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Neutron Activation Calculator - HELP

(last updated 21 Feb 2013)

Contents:


Introduction

This calculator determines radioactive activation products of elements and nuclides exposed to neutron radiation. In addition, the calculator also performs the inverse operation, that is, potential origin nuclides can be determined for given activation products.

Depending on the elements and nuclides contained in the matter, neutron interaction can cause various types of reaction, many of which leading to the formation of radioactive activation products. Thus, matter irradiated by neutrons becomes radioactive and remains, at least for some time, radioactive even after the neutron source is shut off.

In the nuclear power industry, neutron activation is a matter of concern in reactors and in spent fuel management.
In the nuclear fuel industry, neutron activation can occur in case of criticality accidents.

The measurement of activation products in accidentally irradiated matter allows to determine the strength of the neutron exposure.
On the other hand, in case of a known neutron exposure of a sample, the same method can be used to determine concentrations of any (including non-radioactive) nuclides in the sample (Neutron Activation Analysis - NAA).

The decay database contains a total of 838 radionuclides.
The cross section database contains the thermal neutron activation cross sections of 597 nuclides.

The calculator shows all activation products that are covered by the formation rule of the respective reaction - independent of whether they really exist and whether they are easily detectable with monitoring equipment or not.
The calculator does not take into account any reduction of the irradiation efficiency from effects such as self shielding in the target, among others.

The calculator performs a complete decay analysis for the activation products and all their decay products, according to [Bateman 1910]; minor nuclides are listed at the end.
The calculator does not consider further neutron activation of activation products nor their decay products.

The results are presented in numerical form in the Results table for the irradiation period and the post-irradiation delay time specified.
The contents of the numerical result field can be marked and copied to the clipboard for further use.

The contents of the database for any element or nuclide can be checked with the "Query nuclide database" button. It shows, where available, the following data:


For space limitations, the database contains no decay energies and no cross sections for non-thermal neutrons.

This JavaScript calculator is suitable for offline use.

 

Search Mode

Select appropriate mode before any other data entry (this selection resets the complete calculator):

Nuclide Input

Forward Mode

Enter initial masses for up to 52 elements or nuclides, or select one of the pre-defined nuclide mixes from the sample data pick list.

Data import: Longer lists of input data can be imported by pasting the data to the "Results" field first, then clicking the "Import" button. For this purpose, the data must be delimited by space, comma, tab, or new lines. So, direct import from applications such as Excel is possible by copying and pasting, if the data is organized in two columns for name and mass value.
Note: make sure that you get decimal points (not commas!) from your spreadsheet software!

Original Element / Nuclide
Enter element acronym (e.g. Fe) or nuclide name (e.g. Na-23).
The name is checked with the database immediately on entry. If the element or nuclide is not found, the available nuclides resp. elements are listed.
Note: Element acronyms can be looked up with the "Query nuclide database" button.

Mass
Select unit from pick list for all mass entries in this table, and enter mass values.
If no mass value is entered for an Element/Nuclide entry, only a qualitative analysis is performed.

 

Reverse Mode

Enter names of up to 5 radioactive nuclides resulting from neutron activation
Activated Nuclide
Enter name of radioactive nuclide (e.g. Na-24).
The name is checked with the database immediately on entry. If the nuclide is not found, the available nuclides resp. elements are listed.
Note: Element acronyms can be looked up with the "Query nuclide database" button.

 

Output Parameters

Neutron flux [per cm2s], or
Point source neutron emission rate [per s] and
Distance from point source [m] (Forward mode only)
Enter either neutron flux, or emission rate of a point source and distance.
(Note: if a neutron flux is entered, then it supersedes any point source input)

Typical neutron flux values:
Neutron SourceNeutron Flux
[per cm2s]
Cosmic radiation at sea level in Germany0.0122
Alpha (Pb-210, Po-210, Ra-226, Th-228, Pu-239, or Am-241) or Gamma (Sb-124) emitters, with beryllium powder104 - 107
Cyclotron-accelerated Deuterons on H-2, H-3, or Be-9108 - 1010
Uranium reactor108 - 1016
FRM-II reactor Garching (TU München), Germany8 x 1014

Neutron emission rates:
The total number of fissions which occured during the 1999 JCO Co. criticality accident was approx. 2.5·1018. Each fission releases 2-3 neutrons, so the total number of neutrons released was approx. 6·1018. Since the criticality persisted for 20 hours, the average neutron emission rate would have been 8·1013 per s. In fact, however, a first strong peak of a few minutes was followed by a longer phase of decline.

use IAEA 2003 cross sections, where available
If checked, the cross sections are taken from the report [IAEA 2003], where available. If unchecked, or for nuclides not listed in this report, the data is extracted from [BNL 2001].
The value and origin of the cross section data actually used for a particular nuclide can be identified with the Query nuclide database button.

thermal neutrons   (n,Gamma)
fast neutrons   (n,2n)   (n,3n)   (n,p)   (n,t)   (n,Alpha)
Select neutron types of interest
Note: A quantitative analysis is only performed for thermal neutrons.

"The thermal neutron component consists of low-energy neutrons (energies below 0.5 eV) in thermal equilibrium with atoms in the reactor's moderator. At room temperature, the energy spectrum of thermal neutrons is best described by a Maxwell-Boltzmann distribution with a mean energy of 0.025 eV and a most probable velocity of 2200 m/s. In most reactor irradiation positions, 90-95% of the neutrons that bombard a sample are thermal neutrons. In general, a one-megawatt reactor has a peak thermal neutron flux of approximately 1E13 neutrons per square centimeter per second.

The epithermal neutron component consists of neutrons (energies from 0.5 eV to about 0.5 MeV) which have been only partially moderated. A cadmium foil 1 mm thick absorbs all thermal neutrons but will allow epithermal and fast neutrons above 0.5 eV in energy to pass through. In a typical unshielded reactor irradiation position, the epithermal neutron flux represents about 2% the total neutron flux. Both thermal and epithermal neutrons induce (n,gamma) reactions on target nuclei. [...]

The fast neutron component of the neutron spectrum (energies above 0.5 MeV) consists of the primary fission neutrons which still have much of their original energy following fission. Fast neutrons contribute very little to the (n,gamma) reaction, but instead induce nuclear reactions where the ejection of one or more nuclear particles - (n,p), (n,n'), and (n,2n) - are prevalent. In a typical reactor irradiation position, about 5% of the total flux consists of fast neutrons. [...]" [Glascock 2001]

Duration of irradiation (Forward mode only)
Enter time (and appropriate unit from pick list) during which the sample was subject to neutron irradiation

Time delay since end of irradiation (Forward mode only)
Enter time (and appropriate unit from pick list) passed between termination of neutron irradiation and analysis of the activation-induced activities

Max. half-life for consideration of progeny [years]
Enter an appropriate value, if the Results window is cluttered with non-relevant decay products of activated nuclides. Leave open otherwise.

Activation Product Unit (Forward mode only)
Select appropriate mass or activity unit for the activation and decay products

Naturally occuring nuclides only for origin (Reverse mode only)
Check to eliminate all artificial nuclides as origin
 

Bibliography

ORNL is the source of the nuclide data used in this calculator.

[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

[BNL 2001] PCNuDat data base at BNL.

[Glascock 2001] An Overview of Neutron Activation Analysis , by Michael D. Glascock, Missouri University Research Reactor, 2001

[IAEA 2003] Thermal Neutron Capture Cross Sections, Resonance Integrals and g-Factors , International Atomic Energy Agency - International Nuclear Data Committee, IAEA INDC(NDS)-440, February 2003.

 

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