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Neutron Activation Calculator - HELP

(last updated 5 Jul 2004)

Contents:

Introduction · Search Mode · Nuclide Input · Output Parameters · Bibliography

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).

In the Forward mode, initial masses of up to 52 elements or nuclides can be entered in the Nuclide Input table. In case of entry of an element rather than a nuclide, the natural isotope abundance of the element is used for the analysis.
The nuclides entered may be stable or radioactive. Any decay products growing in from the entered nuclides are not considered, however.
Possible neutron capture activation products are determined for each nuclide of the origin material, plus any decay products growing in from the activation products.
The neutron capture reactions covered are:
- (n,Gamma) - thermal neutron capture results in release of gamma radiation
- (n,2n) - fast neutron capture results in release of 2 neutrons
- (n,3n) - fast neutron capture results in release of 3 neutrons
- (n,p) - fast neutron capture results in release of a proton
- (n,t) - fast neutron capture results in release of a triton
- (n,Alpha) - fast neutron capture results in release of an alpha particle
Only radioactive end products of neutron capture are determined.
For thermal neutrons, the activities of the activation products and their decay products are calculated, while, for fast neutrons, no activities are calculated (since the calculator's cross section database only covers thermal neutrons).
In the Reverse mode, the names of up to 5 activation and/or decay products can be entered in the Nuclide Input table. All possible origin nuclides are determined which may result in the listed activation and decay products. In this mode, no masses are calculated; however, calculated origin nuclides of interest can be entered in the Forward mode to perform mass calculations.

The decay database contains a total of 497 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 performs a complete decay analysis for the activation products and all their decay products, except in case of branching in the forward mode, where only the decay with the highest probability is considered.
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 elements: name, acronym, atomic number, isotopic abundance, list of radionuclides in database,
for radionuclides: half-life, specific activity, possible parent nuclides, decay products with branching ratios and decay type.
Note that stable end-products of decay series are not listed, and note that only alpha and beta decays are listed: in addition, each decay emits gamma radiation.
for any nuclides: thermal neutron activation cross sections (1 barn = 10^-24 cm²).

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):

Forward (Search activation products for given origin material)
Reverse (Search original nuclides for given activation products)

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.

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 Source Neutron Flux
[per cm²s]

Cosmic radiation at sea level in Germany 0.0122

Alpha (Pb-210, Po-210, Ra-226, Th-228, Pu-239, or Am-241) or Gamma (Sb-124) emitters, with beryllium powder 10⁴ - 10⁷

Cyclotron-accelerated Deuterons on H-2, H-3, or Be-9 10⁸ - 10¹⁰

Uranium reactor 10⁸ - 10¹⁶

FRM-II reactor Garching (TU München), Germany 8 x 10¹⁴

Neutron Source	Neutron Flux [per cm²s]
Cosmic radiation at sea level in Germany	0.0122
Alpha (Pb-210, Po-210, Ra-226, Th-228, Pu-239, or Am-241) or Gamma (Sb-124) emitters, with beryllium powder	10⁴ - 10⁷
Cyclotron-accelerated Deuterons on H-2, H-3, or Be-9	10⁸ - 10¹⁰
Uranium reactor	10⁸ - 10¹⁶
FRM-II reactor Garching (TU München), Germany	8 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.

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

RSIC

at ORNL is the original source of the nuclide data used in this calculator, and Grove Engineering are the originators of the microcomputer formatted files.

The cross sections are extracted from the NuDat data base at BNL.

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

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