
 
 
 

dose equivalent (Sv) = dose factor x Bq incorporated

dose factor

  • used in Offsite Dose Calculation Manual, Commonwealth Edison Company, April 1999 (in Cache) and
  • specified in Regulatory Guide 1.109 "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977, US Nuclear Regulatory Commission (in Cache)


 
 
 
 
 
 
 
 

Activity (Bq) deposited in an organ depends on intake pathway. Sv in an organ is a function of Bq deposited in that organ. Dose factor depends on intake pathway, because Bq deposited varies with intake pathway.

pathway dose factors, unit: (mrem/hr) / (μCi/ml) ...
... are basically the dose conversion factors (Sv/Bq or mrem/μCi) multiplied by transfer factors relating the radionuclide spill (contamination of released water (μCi/ml), release period (hr)) to radionuclide uptake (μCi) via that pathway:

dose due to spill (mrem) = pathway dose factor x contamination of spilled water (μCi/ml) x duration of spill (hr)

see e.g. vegetation pathway dose factor

Basically, calculation of exposure via a pathway

  1. starts with calculation of contamination Ci of food in that pathway (Ci/kg),
  2. multiplies Ci with amount M of food consumed (kg),
  3. multiplies Ci M with dose conversion factor (rem/Ci).
(Equation (A-23), page A-12 in "Offsite Dose Calculation Manual", Commonwealth Edison Company, April 1999, in cache)

Dose Conversion Factors

Radionuclides Notice of Data Availability Technical Support Document, USEPA,  March 2000
(alternative link, in )

US Nuclear Regulatory Commission
Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents
for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977

-------------- click on text images to enlarge --------------

USNRC Regulatory Guide 1.109, Gleichung 1 - 4
USNRC Regulatory Guide 1.109, Gleichung 5
USNRC Regulatory Guide 1.109, Notation
USNRC Regulatory Guide 1.109, Notation

Annual Limit of Intake (ALI) Annual limit on intake (ALI)

As defined in Title 10, Section 20.1003, of the Code of Federal Regulations (10 CFR 20.1003), ALI is the derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the "reference man" that would result in

  • a committed dose equivalent (CDE) HT.50 = 50 rems (0.5 Sv, in cache) to any individual organ or tissue, or
  • a committed effective dose equivalent (CEDE) HE.50 = 5 rems (0.05 sievert, in cache).

ALI values for intake by ingestion and inhalation of selected radionuclides are given in Table 1, Columns 1 and 2, of Appendix B to 10 CFR Part 20 (in cache), "Standards for Protection Against Radiation." For additional detail, see Information for Radiation Workers .

Displacements per Atom (dpa)


from
"1. Fundamental of Radiation Damage and Defect Production (NE220)"
by
Brian D. Wirth and D.R. Olander
Nuclear Engineering Department, UC Berkeley, Spring 2008
in Google Documents





Primordial Radionuclides

Primordial radionuclides are left over from when the world and the universe was created. They are typically long lived, with half-lives often on the order of hundreds of millions of years. Radionuclides that exist for more than 30 half-lives are not measurable. The progeny or decay products of the long lived radionuclides are also in this heading. Here is some basic information on some common primordial radionuclides:
 
Primordial nuclides
Nuclide Symbol Half-life Natural Activity
Uranium 235 235U 7.04 x 108 yr  0.72% of all natural uranium 
Uranium 238 238U 4.47 x 109 yr  99.2745% of all natural uranium; 0.5 to 4.7 ppm total uranium in the common rock types 
Thorium 232 232Th 1.41 x 1010 yr  1.6 to 20 ppm in the common rock types with a crustal average of 10.7 ppm 
Radium 226 226Ra 1.60 x 103 yr  0.42 pCi/g (16 Bq/kg) in limestone and 1.3 pCi/g (48 Bq/kg) in igneous rock 
Radon 222 222Rn 3.82 days  Noble Gas; annual average air concentrations range in the US from 0.016 pCi/L (0.6 Bq/m3) to 0.75 pCi/L (28 Bq/m3
Potassium 40 40K 1.28 x 109 yr  soil - 1-30 pCi/g (0.037-1.1 Bq/g) 

Some nuclides like 232Th have several members of its decay chain. You can roughly follow the chain starting with 232Th

232Th --> 228Ra --> 228Ac --> 228Th --> 224Ra -->
     220Rn --> 216Po --> 212Pb --> 212Bi --> 212Po --> 208Pb (stable)

Some other primordial radionuclides are 50V, 87Rb, 113Cd, 115In, 123Te, 138La, 142Ce, 144Nd, 147Sm, 152Gd, 174Hf, 176Lu, 187Re, 190Pt, 192Pt, 209Bi.

Nuclear Data

United States Geological Survey Digital maps of estimated potassium, equivalent uranium-238, equivalent thorium-232 concentrations for the conterminous U.S.
 

Russ Brown:
Potassium-40 content of the body can be obtained from its natural abundance of 0.0117% of potassium and calculating the specific activity of natural potassium (30.5 Bg/g) using the half life (1.28 x 109 y). The potassium content of the body is 0.2%, so for a 70 kg man the amount of 40-K will be about 4.26 kBq.

14C content of the body is based on the fact that one 14C atom exists in nature for every 1,000,000,000,000 12C atoms in living material. Using a half life of 5730 y, one obtains a specific activity of 0.19Bq/g of carbon. As carbon is 23 percent of the body weight, the body content of 14-C for a 70 kg man would be about 3.08 kBq.
 
 

Radon in the Air

Rn-222 decay scheme (uranium series), cache,
decay to Po-218

Übersichtskarte der Radonkonzentration in der Bodenluft der Bundesrepublik Deutschland (im Cache)


Bundesland Anzahl Wohngebäude in Tausend Medianwert
in Bq/m3 (mSv/a)
geschätzter Anteil in % mit einer Belastung
>200 Bq/m3
geschätzter Anteil in % mit einer Belastung
>400 Bq/m3
Baden-Württemberg 1 831,8 a) 38 (0.65) 2,9 0,6
Bayern 2 218,8 a) 41 (0.70) 3,3 1,2
Berlin 265,8 a)b) 27 (0.46) 0,3 0
Brandenburg 472,6 b) 26 (0.44) 0 0
Bremen 121,1 a) 30 (0.51) 0 0
Hamburg 207,2 a) 27 (0.46) 0,2 0
Hessen 1 102,8 a) 40 (0.68) 1 0,2
Mecklenburg-Vorpommern 294,6 b) 39 (0.66) > 0 1) > 0 1)
Niedersachsen 1 577,1 a) 32 (0.54) 0,8 0,1
Nordrhein-Westfalen 2 976,9 a) 35 (0.60) 0,6 0,2
Rheinland-Pfalz 884,1 a) 51 (0.87) 2,3 0,4
Saarland 261,5 a) 46 (0.78) 1,8 0,4
Sachsen 692,9 b) 47 (0.80) 6,5 2
Sachsen-Anhalt 502,6 b) 34 (0.58) 0,8 > 0 1)
Schleswig-Holstein 576,5 a) 36 (0.61) 1,1 > 0 1)
Thüringen 462,1 b) 54 (0.92) 3,9 0,7
a) Gebäude- und Wohnungszählung vom 25. Mai 1987
b) Gebäude- und Wohnungszählung 1995
1)aufgrund der Erfahrungen über die regionale Verteilung des Radonpotentials im Untergrund ggf. tatsächlicher Prozentsatz höher als die über den aus Stichproben ermittelten Ergebnisse.
Quelle: Bundesamt für Strahlenschutz [22-6]
 


EPA Radon Risk Comparison Charts


RADON RISK IF YOU SMOKE
Radon Level If 1,000 people who smoked were exposed to this level over a lifetime... The risk of cancer from radon exposure compares to... WHAT TO DO:
Stop smoking and...
20 pCi/L About 135 people could get lung cancer 100 times the risk of drowning Fix your home
10 pCi/L About 71 people could get lung cancer 100 times the risk of dying in a home fire Fix your home
8 pCi/L About 57 people could get lung cancer Fix your home
4 pCi/L About 29 people could get lung cancer 100 times the risk of dying in an airplane crash Fix your home
2 pCi/L About 15 people could get lung cancer 2 times the risk of dying in a car crash Consider fixing between 2 and 4 pCi/L
1.3 pCi/L
= 48 Bq/m3
About 9 people could get lung cancer (Average indoor radon level) (Reducing radon levels below 2 pCi/L is difficult.)
0.4 pCi/L
= 15 Bq/m3
About 3 people could get lung cancer (Average outdoor radon level) (Reducing radon levels below 2 pCi/L is difficult.)
Note: If you are a former smoker, your risk may be lower.

 
 
 
RADON RISK IF YOU HAVE NEVER SMOKED
Radon Level If 1,000 people who never smoked were exposed to this level over a lifetime... The risk of cancer from radon exposure compares to... WHAT TO DO:
20 pCi/L About 8 people could get lung cancer The risk of being killed in a violent crime Fix your home
10 pCi/L About 4 people could get lung cancer Fix your home
8 pCi/L About 3 people could get lung cancer 10 times the risk of dying in an airplane crash Fix your home
4 pCi/L About 2 people could get lung cancer The risk of drowning Fix your home
2 pCi/L About 1 person could get lung cancer The risk of dying in a home fire Consider fixing between 2 and 4 pCi/L
1.3 pCi/L
= 48 Bq/m3
Less than 1 person could get lung cancer (Average indoor radon level) (Reducing radon levels below 2 pCi/L is difficult.)
0.4 pCi/L
= 15 Bq/m3
Less than 1 person could get lung cancer (Average outdoor radon level) (Reducing radon levels below 2 pCi/L is difficult.)
Note: If you are a former smoker, your risk may be higher.

Radon Frequently Asked Questions
 
 

Cosmogenic Radioactivity

(in cache)

Cosmic radiation is really divided into two types, primary and secondary.

The atmosphere and the Earth's magnetic fields also act as shields against cosmic radiation, reducing the amount that reaches the Earth's surface. With that in mind, it is easy to see that the annual dose you get from cosmic radiation depends on what altitude you are at. From cosmic radiation the U.S., the average person will receive a dose of 27 mrem per year and this roughly doubles every 6,000 foot increase in elevation.

Cosmic radiation permeates all of space, the source being primarily outside of our solar system. The radiation is in many forms, from high speed heavy particles to high energy photons and muons. The upper atmosphere interacts with many of the cosmic radiations, and produces radioactive nuclides. They can have long half-lives, but the majority have shorter half-lives than the primordial nuclides. Here is an table with some common cosmogenic nuclides:
 
Cosmogenic Nuclides
Nuclide Symbol Half-life  Source Natural Activity
Carbon 14 14C 5730 yr  Cosmic-ray interactions, 14N(n,p)14C; 6 pCi/g (0.22 Bq/g) in organic material 
Tritium 3 3H 12.3 yr  Cosmic-ray interactions with N and O; spallation from cosmic-rays, 6Li(n,alpha)3H 0.032 pCi/kg (1.2 x 10-3 Bq/kg) 
Beryllium 7 7Be 53.28 days  Cosmic-ray interactions with N and O;  0.27 pCi/kg (0.01 Bq/kg) 

Some other cosmogenic radionuclides are 10Be, 26Al, 36Cl, 80Kr, 14C, 32Si, 39Ar, 22Na, 35S, 37Ar, 33P, 32P, 38Mg, 24Na, 38S, 31Si, 18F, 39Cl, 38Cl, 34mCl.
 
 


Natürlicher Neutronenhintergrund

Neutronenhintergrundstrahlung

zum Vergrößern auf Bild klicken
Quelle: Messungen von Neutronenspektren induziert durch kosmische Strahlung in Höhen von 85 m, 1195 m und 2650 m, Physikalisch-Technische Bundesanstalt, 2010 (im Cache: pdf, html)

Abbildung: Neutronenhintergrundstrahlung in Abhängigkeit von der Höhe

"Da die Anzahldichte der Neutronen aus der kosmischen Strahlung sehr klein ist, dauern die Messkampagnen in der Regel 6 bis 8 Wochen. Um für den gesamten Zeitraum aus den gesammelten Daten ein gemitteltes Ergebnis zu bekommen, ist es unerlässlich auf den wechselnden Luftdruck und andere Umwelteinflüsse (wie z. B. auf zusätzliche Moderation bei Schneefall) zu korrigieren."

Quelle: Messung des natürlichen Neutronenuntergrunds, hervorgerufen von der kosmischen Strahlung, Physikalisch-Technische Bundesanstalt, 18.8.2010 (im Cache)

"Die Erde wird fortwährend von hochenergetischen (schnellen), geladenen Teilchen aus den Tiefen des Alls getroffen. Diese kosmische Strahlung besteht

  • zu etwa 98% aus Atomkernen:
    • 87% Wasserstoffkerne, d.h. Protonen,
    • 12% Heliumkerne, und
    • etwa 1% schwere Kerne.
  • und zu 2% aus Elektronen.

    "Ein Teil dieser Atomkerne wechselwirkt mit den Molekülen und Atomkernen der Erdatmosphäre und erzeugt dabei schnelle Neutronen. Während die geladenen Teilchen mit größerer Eindringtiefe in die Atmosphäre stark abgebremst und ihre Anzahl auf dem Weg bis zum Erdboden sehr stark reduziert wird, erreicht ein weitaus größerer Teil der erzeugten Neutronen die Erdoberfläche mit Energien im meV Bereich (thermische Neutronen) bis hin zu einigen hundert MeV."

    "Die Neutronenfluenz variiert

    • mit der Höhe über dem Meeresspiegel - genauer mit der Dicke der durchlaufenen Luftschicht - und folglich auch mit dem Luftdruck,
    • mit der geografischen Breite und
    • langfristig mit dem 11-jährigen Sonnenzyklus."

    "Diese natürliche Neutronenstrahlung verursacht einen wesentlichen Beitrag der Strahlenbelastung in Flughöhen. Auf dem Erdboden ist ihr Anteil zwar gering; weniger als 10% der Äquivalentdosis wird hier durch Neutronen verursacht."

    "Trotzdem sind Verfahren für die Messung dieser geringen Dosisleistung von Interesse, um zum Beispiel in der Umgebung von kerntechnischen Anlagen (Kernkraftwerke, Zwischenlager für Brennelemente) erkennen zu können, ob Neutronenstrahlung von diesen Anlagen ausgeht. Herkömmliche Ortsdosisleistungs-Messgeräte sind für diese Messaufgabe auf Grund ihrer geringen Empfindlichkeiten nicht geeignet."

    "Die Eignung des PTB-Spektrometers für die Messung von sehr kleinen Flussdichten und Dosisleistungen wurde mit Hilfe von Messungen auf dem PTB-Gelände (70 m Höhe über dem Meeresspiegel) überprüft."

    "Zwischen Dezember 1999 und Januar 2000 ergab sich eine mittlere Neutronenflussdichte von 46 cm-2 h-1. Aus der gemessenen Energieverteilung und den empfohlenen Fluenz-zu-Äquivalentdosis Konversionsfaktoren ergibt sich daraus eine mittlere Äquivalentdosisleistung von 9.3 nSv h-1 (0.081 mSv/a)".

    "Es hat sich gezeigt, dass bei sorgfältiger Charakterisierung des Spektrometers und Prüfung der Messergebnisse diese geringen Flussdichten und Dosisleistungen bestimmt werden können. Einzelheiten zu diesen Messungen finden Sie auf der weiterführenden Seite".


Human Body

You are made up of chemicals, and it should be of no surprise that some of them are radionuclides, many of which you ingest daily in your water and food. Here are the estimated concentrations of radionuclides calculated for a 70,000 gram adult based ICRP 30 data:
 
Natural Radioactivity in Your Body
Nuclide
Total Mass of Nuclide 
Found in the Body
Total Activity of Nuclide 
Found in the Body
Daily Intake of Nuclides
Uranium
90 µg
30 pCi (1.1 Bq)
1.9 µg 
Thorium
30 µg
3 pCi (0.11 Bq)
3 µg 
Potassium 40
17 mg
120 nCi (4.4 kBq)
0.39 mg 
Radium
31 pg
30 pCi (1.1 Bq)
2.3 pg 
Carbon 14
95 µg
0.4 µCi (15 kBq)
1.8 µg 
Tritium
0.06 pg
0.6 nCi (23 Bq)
0.003 pg 
Polonium
0.2 pg
1 nCi (37 Bq)
~0.6 µg 
The average annual dose equivalent from the internal deposited radionuclides is given in the entry Internal in table Radiation Doses to the US Pop.


Population Dose

Radiation Safety, IAEA (acamedia cache: 24. November 2011)

Average annual radiation exposure (millisievert), Source: Wikipedia, Background Radiation
Radiation UNSCEAR[1][2] Princeton[3] Wa State[4] MEXT[5]
Type Source World
average
Typical range USA USA Japan remark
Natural Air 1.26 0.2-10.0a 2.29 2.00 0.40 mainly from Radon, (a)depend on indoor accumulation of radon gas
Internal 0.29 0.2-1.0b 0.16 0.40 0.40 mainly from food (K-40, C-14, etc.) (b)Depend on diets
Terrestrial 0.48 0.3-1.0c 0.19 0.29 0.40 (c)depend on soil and building material
Cosmic 0.39 0.3-1.0d 0.31 0.26 0.30 (d)from sea level to high elevation
sub total 2.40 1.0-13.0 2.95 2.95 1.50
Man made Medical 0.60 0.03-2.0 3.00 0.53 2.30
Fallout 0.007 0 - 1+ - - 0.01 peak at 1963 and spike at 1986. still high near test and accident sites. US; Fallout is included in others
others 0.0052 0-20 0.25 0.13 0.001 average occupational exposure 0.7mSv, mining workers are high, population near Nuclear plant 0.02mSv
sub total 0.6 0 to tens 3.25 0.66 2.311
Total 3.00 0 to tens 6.20 3.61 3.81

figures are pre "2011 Fukushima Nuclear Plant Accident"
Human-made values at UNCEAR are from Japan NIRS which summarized UNCEAR data.


Average Annual Population Dose:


Cosmic Radiation:

Umweltradioaktivität und Strahlenbelastung
(Jahresberichte 2001 - 2009)
Bundesamt für Strahlenschutz

Siehe auch T. Hehl, "Radioaktivität in Umwelt, Technik und Medizin", Universität Tübingen, 2005 (im Cache)


 

Effektive Dosis durch ionisierende Strahlung im Jahr 1999:
Mittlere Effektive Dosis ca. 4,5 mSv
Quelle: Bundesamt für Strahlenschutz (im Cache)

Strahlungsart
Gesamtbevölkerung
Berufliche 
Strahlenexposition 
gemittelt über 
ca. 334 000 Personen
Natürliche Strahlenquellen
Kosmische Strahlung 0,3 mSv
Nahrung 0,3 mSv
Inhalation von Radon und seinen Zerfallsprodukten 1,4 mSv
Terrestrische Strahlung 0,4 mSv
Künstliche Strahlenquellen
Reaktorunfall Tschernobyl 0.02 mSv
Atombomben-Fallout 0.01 mSv
Forschung,Technik,Haushalt 0,01 mSv
Kerntechnische Anlagen 0,01 mSv 0.17 mSv
Medizin 2.00 mSv

Yearly Dose Examples
 

Tab.  III.4 S. 25 aus: Umweltradioaktivität und Strahlenbelastung im Jahr 2001 (im Cache, 24. Nov. 2011),


 

Princeton University:
Radiation Doses to the U.S. Population (NCRP 93)
 
 

Population Dose. Source: Idaho State University Source: Idaho State University


Radiation source 
Average annual whole body dose  (millirem/year)
 Natural:  Cosmic
(cosmic + terrestrial + radon  in Pittsburgh = 10 mR/a)
                             29
                 Terrestrial                              29
                 Radon                      200
Internal (K-40, C-14, etc.)                              40
Manmade:  Diagnostic x-ray                              39
                    Nuclear Medicine                              14
                    Consumer Products                              11
                    All others (fallout, air 
                    travel, occupational, etc.)
                               2
                     Average annual total
360 millirem/year


Average doses from some common activities

Activity Typical Dose
Smoking 280 millirem/year
Using radioactive materials 
 in a Princeton University lab
<10 millirem/year
Dental x-ray 10 millirem per x-ray
Chest x-ray  8 millirem per x-ray
Drinking water  5 millirem/year
Cross country round trip by air  5 millirem per trip 
Coal Burning power plant 0.165 millirem/year 


2005 Recommendations of the International Commission on Radiological Protection, Draft for Consultation

(in cache)
"Fundamental aim of the International Commission on Radiological Protection (ICRP) set out in the 1990 Recommendations (1990 Recommendations of the International Commission on Radiological Protection. ICRP Publication 60. Annals of the ICRP 21, 1-3, Elsevier, April 1991):

The primary aim of radiological protection is to
provide an appropriate standard of protection for man without unduly limiting the beneficial actions giving rise to radiation exposure.
This aim cannot be achieved on the basis of scientific concepts alone. All those concerned with radiological protection have to make value judgements about the relative importance of different kinds of risk and about the balancing of risks and benefits. In this, they are no different from those working in other fields concerned with the control of hazards."

COUNCIL DIRECTIVE 96/29/EURATOM of 13 May 1996
laying down basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionizing radiation

(in cache)
"Member States shall ensure that all new classes or types of practice resulting in exposure to ionizing radiation are
justified
in advance of being first adopted or first approved
by their economic, social or other benefits in relation to the health detriment they may cause."

Health Physics Society's Position Statement

Radiation Risk in Perspective.

(in cache)
"Epidemiological studies have not demonstrated adverse health effects in individuals exposed to small doses (less than 10 rem) delivered in a period of many years."


Roger Clarke, Control of low-level radiation exposure: time for a change?

J. Radiol. Prot. 1999 Vol. 19, No 2, 107-115 (in cache)

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
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)
    INTERNATIONAL ATOMIC ENERGY AGENCY, "Clearance levels for radionuclides in solid materials: Application of exemption principles", Interim report for comment, IAEA-TECDOC-8S5, 1996
    INTERNATIONAL GUIDANCE ON EXEMPTION AND CLEARANCE PRINCIPLES
    207. There are two main approaches that can be considered indeciding whether a level of dose is trivial:
    1. to choose a level of risk and a corresponding dose which have no significant effect as regards individuals; and
    2. to use the exposure to natural background radiation, to the extent that it is normal and unavoidable, as a relevant reference level.
    In Ref.[3] these approaches are evaluated and it is concluded that for the purpose of exemption a level of individual dose of some tens of microsieverts in a year can reasonably be regarded as trivial.

    208. Because an individual may be exposed to radiation from several exempted practices,it is necessary to ensure that the total dose does not rise above the trivial dose level. It is therefore recommended that each exempt practice should contribute only apart of the identified trivial dose [3]. The apportionment suggested could lead to individual doses to average members of the critical group of the order of 10 μSv/a from each exempt practice [3]. The value of 10 μSv/a is used in this report as the basis for evaluating unconditional clearance levels.
    (Unless otherwise stated,theterm 'dose' refers to the sum of the effective dose from external exposure in a given period and the committed effective dose from radionuclides taken into the body in the same period.)

    [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Principles for the Exemption of Radiation Sources and Practices from Regulatory Control, Safety Series No. 89, IAEA, Vienna (1988).

    vgl. auch
    Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit, "Sicherheitsanforderungen an die Endlagerung wärmeentwickelnder radioaktiver Abfälle", Stand 30. September 2010 (im Cache).

    Die Strahlenschutzverordnung enthält keine Kriterien, mit denen der Schutz zukünftiger Ge- nerationen und der Umwelt vor ionisierender Strahlung zu bewerten ist. Bei einem unter Beachtung dieser Sicherheitsanforderungen geplanten, errichteten, betriebenen und stillgelegten Endlager werden alle wesentlichen Maßnahmen getroffen, um nachfolgende Generationen und die Umwelt vor Schäden durch ionisierende Strahlen zu schützen, so dass weitergehende Nachweise von daher grundsätzlich entfallen können. International besteht Einvernehmen, dass berechnete oder abgeschätzte Risiken oder Dosen in dieser Phase nur als Indikatoren für das mit der Endlagerung zu erzielende Schutzniveau interpretiert werden dürfen. Für diese Indikatoren gelten folgende Bewertungskriterien.

    6.1 Maßgeblich für den Schutz vor Schäden durch ionisierende Strahlen in der Nachverschlussphase ist die Integrität des einschlusswirksamen Gebirgsbereichs. Die radioaktiven Abfälle müssen in diesem Gebirgsbereich so einge- schlossen sein, dass sie dort verbleiben und allenfalls geringfügige Stoffmengen diesen Gebirgsbereich verlassen können. Zusätzliche Strahlenexpositionen sollen nur in einem begrenzten Gebiet auftreten können, so dass möglichst wenige Personen einer Generation betroffen sein können.

    6.2 Für die Nachverschlussphase ist nachzuweisen, dass für wahrscheinliche Entwicklungen durch Freisetzung von Radionukliden, die aus den eingelagerten radioaktiven Abfällen stammen, für Einzelpersonen der Bevölkerung nur eine zusätzliche effektive Dosis im Bereich von 10 μSv/Jahr (in Anlehnung an ICRP 104 (triviale Dosis)) auftreten kann. Dabei sind Einzelpersonen mit einer heutigen Lebenserwartung, die während der gesamten Lebenszeit exponiert werden, zu betrachten.

    6.3 Für weniger wahrscheinliche Entwicklungen in der Nachverschlussphase ist nachzuweisen, dass die durch Freisetzung von Radionukliden, die aus den eingelagerten radioaktiven Abfällen stammen, verursachte zusätzliche effektive Dosis für die dadurch betroffenen Menschen 0,1 mSv/Jahr nicht überschreitet (in Anlehnung an ICRP 81 (Risiko kleiner 10-5/a)). Dabei sind ebenfalls Einzelpersonen mit einer heutigen Lebenserwartung, die während der gesamten Lebenszeit exponiert werden, zu betrachten.

    Für derartige Entwicklungen sind höhere Freisetzungen radioaktiver Stoffe zulässig, da das Eintreten solcher Entwicklungen eine geringere Wahrscheinlichkeit aufweist.

    Wahrscheinliche Entwicklungen sind die für diesen Standort prognostizierten normalen Entwicklungen und für vergleichbare Standorte oder ähnliche geologische Situationen normalerweise beobachtete Entwicklungen. Dabei ist für die technischen Komponenten des Endlagers die als normal prognostizierte Entwicklung ihrer Eigenschaften zugrunde zu legen. Falls eine quantitative Angabe zur Eintrittswahrscheinlichkeit einer bestimmten Entwicklung möglich ist, und ihre Eintrittswahrscheinlichkeit bezogen auf den Nachweiszeitraum mindestens 10 % beträgt, gilt diese als wahrscheinliche Entwicklung.

    Weniger wahrscheinliche Entwicklungen sind solche, die für diesen Standort unter ungünstigen geologischen oder klimatischen Annahmen eintreten können und die bei vergleichbaren Standorten oder vergleichbaren geologischen Situationen selten aufgetreten sind. Für die technischen Komponenten des Endlagers ist dabei eine als normal prognostizierte Entwicklung ihrer Eigenschaften bei Eintreten der jeweiligen geologischen Entwicklung zugrunde zu legen. Außerdem sind auch von der normalen Entwicklung abweichende ungünstige Entwicklungen der Eigenschaften der technischen Komponenten zu untersuchen. Rückwirkungen auf das geologische Umfeld sind zu betrachten. Abgesehen von diesen Rückwirkungen sind dabei die jeweilig erwarteten geologischen Entwicklungen zu berücksichtigen. Innerhalb einer derartigen Entwicklung ist das gleichzeitige Auftreten mehrerer unabhängiger Fehler nicht zu unterstellen. Falls eine quantitative Angabe zur Wahrscheinlichkeit einer bestimmten Entwicklung oder einer ungünstigen Entwicklung der Eigenschaften einer technischen Kompo-nente möglich ist, sind diese hier zu betrachten, wenn diese Wahrscheinlichkeit bezogen auf den Nachweiszeitraum mindestens 1 % beträgt.

  • Roger Clarke, Control of low-level radiation exposure: time for a change?, 1999
    Excerpts:
    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 (50 mSv, in cache) 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
    (alternative link, in cache)

    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
    was a reasonable estimate of the annual risk from a lifetime ingestion of drinking water.

    TERMS

    (alternative link, in cache)

    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).



    Radioactive Invenory of Spent Reactor Fuel, measured in SV/tHM

    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


    Health Physics Society: Population Dose 25. May 2001,

    Health Physics Society: Population Dose
    25. May 2001

    The radiation dose from inhalation of radon and thoron and their daughter products can be expressed several ways, including (a) absorbed dose to the irradiated long tissue (primarily the bronchial epithelium), (b) average absorbed dose to the whole lung, (c) dose equivalent (or equivalent dose) to the whole lung, or (d) effective dose (i.e., the average uniform whole-body dose that is estimated to represent the same risk as the actual partial-body irradiation). For this answer, we will assume that you are interested in the effective dose.

    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) đ 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 ÁSv 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 ÁSv 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 ÁSv 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. ===================


    Title 40: Protection of Environment

    PART 191 - ENVIRONMENTAL RADIATION PROTECTION STANDARDS FOR MANAGEMENT AND DISPOSAL OF SPENT NUCLEAR FUEL, HIGH-LEVEL AND TRANSURANIC RADIOACTIVE WASTES

    Subpart C - Environmental Standards for Ground-Water Protection

    § 191.13   Containment requirements.
    (a) Disposal systems for spent nuclear fuel or high-level or transuranic radioactive wastes shall be designed to provide a reasonable expectation, based upon performance assessments, that the cumulative releases of radionuclides to the accessible environment for 10,000 years after disposal from all significant processes and events that may affect the disposal system shall:

    (1) Have a likelihood of less than one chance in 10 of exceeding the quantities calculated according to Table 1 (appendix A); and

    (2) Have a likelihood of less than one chance in 1,000 of exceeding ten times the quantities calculated according to Table 1 (appendix A).

    (b) Performance assessments need not provide complete assurance that the requirements of §191.13(a) will be met. Because of the long time period involved and the nature of the events and processes of interest, there will inevitably be substantial uncertainties in projecting disposal system performance. Proof of the future performance of a disposal system is not to be had in the ordinary sense of the word in situations that deal with much shorter time frames. Instead, what is required is a reasonable expectation, on the basis of the record before the implementing agency, that compliance with §191.13 (a) will be achieved.

    Table 1—Release Limits for Containment Requirements

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

    RadionuclideRelease limit per 1,000 MTHM or other unit of waste (see notes) (curies)
    Americium-241 or -243100
    Carbon-14100
    Cesium-135 or -1371,000
    Iodine-129100
    Neptunium-237100
    Plutonium-238, -239, -240, or -242100
    Radium-226100
    Strontium-901,000
    Technetium-9910,000
    Thorium-230 or -23210
    Tin-1261,000
    Uranium-233, -234, -235, -236, or -238100
    Any other alpha-emitting radionuclide with a half-life greater than 20 years100
    Any other radionuclide with a half-life greater than 20 years that does not emit alpha particles1,000

    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 Qa, Qb, and Qc, and if the applicable Release Limits are RLa, RLb, and RLc, 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, HT, to a tissue or organ, T, listed in Table B.2 below by using the equation:



    where DT,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 wRis 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, wR1

    Radiation type and energy range2wRvalue
    Photons, all energies1
    Electrons and muons, all energies1
    Neutrons, energy < 10 keV5
    10 keV to 100 keV10
    >100 keV to 2 MeV20
    >2 MeV to 20 MeV10
    >20 MeV5
    Protons, other than recoil protons, >2 MeV5
    Alpha particles, fission fragments, heavy nuclei20

    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, wTin 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 HTis the equivalent dose in the tissue or organ, T, in the equation:



    Table B.2—Tissue Weighting Factors, wT1

    Tissue or organwTvalue
    Gonads0.25
    Breast0.15
    Red bone marrow0.12
    Lung0.12
    Thyroid0.03
    Bone surfaces0.03
    Remainder20.30

    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 HE.

    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 HT(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, wT, and then summed, the result will be the annual committed effective dose, E(τ):



    [58 FR 66415, Dec. 20, 1993]


    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

      1. the total effective dose equivalent being equal to 5 rems (0.05 Sv, in cache); or

      2. 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, in cache).

    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:

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

      2. A shallow-dose equivalent of 50 rem (0.5 Sv, in cache) 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",


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

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    Joachim Gruber