External Radiation Dose Calculator - HELP
(Virtual Geiger Counter)
(last updated 12 Feb 2026)
Contents:
The External Radiation Dose Calculator determines the radiation dose from a shielded gamma source. The source can be a point source, or a cylindrical volume source with an evenly distributed concentration of radionuclides. The shield may consist of consecutive layers, each of which may also contain additional radionuclides.
For source and each shield layer, a number of common materials and compositions of natural radionuclides can be selected, or a custom mix of elements and radionuclides can be entered.
The receptor location can be varied in two dimensions.
Typical situations covered by the calculator:
The calculator determines the radiation dose for the actual situation, or for some later time. In the latter case, the calculator performs a complete decay analysis for the nuclides entered and all their decay products, according to [Bateman 1910]; minor nuclides are listed at the end.
The Calculator does take into account:
- Nuclide-specific Gamma radiation energies (ICRP38).
- in case of a volume source, the Gamma radiation emitted from the top of the cylinder.
- Dose contributions from the source, plus from radionuclides contained in the shield material(s), if applicable
- Self-shielding within a volume source, and within shield layers containing radioactive materials.
Note: Self-shielding can also be taken into account for sources otherwise treated as point sources, if sufficient material properties are known (density and attenuation coefficients); in this case, the effective radiation emitted from the top of a cylinder with height equaling diameter is used as the activity of the point source, and the source is labeled "Effective Point Source".
- Attenuation of Gamma radiation through a selection of common shield materials, also when arranged in multiple layers. The calculator uses the point-kernel method with buildup factors (see Calculation Details).
- Gamma attenuation in air
- Cosmic ionizing radiation, if required (calculated according to UNSCEAR 2000)
The Calculator does not take into account:
- Alpha or Beta radiation, nor secondary X-Rays (Bremsstrahlung) from shielding of Beta radiation, nor neutron radiation,
- Radiation from the side walls of the cylinder (volume source, or shields containing radionuclides),
- Radiation from radionuclides other than those for which decay energies are contained in its database,
- Gamma attenuation from shield materials not contained in its database, nor composed of elements for which shielding properties are stored,
- Cosmic neutron radiation
With these properties, the calculator is suitable for a rough assessment of the following situations, for example:
- Gamma radiation on a bare or covered uranium mill tailings pile, also with multi-layer covers (Volume Source mode)
- Gamma radiation near transport containers carrying uranium ore, U3O8, UF6, or the like (Volume Source mode)
- Gamma radiation near transport containers containing heels after unloading UF6 (Point Source mode)
- Gamma radiation from DU penetrator lying on the ground, or buried in the ground (Point, or Volume Source mode)
- Gamma radiation from storage of reprocessed uranium or depleted uranium in a warehouse (Volume Source mode)
Predefined example parameter sets covering some of these cases can be used to obtain quick first results.
References to other calculators:
The geometry of the situation and the properties of the materials involved are defined in the Input Data table.
With a HTML 5-enabled browser, a section of the geometry of the situation is shown in a Graph.
Upon execution of the calculation, the graph shows the Gamma dose rate at the selected receptor location. The receptor can easily be moved to other locations by mouse clicking.
In addition, the gamma dose can be visualized in a color map, if the Show color map box is checked.
⚠ Note 1: Each square in the color dose map represents the dose value in its center, not the value averaged over the square. Due to the sharp dose increase near point sources, the color scheme may change considerably with a decrease of the raster width, therefore.
Note 2: In logarithmic mode, the color of the square with the highest dose value is red. The span given by the selected number of decades is represented by the rainbow colors red - yellow - green - cyan - blue - violet. If the range of values is not covered by the number of decades selected, the remaining (too low) values are displayed in a linear fade-out of violet.
Computing time increases considerably with number of shields, number of elements, radionuclides per layer, Time delay > 0, and in particular with color map enabled, and then even more with small raster widths. The example parameter sets take a few seconds to compute on current machines.
Output graph (with linear color map): Click image to view animation 
!
Also upon the calculation, the layer colors change according to the following color scheme, allowing for a simple overview and for easy detection of data problems:
| Layer Color Scheme |
|---|
Point
Source | Effective
Point Source | Volume
Source | Shield | Meaning |
|---|
| dark red | n.a. | red  | pink | layer contains radionuclides, no attenuation data defined |
| n.a. | yellow | yellow | yellow | attenuation data is defined for this layer, no radionuclides contained |
| n.a. | dark orange | orange | light orange | layer contains radionuclides, plus attenuation data is defined for this layer |
| white | n.a. | white | white | no, or insufficient information defined for this layer |
indicates problem with missing data
n.a. = not applicable
The Result field repeats some important input data, reports any warnings about missing input data, and shows the calculation results for the current receptor location. The contributions from the source and from each layer containing radionuclides are shown separately and in summary.
The contents of the Result field can be highlighted and copied for further use.
Note: The figures are for the current geometry - so, if the effect of a shield is to be compared to the open source, disable the shield (select Layer OFF) and compare the results manually.
The Query Database button allows to check the contents of the calculator's database.
See instructions for offline use of this calculator.
Mode ·
Point/Volume Source Material ·
Shield #n Material ·
Cosmic Radiation ·
Geometry ·
Output ·
Element and Radionuclide Compositions and Series
Notes:
⚠ This selection must be made before any other entry, since it resets the complete calculator!
After a selection is made, the calculator can be bookmarked to obtain future direct access to the mode selected.
- Number of Shield layers
- Note: Calculation time increases with the number of shield layers.
- Point / Volume Source
- Note: Calculation time increases for volume sources.
- Example data sets
- These data sets preset the whole calculator for certain typical cases of interest. A first calculation result can then be obtained by just clicking the Calculate button. After selection, the example data can also be modified as required, to study the effect of parameter variations.
- Ex. 1: Shielding in free air (for Point Source Mode)
- Study the impacts of distance, shield thickness, and shield material. With the highest Result Detail setting, the factors affecting attenuation can be checked for each single energy emitted by the radiation source.
⚠ Note: when entering individual source radionuclides, make sure that the calculator's database actually contains the gamma energy data for these nuclides (Query Database button): for space considerations, the calculator contains energy data for a few selected nuclides only.
- Ex. 2: 48Y Cylinder with Heels from UF6_nat (for Point Source Mode)
- After unloading of a type 48Y cylinder containing 12,500 kg of natural uranium hexafluoride (UF6) by heating in an autoclave, the decay products of the uranium remain in the cylinder as so-called Heels. See, how an "empty" cylinder emits much more gamma radiation than a full one.
Note 1: This data set conservatively assumes that no residual UF6 remains in the cylinder.
Note 2: In addition, the cylinder emits Bremsstrahlung which is not covered by this calculator.
- Ex. 3: 30B Cylinder with Heels from UF6_enr (for Point Source Mode)
- After unloading of a type 30B cylinder containing 2,277 kg of enriched (3.5 wt% U-235) uranium hexafluoride (UF6) by heating in an autoclave, the decay products of the uranium remain in the cylinder as so-called Heels. See, how an "empty" cylinder emits much more gamma radiation than a full one.
Note 1: This data set conservatively assumes that no residual UF6 remains in the cylinder.
Note 2: In addition, the cylinder emits Bremsstrahlung which is not covered by this calculator.
- Ex. 4: DU bullet buried in soil (for Point Source Mode)
- Depleted uranium bullets buried in soil are difficult to locate by their gamma radiation emission.
- Ex. 5: Soil cover 40 CFR 192 (for Volume Source Mode)
- Gamma radiation from soil containing 15 pCi/g [0.555 Bq/g] Ra-226, covered by 15 cm of soil containing 5 pCi/g [0.185 Bq/g] Ra-226. These parameters correspond to the EPA standard 40 CFR 192 for residual contamination of soil at reclaimed uranium mill sites.
Note 1: This data set assumes that Ra-226 is in secular equilibrium with its parent and decay products.
- Ex. 6: Soil cover for uranium mill tailings (for Volume Source Mode)
- Gamma radiation from uranium mill tailings can be reduced by soil covers, often applied in several layers. Check the effects of layer material, cover thickness and of residual radioactivity contained in the cover material.
- Ex. 7: 200L Drum with uranium ore concentrate (UOC) (for Volume Source Mode)
- Determine the gamma dose rate outside the top of a typical 200 liter transport drum for Uranium Ore Concentrate.
- Ex. 8: 48Y Cylinder with UF6_nat (for Volume Source Mode)
- Determine the gamma dose rate outside the top of a typical transport cylinder for natural uranium hexafluoride.
⚠ Note 1: This data set assumes that the cylinder is completely filled with solid UF6, which it normally isn't. Modify the Source density rhoso to 2.897 g/cm3 to account for the regular partial filling state.
Note 2: In addition, the cylinder emits Bremsstrahlung and neutron radiation which are not covered by this calculator.
- Ex. 9: 30B Cylinder with UF6_enr (for Volume Source Mode)
- Determine the gamma dose rate outside the top of a typical transport cylinder for enriched uranium hexafluoride.
⚠ Note 1: This data set assumes that the cylinder is completely filled with solid UF6, which it normally isn't. Modify the Source density rhoso to 2.76 g/cm3 to account for the regular partial filling state.
Note 2: In addition, the cylinder emits Bremsstrahlung and neutron radiation which are not covered by this calculator.
- Ex. 10: Reprocessed uranium storage (for Volume Source Mode)
- Determine the gamma dose rate from reprocessed uranium stored in a warehouse as bulk U3O8 behind a shield of depleted uranium in the form of bulk U3O8.
Note 1: Shielding by container walls neglected.
Note 2: In addition, Bremsstrahlung and neutron radiation are emitted which are not covered by this calculator.
- Ex. 11: Depleted uranium storage (for Volume Source Mode)
- Determine the gamma dose rate from depleted uranium stored in a warehouse as bulk U3O8 behind a shield of ordinary concrete.
Note 1: Shielding by container wall neglected.
Note 2: In addition, Bremsstrahlung and neutron radiation are emitted which are not covered by this calculator.
Note: The dose color map configuration is initialized individually for each example data set, in order to obtain reasonable calculation times.
- Layer usage selector
- This drop down list allows to easily disable parts or all of the properties of the source layer for test purposes:
| NORMAL |
the layer is fully operational |
| RAD. ONLY |
the attenuation properties of the layer are set to vacuum |
| ATTEN. ONLY |
any radiation emission from the layer is disabled |
| Layer VOID |
the attenuation properties of the layer are set to vacuum and any radiation emission from the layer is disabled |
| Layer OFF |
the layer is completely removed |
- Total amount / Individual amounts below (Point Source only)
- Choose, whether the composition of the source material is to be entered as total mass plus activity or mass concentrations, or as individual activities and masses.
- Total amount: Mass figure (Point Source only)
- Enter number
- Total amount: Mass Unit (Point Source only)
- Select unit from pick list
- Source Material
- Material data for the radiation source
Select sample material from the pick list (and modify any entries, if required), or enter a chemical formula below, or enter new data in the table.
The pick list contains some elemental compositions, as well as radionuclide compositions and radionuclide series.
Note: In the Element/Nuclide column, only Elements, Nuclides, and radionuclide compositions and radionuclide series are allowed.
Note: The decay energy and attenuation data can be viewed with the "Query database" button.
For any elemental compositions not contained in the pick list, a chemical formula can be entered according to the scheme shown in the following examples. In addition to the molecular information, the scheme includes also the radiation properties of the material:
| Material | Original formula *) | Entered formula | Comments |
|---|
| Natural U3O8, without progeny | U3O8 | U3O8 | equivalent to: U_nat3O8 |
| Natural U3O8, with short-lived progeny | U3O8 | U_nat+3O8 | |
| Natural U3O8, with all progeny | U3O8 | U_nat++3O8 | |
| Natural UF4, with short-lived progeny | UF4 | U_nat+F4 | |
| Enriched uranyl fluoride, with short-lived progeny | UO2F2 | U_enr+O2F2 | |
| Enriched uranium trioxide, with short-lived progeny | UO3 | U_enr+O3 | |
Recycled uranyl nitrate hexahydrate (UNH),
with short-lived progeny | UO2(NO3)2 · 6 H2O | U_rec+O2(NO3)2.6H2O | |
| Plutonium-238 dioxide | 238PuO2 | Pu-238O2 | |
*) Elements shown in red, if their radionuclide composition differs from natural, or no natural abundance is defined
Note 1: Check Radionuclide Compositions and Radionuclide Series for accepted symbols and details.
Note 2: Nested parentheses ((...)), points (.) replacing middle dots (·), and additional spaces are allowed, where applicable.
⚠ Note 3: Isotopes must be entered e.g. as U-238; use parentheses, where ambiguity could arise.
⚠ Note 4: Formula entry is case sensitive!
⚠ Note 5: The Source density must be entered explicitly, when using chemical formulas -- other than for the pick list data sets!
- Consider self-shielding of point source with: rhoso - Source density [g/cm3] (Point Source only)
Check box and enter density value to take self-shielding within the point source into account: if attenuation coefficients are available for the source material, then the activity of the point source is taken as the activity at the top surface of an upright cylinder with diameter equalling height, and the source is labeled "Effective Point Source".
- rhoso - Source density [g/cm3] (Volume Source only)
Value must be larger than zero.
Note: For granular materials, this is the density of the bulk material, not the density of the grains.
- Element / Nuclide
- Enter short names of elements (e.g. Th) or radionuclides (e.g. U-238) and associated concentrations in weight-percent, or, for radionuclides, alternatively the activity concentration in Bq per gram of source material. For point sources, there is also the possibility to enter individual masses in grams and activities in Bq directly, if the box "Individual amounts below" is checked.
In addition, the short names of a number of pre-defined radionuclide compositions (e.g. U_nat) and radionuclide series (e.g. U-238++) can be entered.
Note 1: For elements, the natural isotopic abundance of the element is assumed, if there is one defined (check with the "Query database" button) - otherwise enter individual nuclides.
Note 2: For radionuclides, check the availability of the associated decay data with the "Query database" button.
⚠ Note 3: Mass and activity concentrations entered for a radionuclide composition refer to the total mass/activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++), while those entered for a radionuclide series refer to the mass/activity of the nominal nuclide only (e.g. U-238 for U-238++).
Data import: Longer lists of input data can be imported by pasting the data to the input field first, then clicking the "IMPORT" button next to it. For this purpose, the data must be delimited by space, comma, tab, or new lines, and "wt_%" must be encoded as 0, "Bq/g" as 1. So, direct import from spreadsheet applications such as Excel is possible by copying and pasting, if the data is organized in three columns for name, value, and unit.
⚠ Note: make sure that you get decimal points (not decimal commas!) from your spreadsheet software!
Shield #n Material Composition
- Layer usage selector
- This drop down list allows to easily disable parts or all of the properties of each shield layer for test purposes:
| NORMAL |
the layer is fully operational |
| RAD. ONLY |
the attenuation properties of the layer are set to vacuum |
| ATTEN. ONLY |
any radiation emission from the layer is disabled |
| Layer VOID |
the attenuation properties of the layer are set to vacuum and any radiation emission from the layer is disabled |
| Layer OFF |
the layer is completely removed |
- Shield Material
- Material data for each shield
Select sample material from the pick list (and modify any entries, if required), or enter a chemical formula below, or enter new data in the table.
The pick list contains some elemental compositions, as well as radionuclide compositions and radionuclide series.
Note: In the Element/Nuclide column, only Elements, Nuclides, and radionuclide compositions and radionuclide series are allowed.
Note: The decay energy and attenuation data can be viewed with the "Query database" button.
For any elemental compositions not contained in the pick list, a chemical formula can be entered according to the scheme shown in the following examples. In addition to the molecular information, the scheme includes also the radiation properties of the material:
| Material | Original formula *) | Entered formula | Comments |
|---|
| Polypropylene | (C3H6)n | C3H6 | |
| Polyvinyl chloride | (C2H3Cl)n | C2H3Cl | |
| Gypsum | CaSO4·2H2O | CaSO4.2H2O | |
| Barium sulfate | BaSO4 | BaSO4 | |
| ZnO-PbO-B2O3 glass | xZnO · 2xPbO · (1-3x)B2O3 | 0.23ZnO.0.46PbO.0.31B2O3 | e.g. for x=0.23 |
| Depleted U3O8, with short-lived progeny | U3O8 | U_dep+3O8 | |
*) Elements shown in red, if their radionuclide composition differs from natural, or no natural abundance is defined
Note 1: Check Radionuclide Compositions and Radionuclide Series for accepted symbols and details.
Note 2: Nested parentheses ((...)), points (.) replacing middle dots (·), and additional spaces are allowed, where applicable.
⚠ Note 3: Isotopes must be entered e.g. as U-238, while 238U (for 238U) is not allowed, to avoid ambiguity.
⚠ Note 4: Formula entry is case sensitive!
⚠ Note 5: The Source density must be entered explicitly, when using chemical formulas -- other than for the pick list data sets!
- rhoshn - Shield density [g/cm3]
Value must be larger than zero, for the shield to be effective.
Note 1: If no value, or zero, is entered, this layer is treated as vacuum.
Note 2: For granular materials, this is the density of the bulk material, not the density of the grains.
- Element / Nuclide [wt_% / Bq/g]
- Enter short names of elements (e.g. Pb) or radionuclides (e.g. U-238) and associated abundance in weight-percent, or, for radionuclides, alternatively the activity concentration in Bq per gram of source material. In addition, the short names of a number of pre-defined radionuclide compositions (e.g. U_dep) and radionuclide series (e.g. U-238++) can be entered.
Note 1: For elements, the natural isotopic abundance of the element is assumed, if there is one defined (check with the "Query database" button) - otherwise enter individual nuclides.
Note 2: For radionuclides, check the availability of the associated decay data with the "Query database" button.
⚠ Note 3: Mass and activity concentrations entered for a radionuclide composition refer to the total mass/activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++), while those entered for a radionuclide series refer to the mass/activity of the nominal nuclide only (e.g. U-238 for U-238++).
Data import: Longer lists of input data can be imported by pasting the data to the input field first, then clicking the "IMPORT" button next to it. For this purpose, the data must be delimited by space, comma, tab, or new lines, and "wt_%" must be encoded as 0, "Bq/g" as 1. So, direct import from spreadsheet applications such as Excel is possible by copying and pasting, if the data is organized in three columns for name, value, and unit.
⚠ Note: make sure that you get decimal points (not commas!) from your spreadsheet software!
- Layer usage selector
- This drop down list allows to enable/disable the inclusion of cosmic radiation to the dose calculations:
| NORMAL |
cosmic radiation is considered |
| Layer OFF |
cosmic radiation is neglected |
- Altitude [m above sea level]
- Cosmic radiation increases with altitude.
(Note: for terrestrial use only)
- Latitude [°]
- Cosmic radiation is slightly lower at latitudes below 30°.
- Outdoor/Indoor
- The ceiling of a building provides some shielding from cosmic radiation.
- Indoor: Shielding factor
- Dimensionless ratio of indoor to outdoor radiation level from ionizing cosmic radiation.
Point Source Geometry: (2 shield layers)
Volume Source Geometry: (2 shield layers)
The geometry parameters can be initialized with predefined data set examples, corresponding to the example buttons in the Mode section.
Note: While the buttons in the Mode section initialize all parameters, here only the geometry is affected.
- Ex. 1: Shielding in free air (for Point Source Mode)
- Ex. 2: 48Y Cylinder (for Point Source Mode)
- Ex. 3: 30B Cylinder (for Point Source Mode)
- Ex. 4: Source buried in soil (for Point Source Mode)
- Ex. 5: Soil cover 40 CFR 192 (for Volume Source Mode)
- Ex. 6: Soil cover for tailings (for Volume Source Mode)
- Ex. 7: Drum with uranium ore concentrate (UOC) (for Volume Source Mode)
- Ex. 8: 48Y Cylinder (for Volume Source Mode)
- Ex. 9: 30B Cylinder (for Volume Source Mode)
- Ex. 10: Reprocessed uranium storage (for Volume Source Mode)
- Ex. 11: Depleted uranium storage (for Volume Source Mode)
- a - Distance of source from shield front surface [cm] (Point Source only)
If no value is entered, zero is assumed.
If a negative value is entered, the point source is located inside the shield (Shield #1 only).
- x - Horizontal displacement of receptor [cm]
If no value is entered, zero is assumed. For volume sources or shields containing radionuclides, the horizontal displacement is limited to the source or shield radius, respectively.
- y - Distance of receptor from x-axis [cm] - or -
b - Distance of receptor from shield rear surface [cm]
Enter either y or b.
Upon entry of y, b is erased, and vice versa.
Note: The x-axis runs through the point source, the top of the effective point source, or the top of the volume source, respectively.
- d - Source depth [cm] (Volume Source only)
Value must be larger than zero.
- dsn - Shield #n thickness [cm]
If no value is entered, zero is assumed.
- r - Source radius [cm] - or -
sa - Source surface area [m2] (Volume Source only)
Enter either r or sa. The value must be larger than zero.
Upon entry of r, sa is calculated automatically, and vice versa.
- rs - Shield radius [cm] - or -
sas - Shield surface area [m2]
Enter either rs or sas. The value must be equal or larger than the source radius.
Upon entry of rs, sas is calculated automatically, and vice versa.
(In the case of a point source, this parameter is only required if the shield contains radionuclides)
- integration step width [cm]
- Maximum horizontal size of the section of the volume elements used for the point-kernel method.
Note: smaller sizes improve the precision, but increase computing times and memory requirements considerably
(In the case of a point source, this parameter is only required if the shield contains radionuclides)
- integration step height [cm]
- Maximum vertical size of the section of the volume elements used for the point-kernel method.
Note: smaller sizes improve the precision, but increase computing times and memory requirements considerably
(In the case of a point source, this parameter is only required if the shield contains radionuclides)
Decay Options
- Time delay for dose assessment
- Enter number and select unit for time passed, until the dose assessment is performed: this allows to examine the effect of radioactive decay in the nuclide composition. Leave open, or enter 0, if no time delay is requested.
(time units: s=seconds, m=minutes, h=hours, d=days, a=years)
- Min. branching for progeny [%]
- Enter branching percentage, below which branching in the decay chain will be truncated: this allows to eliminate less relevant nuclides. Leave open, or enter 0, if no truncation of branching is requested.
- Max. half life for short-lived progeny:
- Enter number and select unit for the maximum half life, beyond which the decay chain will be truncated for short-lived progeny (+).
(time units: s=seconds, m=minutes, h=hours, d=days, a=years)
Note: The half life definitions for "short-lived" vary widely:
- ~15 h - Medical imaging
- ~10 d - Medical treatment
- 10 d - IAEA transport regulations
- 100 d - European Nuclear Society (ENS)
- 5 a - French Centre National de la Recherche Scientifique (CNRS)
- 31 a - French National Radioactive Waste Management Agency (Andra)
- 40 a - Australian Radioactive Waste Agency (ARWA)
- 100e6 a - Cosmology
Dose Options
- Dose Rate Unit
- Select from pick list
Note: The primary unit calculated is Gy/h. All other units are derived from this one.
- Exposure for annual dose rates
- Select occupancy form pick list
- Receptor Material
- Select from pick list
Note: The absorbed gamma energy dose usually is calculated for air, and any further dose figures are derived from this value.
- Terrestrial gamma dose coeff. in air [Sv/Gy]
- Conversion coefficient from absorbed dose in air to effective dose equivalent for terrestrial gamma rays.
UNSCEAR (2000) recommends 0.7 Sv/Gy for adults, 0.8 for children, and 0.9 for infants.
Note 1: this coefficient is used energy-independently
Note 2: this coefficient is only used for receptor air, otherwise unity is used
Note 3: the coefficient used for the contribution from cosmic ionizing radiation is unity
- Use buildup factors for:
- Check to enable computing of buildup factors to compensate for the non-ideal geometry of the situation.
Note: although the use of buildup factors increases computing time, unchecking is not advisable, except for test purposes.
For each layer, the material to be used for the buildup factor calculations can be selected manually from the respective drop-down list, or "auto select" can be chosen.
Note: For "auto select", the buildup data for the layer material is used, if available; otherwise the buildup data for an estimated effective atomic number of the layer material is used.
Result Detail
- Dose from each layer
- Select desired level of detail for the dose summary in the Result field.
(The selections "by Radiations" are available in Point Source mode only and affect the contributions from the point source itself only)
Graph Detail
- Show total situation
- Check to show total view of the situation defined by the Geometry Parameters, in case outer lateral parts of volume source or shields are cut off.
- Show integration grid
- Check to show the grid defined by the integration step width/height entered under Geometry Parameters
- Show dose color map
- Check to calculate dose values for the following raster of x and y positions and show the result as a coloured map.
⚠ Note: If this checkbox is checked, computing time increases considerably, in particular for volume sources and/or shields containing radionuclides!
- Raster width [pixel]
- Horizontal and vertical raster width in pixels for calculation of dose values in color map
⚠ Note: Small raster widths increase computing time considerably!
- Logarithmic color scale
- Check for logarithmic color scale, otherwise a linear scale is used
- Decades
- Number of decades to be covered by color map in logarithmic mode.
If no value, or zero, is entered, an appropriate scale is selected automatically.
Note: If, for logarithmic scale, the range of values is too large for the number of decades selected, the remaining (too low) values are displayed in a linear fade-out of violet.
Element Compositions ·
Radionuclide Compositions ·
Radionuclide Series
| Element Compositions |
|---|
| Name | Description |
|---|
| Air | Air, Dry (Near Sea Level) |
| Water | Water, Liquid |
| Brick_cs | Bricks, Common Silica |
| Concr_ord | Concrete, Ordinary |
| Concr_BA | Concrete, Barite (Type BA) |
| Wood_SP | Wood, Southern Pine |
| Glass_Pb | Glass, Lead |
| Glass_BS | Glass, Borosilicate ("Pyrex") |
| Glass_plt | Glass, Plate |
| Glass_Unat | Uranium glass with 0.373% U, natural, with short-lived progeny |
| Glass_Udep | Uranium glass with 0.373% U, depleted to 0.2% U-235, with short-lived progeny |
| Glass_acr | Glass, Acrylic ("Lucite") |
| Tiss_sft | Tissue, Soft (ICRU-44) |
| HDPE | High Density Polyethylene (HDPE) |
| St_304 | Stainless Steel (Type 304) |
| Soil_US | U.S. Soil |
| Soil_05 | U.S. Soil with Ra-226 @ 5 pCi/g = 0.185 Bq/g (U-series in equil.) a) |
| Soil_15 | U.S. Soil with Ra-226 @ 15 pCi/g = 0.555 Bq/g (U-series in equil.) a) |
| Rock_cru | Rock, Crustal |
| Salt_rck | Salt, Rock |
| Asph_05 | Asphalt mixture (5% bitumen, 95% crustal rock) |
| Uore_dgo | Uranium ore 0.33 wt-% U, Durango, Colorado, USA c) |
| Utail_dgo | Uranium mill tailings, Durango, Colorado, USA |
| Uore_01 | Uranium ore 0.1 wt-% U b) |
| Utail_01 | Uranium mill tailings from 0.1 wt-% U ore, extraction = 90% b) |
| Uore_nor | Uranium ore 0.066 wt-% U, Nordic Lake, Elliot Lake, Ontario, Canada b) |
| Utail_nor | Uranium mill tailings, Nordic Lake, Elliot Lake, Ontario, Canada |
| UF6_nat+ | Uranium hexafluoride, natural, solid, with short-lived progeny |
| UF6_rec+ | Uranium hexafluoride, recycled uranium, solid, init. enr. 3.5 wt-%, burnup 39 GWd/tHM, 5 y delay, with progeny |
| UF6_enr+ | Uranium hexafluoride, enriched to 3.5 wt-% U-235, solid, from natural uranium, with short-lived progeny |
| UF6_ere+ | Uranium hexafluoride, enriched to 3.5 wt-% U-235 equiv., solid, from recycled U (3.5 wt-% init.enr. 39 GWd/tHM, 5 y), with short-lived progeny |
| UF6_dep+ | Uranium hexafluoride, depleted to 0.2 wt-% U-235, solid, from natural uranium, with short-lived progeny |
| UF6_dre+ | Uranium hexafluoride, depleted to 0.2 wt-% U-235, solid, from recycled U (3.5 wt-% init.enr. 39 GWd/tHM, 5 y), with short-lived progeny |
| U3O8_nat+ | U3O8, natural, with short-lived progeny, bulk |
| U3O8_nat++ | U3O8, natural, with all major progeny, bulk |
| SPF_33y0 | Spent fuel, burnup 33 GWd/tHM, at unload |
| U3O8_rec+ | U3O8, recycled uranium, init. enr. 3.5 wt-%, burnup 39 GWd/tHM, 5 y delay, with progeny, bulk |
| U3O8_dep+ | U3O8, depleted to 0.2 wt-% U-235, from natural uranium, with short-lived progeny, bulk |
| U3O8_dep++ | U3O8, depleted to 0.2 wt-% U-235, from natural uranium, with all major progeny grown in (takes a while...), bulk |
| UO2_nat+ | UO2, natural, with short-lived progeny |
| UO2_enr+ | UO2, enriched to 3.5 wt-% U-235, from natural uranium, with short-lived progeny |
| UO2_dep+ | UO2, depleted to 0.2 wt-% U-235, from natural uranium, with short-lived progeny |
| UO2_ere+ | UO2, enriched to 3.5 wt-% U-235 equiv., from recycled U (3.5 wt-% init.enr. 39 GWd/tHM, 5 y), with short-lived progeny |
| Heels_nat+ | Heels from sublimation of natural uranium hexafluoride, radionuclides only |
| Heels_enr+ | Heels from sublimation of enriched uranium hexafluoride (3.5 wt-% U-235), radionuclides only |
| UNH_rec+ | UO2(NO3)2 · 6 H2O, uranyl nitrate hexahydrate, recycled uranium, init. enr. 3.5 wt-%, burnup 39 GWd/tHM, 5 y delay, with progeny, bulk |
| 238PuO2 | 238PuO2: Plutonium-238 dioxide |
| 241AmO2 | 241AmO2: Americium-241 dioxide |
a) based on Soil_US, density and/or radionuclides modified
b) based on Utail_nor, density and/or radionuclides modified
c) based on Utail_dgo, density and/or radionuclides modified
⚠ Note 1: Mass and activity concentrations entered for an element composition refer to the total mass/activity of all nuclides contained, except for those only contained in the decay series indicated by "+".
Note 2: For short-lived progeny (+), the decay chain is truncated, once a nuclide with a half life higher than the Max. half life for short-lived progeny specified is encountered.
Note 3: Decay branches are only considered, if they are higher than the Min. branching for progeny specified.
| Radionuclide Compositions |
|---|
| Name | Description |
|---|
| U_nat | Natural Uranium, without progeny |
| U_nat+ | Natural Uranium, with short-lived progeny |
| U_nat++ | Natural Uranium, with all major progeny in sec. equilibrium |
| U_tailx90++ | Uranium in mill tailings, extraction = 90%, with all major progeny @ 1 year |
| U_rec | Recycled Uranium, init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay |
| U_rec+ | Recycled Uranium, init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay, with progeny |
| U_dep | Depleted Uranium, 0.2 wt-% U-235, without progeny |
| U_dep+ | Depleted Uranium, 0.2 wt-% U-235, with short-lived progeny |
| U_dre | Depleted Recycled Uranium, 0.2 wt-% U-235, init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay |
| U_dre+ | Depleted Recycled Uranium, 0.2 wt-% U-235, init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay, with short-lived progeny |
| U_enr | Enriched Uranium, 3.5 wt-% U-235, without progeny |
| U_enr+ | Enriched Uranium, 3.5 wt-% U-235, with short-lived progeny |
| U_ere | Enriched Recycled Uranium, 3.5 wt-% U-235 equiv., init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay, without progeny |
| U_ere+ | Enriched Recycled Uranium, 3.5 wt-% U-235 equiv., init. enr. 3.5 wt-% U-235, burnup 39 GWd/tHM, 5 y delay, with short-lived progeny |
⚠ Note 1: Mass and activity concentrations entered for a radionuclide composition refer to the total activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++).
Note 2: Radionuclide compositions are calculated with the material properties of the nominal element.
Note 3: For short-lived progeny (+), the decay chain is truncated, once a nuclide with a half life higher than the Max. half life for short-lived progeny specified is encountered.
Note 4: Decay branches are only considered, if they are higher than the Min. branching for progeny specified.
| Radionuclide Series |
|---|
| Name | Description |
|---|
| Definition |
|---|
| Nn-XXX+ | includes only short-lived progeny |
| Nn-XXX++ | includes progeny with any half lives |
| Examples |
|---|
| U-238+ | Uranium-238, with short-lived progeny |
| U-238++ | Uranium-238, with all progeny in secular equilibrium |
⚠ Note 1: Mass and activity concentrations entered for a radionuclide series refer to the activity of the nominal nuclide only (e.g. U-238 for U-238++).
Note 2: Radionuclide series are calculated with the material properties of the nominal nuclide.
Note 3: For short-lived progeny (+), the decay chain is truncated, once a nuclide with a half life higher than the Max. half life for short-lived progeny specified is encountered.
Note 4: Decay chain branches are only considered, if they are higher than the Min. branching for progeny specified.
The calculator uses the point-kernel method with buildup factors.
- For point sources, the attenuation of a radiation beam is calculated, as it passes through the shield (which may be composed of consecutive shield layers). For each decay energy of each radionuclide in the source, the energy-specific attenuation on the beam's path through the shield is taken into account. For sections of the path outside of the shield, attenuation by air is assumed.
If no attenuation data is stored for a shield layer material, it is calculated from the layer's elemental composition, if given, and if attenuation data is stored for the relevant elements.
Shields containing radionuclides are treated as additional volume sources (see below).
To compensate for the non-ideal geometry of the situation, buildup-factors are used for the shield layers. In case the material of a shield layer consists of more than one element, and buildup factors for the material are not available, an energy-specific effective atomic number (Zeff) is calculated for the layer material according to [Singh 2003], which is then used to determine the buildup factors. In case the buildup factors for a layer element (real or effective) are not available, the buildup factors are interpolated between those known for the elements with the nearest atomic numbers. For multi-layer shields, buildup factors are calculated according to the empirical formula provided in [Broder 1963].
- For volume sources, the cylindrical source is divided into small volume elements in the form of rings with a rectangular section. This allows to handle the three-dimensional problem in two dimensions to save computing time. For each volume element, the attenuation of a radiation beam is calculated, as it passes through the rest of the source and through the shield (which may be composed of consecutive shield layers). For each decay energy of each radionuclide in the source, the energy-specific attenuation on the beam's path is taken into account. For sections of the path above the shield, attenuation by air is assumed.
If no attenuation data is stored for a layer material, it is calculated from the layer's elemental composition, if given, and if attenuation data is stored for the relevant elements.
If the receptor is horizontally displaced away from the common axis of the source and shield cylinders, the volume rings follow the displacement, and only their intersections with the source/shield cylinders are taken into account.
Shields containing radionuclides are treated as additional volume sources.
To compensate for the non-ideal geometry of the situation, buildup-factors are used for each layer. In case the material of a source or shield layer consists of more than one element, and buildup factors for the material are not available, an energy-specific effective atomic number (Zeff) is calculated for the layer material according to [Singh 2003], which is then used to determine the buildup factors. In case the buildup factors for a layer element (real or effective) are not available, the buildup factors are interpolated between those known for the elements with the nearest atomic numbers. For multi-layer situations (i.e. source plus at least one shield), buildup factors are calculated according to the empirical formula provided in [Broder 1963].
- [Bateman 1910] Harry Bateman: Solution of a system of differential equations occurring in the theory of radioactive transformations
, in: Proceedings of the Cambridge Philosophical Society, Mathematical and physical sciences. Cambridge [etc.] Cambridge Philosophical Society. v. 15 (1908-10): Pages 423-427
- [Broder 1963] The transmission of gamma-radiation through heterogeneous media, by D. L. Broder, Yu. P. Kayurin, A. A. Kutuzov, in: Problems in reactor shielding physics, Collection of articles , D. L. Broder, et al., Editors, Gosatomizdat, Moscow, 1963, NASA Technical Translation TT F-411, Washington, D.C., June 1967 (26.3 MB PDF), p. 287-299
- [Parks 1988] Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications , by C. V. Parks, B. L. Broadhead, O. W. Hermann, et al., Oak Ridge National Laboratory, ORNL/CSD/TM-246, July 1988
- [Singh 2003] ZnO-PbO-B2O3 glasses as gamma-ray shielding materials, by H. Singh, K. Singh, L. Gerward, et al., in: Nuclear Instruments and Methods in Physics Research B, Vol. 207 (2003), p. 257–262
- [UNSCEAR 2000] Sources and Effects of Ionizing Radiation, UNSCEAR 2000 Report to the General Assembly, with Scientific Annexes, United Nations Scientific Committee on the Effects of Atomic Radiation, United Nations, New York, 2000
> Download full text: Vol. I: Sources · Vol. II: Effects
- Source for the decay data:
ICRP-07 Data Files © A. Endo and K.F. Eckerman 2008; decay energies: Oak Ridge National Laboratory (ORNL): ICRP38 program
- Source for the photon mass attenuation and mass energy-absorption coefficients:
- Source for the Buildup Factors:
New gamma-ray buildup factor data for point kernel calculations: ANS-6.4.3 standard reference data , by D. K. Trubey, NUREG/CR-5740, U.S. Nuclear Regulatory Commission, ORNL/RSIC-49, Oak Ridge National Laboratory, September 1988 (3.1MB PDF)
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