ASTM E3063-24

Standard Test Method for Antimony Content Using Neutron Activation Analysis (NAA)

SIGNIFICANCE AND USE
5.1 High levels of antimony are commonly used in flame retardant formulations for various materials. NAA is a test method that can be useful for verifying these levels and, for other materials, NAA can also be useful in establishing the amount of low level contamination, if any, with high sensitivity and high precision.  
5.2 Neutron activation analysis provides a rapid, highly sensitive, nondestructive procedure for antimony determination in a wide range of matrices. This test method is independent of the chemical form of the antimony.  
5.3 This test method can be used for quality and process control in the petrochemical and other manufacturing industries, and for research purposes in a broad spectrum of applications.
SCOPE
1.1 This test method covers the measurement of antimony concentration in plastics or other hydrocarbon or organic matrix by using neutron activation analysis (NAA). The sample is activated by irradiation with neutrons from a research reactor and the subsequently emitted gamma-rays are detected with a germanium semiconductor detector. The same system may be used to determine antimony concentrations ranging from 1 ng/g to 10 000 μg/g with the lower end of the range limited by numerous interferences and the upper limit established by the demonstrated practical application of NAA.  
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size and shape during irradiation and counting to a standard sample of known antimony content using very simple sample preparation. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of neutron activation analysis using reactor neutrons (1).2  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. Specific precautions are given in Section 9.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

General Information

• Status: Published
• Publication Date: 31-Dec-2023
• ICS: 71.040.40 - Chemical analysis
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.05 - Nuclear Radiation Metrology

Relations

• Replaces: ASTM E3063-17 - Test Method for Antimony Content Using Neutron Activation Analysis (NAA)
• Effective Date: 01-Jan-2024

Overview

ASTM E3063-24: Standard Test Method for Antimony Content Using Neutron Activation Analysis (NAA) is an internationally recognized analytical standard developed by ASTM for the quantification of antimony (Sb) in plastics and other organic matrices. Employing neutron activation analysis, this method enables highly sensitive, rapid, and nondestructive assessment of antimony content, independent of its chemical form. With a broad detection range from trace levels (1 ng/g) to higher concentrations (10,000 μg/g), ASTM E3063-24 is critical for compliance, quality control, and process verification across a variety of industries.

Key Topics

• Scope of Application: This standard method applies to the analysis of solid and liquid samples, particularly plastics and hydrocarbons, using NAA.
• Neutron Activation Analysis (NAA): The sample is irradiated with neutrons in a nuclear reactor, leading to the emission of characteristic gamma-rays detectable by a high-resolution germanium detector.
• Detection Range: Capable of identifying antimony content from very low concentrations (1 ng/g), useful for both contamination assessment and verification of intentional addition in flame retardant products.
• Sensitivity and Precision: NAA provides high sensitivity and precision, allowing for accurate determination of both high and trace antimony levels with minimal sample preparation.
• Non-Destructive Analysis: The technique preserves the integrity of the sample, which is vital for both research and industrial quality assurance.
• Interference and Correction: The method includes guidelines for handling potential interferences from other radionuclides and outlines the necessary corrections for accurate measurements.

Applications

ASTM E3063-24 offers significant practical value across diverse application areas:

• Flame Retardant Verification: Ensures accurate measurement of antimony levels in flame-retarded plastics, vital for product safety and regulatory compliance.
• Contamination Assessment: Detects trace antimony contamination in petrochemical products, organic matrices, and environmental samples, aiding in quality control and environmental monitoring.
• Industrial Quality and Process Control: Key for maintaining consistent production standards in the manufacturing of plastics, cables, and petrochemical derivatives.
• Research and Development: Supports fundamental and applied research by providing a robust method for antimony quantitation, especially in polymer chemistry and material sciences.
• Regulatory Compliance: Facilitates adherence to international regulations concerning antimony usage and exposure levels.

Related Standards

To ensure complete and consistent results, ASTM E3063-24 should be used in conjunction with other related standards and references:

• ASTM E170 - Terminology Relating to Radiation Measurements and Dosimetry
• ASTM E177 - Practice for Use of the Terms Precision and Bias in ASTM Test Methods
• ASTM E691 - Practice for Conducting an Interlaboratory Study to Determine the Precision of a Test Method
• Code of Federal Regulations, Title 10, Part 20 - U.S. Government regulations on radiation protection and safety
• JCGM 100:2008 - Guide to the Expression of Uncertainty in Measurement (GUM)

These referenced documents provide essential terminology, quality assurance practices, and regulatory context that underpin the application of ASTM E3063-24.

Practical Benefits

In summary, ASTM E3063-24 provides a scientifically robust and industry-accepted methodology for accurately measuring antimony content in plastics and other organic materials. The use of neutron activation analysis delivers reliable, nondestructive results with high sensitivity, supporting a wide range of industrial, research, and compliance needs. Adopting this standard enables manufacturers, laboratories, and regulatory bodies to assure product safety, quality, and integrity, while meeting global standards and regulations regarding antimony content.

Standards Content (Sample)

This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the
Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
Designation: E3063 − 24
Standard Test Method for
Antimony Content Using Neutron Activation Analysis (NAA)
This standard is issued under the fixed designation E3063; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope* 2. Referenced Documents
1.1 This test method covers the measurement of antimony 2.1 ASTM Standards:
concentration in plastics or other hydrocarbon or organic E170 Terminology Relating to Radiation Measurements and
Dosimetry
matrix by using neutron activation analysis (NAA). The
sample is activated by irradiation with neutrons from a research E177 Practice for Use of the Terms Precision and Bias in
ASTM Test Methods
reactor and the subsequently emitted gamma-rays are detected
with a germanium semiconductor detector. The same system E691 Practice for Conducting an Interlaboratory Study to
Determine the Precision of a Test Method
may be used to determine antimony concentrations ranging
from 1 ng/g to 10 000 μg/g with the lower end of the range
2.2 U.S. Government Document:
limited by numerous interferences and the upper limit estab-
Code of Federal Regulations, Title 10, Part 20
lished by the demonstrated practical application of NAA.
2.3 Joint Committee for Guides in Metrology (JCGM)
1.2 This test method may be used on either solid or liquid
Reports:
samples, provided that they can be made to conform in size and
JCGM 100:2008, GUM 1995, with minor corrections Evalu-
shape during irradiation and counting to a standard sample of
ation of measurement data—Guide to the expression of
known antimony content using very simple sample prepara-
uncertainty in measurement
tion. Several variants of this method have been described in the
technical literature. A monograph is available which provides a
3. Terminology
comprehensive description of the principles of neutron activa-
2 3.1 Definitions: See also Terminology E170.
tion analysis using reactor neutrons (1).
3.1.1 comparator standard—a reference standard of known
1.3 The values stated in SI units are to be regarded as
antimony content whose specific activation and counting
−1
standard. No other units of measurement are included in this
sensitivity (counts (mg of antimony) ) may be used to
standard.
quantify the antimony content of a sample irradiated and
1.4 This standard does not purport to address all of the counted under the same conditions. Often, a comparator
safety concerns, if any, associated with its use. It is the standard is selected to have a matrix composition, physical
responsibility of the user of this standard to establish appro- size, density, and shape very similar to the corresponding
priate safety, health, and environmental practices and deter- parameters of the sample to be analyzed. Differences in size,
mine the applicability of regulatory limitations prior to use. density, shape, and matrix composition between sample and
Specific precautions are given in Section 9. standard may be corrected for using physical or empirical
1.5 This international standard was developed in accor- models.
dance with internationally recognized principles on standard-
3.1.2 gamma-ray spectrometer—a system comprising a de-
ization established in the Decision on Principles for the
tector which detects individual gamma-rays and converts their
Development of International Standards, Guides and Recom-
energy into an electronic pulse whose voltage is proportional to
mendations issued by the World Trade Organization Technical
the energy deposited in the detector, and a multichannel
Barriers to Trade (TBT) Committee.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
This test method is under the jurisdiction of ASTM Committee E10 on Nuclear contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
Technology and Applications and is the direct responsibility of Subcommittee Standards volume information, refer to the standard’s Document Summary page on
E10.05 on Nuclear Radiation Metrology. the ASTM website.
Current edition approved Jan. 1, 2024. Published February 2024. Originally Available from the Superintendent of Documents, U.S. Government Printing
approved in 2016. Last previous edition approved in 2017 as E3063 – 17. DOI: Office, Washington, DC 20402.
10.1520/E3063-24. Document produced by Working Groups of the Joint Committee for Guides in
The boldface numbers in parentheses refer to a list of references at the end of Metrology (JCGM). Available free of charge at BIPM website (http://
this standard. www.bipm.org).
*A Summary of Changes section appears at the end of this standard
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E3063 − 24
pulse-height analyzer which measures the pulse heights, as- The weighed sample to be analyzed is placed in a polyethylene
signs a digital value, and stores the individual counts in the container for transfer from the sample loading port to the
channels of a gamma-ray spectrum according to the digital irradiation site in the reactor. Several samples, standards, and
values assigned. monitors may be irradiated simultaneously provided that the
self-shielding effects of multiple samples on each other are
3.1.3 intensity—the probability of emission of a gamma-ray
well understood (7). After irradiation for a pre-selected time,
of a given energy per decay. Another commonly used term is
the samples are returned to the sample receiver. After an
gamma abundance.
appropriate decay period to allow the decay of short-lived
3.1.4 monitor—any type of detector or comparison refer-
radio-isotopes, typically 24 h, the samples are manually
ence material that can be used to produce a response propor-
unpacked and transferred from the receiver to the germanium
tional to the neutron fluence rate in the irradiation position, or
semiconductor detector or to a mechanical sample changer
to the radionuclide decay events recorded by the sample
which transports them one by one at the appropriate time to the
detector.
counting position at the detector. The signals from the detector
3.1.4.1 Discussion—An aluminum wire with 1 mg/g Au are sent to a multichannel pulse-height analyzer which mea-
content is often used as a fluence rate monitor. Iron wires are sures the energies of the individual gamma-rays and places
them in a gamma-ray spectrum. The spectrum has peaks at the
used as well. It is important to distinguish that the monitor is
not a standard used to scale the antimony content of the characteristic energies of the elements present in the sample.
The spectrum for each sample is stored for subsequent analy-
samples to be measured, but rather is used to normalize the
analysis system among samples irradiated simultaneously at sis.
different positions in the polyethylene irradiation sample con-
4.2 The amount of total antimony (all chemical forms) in
tainer or among successive analytical passes within the proce-
the sample is proportional to the corrected and normalized peak
dure. When using reactors with highly reproducible fluence
area and is quantified by use of the corrected and normalized
rate, such as those with 1 % variation over long periods of time,
peak area of the comparator standard(s).
monitors may not be necessary for every irradiation.
4.3 When antimony is irradiated with neutrons, the atoms of
3.1.5 neutron fluence rate—the fluence rate (see definition
121 122
Sb capture neutrons and are converted to Sb
the isotope
in Terminology E170) of neutrons. In this test method it refers
which is radioactive with a half-life of 2.72 days. Sb decays
to the value at the site in the reactor where sample and
by emitting a beta-ray and gamma-rays of several possible
comparator standard are irradiated.
energies. From Ref (8), the main gamma-ray, at 564.2 keV, is
3.1.6 pneumatic transfer system—a system used to transport
emitted in 70.67 % of decays. The amount of total antimony
the sample to the irradiation site in the reactor and then to a
(all chemical forms) in the sample is proportional to the
sample receiver.
corrected number of counts in the peak at 564.2 keV. The area
3.1.6.1 Discussion—It may also be used to transport the
of the peak at 564.2 keV is corrected for counts beneath the
sample directly to the counting station where the activity of the
peak due to Compton scattered gamma-rays and for pulse
sample is measured. For the measurement of antimony, where
losses (dead-time) in the combined detector-multichannel
a long decay time between irradiation and counting is usually
pulse-height analyzer system. All modern multichannel pulse-
required, the samples are manually transferred from the re-
height analyzers accurately correct for pulse losses up to their
ceiver to the germanium semiconductor detector or to a
maximum useable count-rates.
mechanical sample changer which transports them one by one
4.3.1 The detector must have good energy resolution be-
at the appropriate time to the counting position at the detector.
cause the peak at 564.2 keV must be well separated from
3.1.7 research nuclear reactor, n—a nuclear reactor that 82 76
nearby peaks such as those from Br at 554.3 keV and As at
uses the fission of uranium to operate at a well-controlled
559.1 keV.
power level and produces neutrons that can be used for
4.4 In addition to Sb capturing neutrons to produce
experiments and for neutron activation analysis. The opera-
122 123 124
Sb, the Sb isotope captures neutrons to produce Sb,
tional characteristics of reactor types believed to be applicable
with a 60-day half-life. This isotope has strong gamma lines at
to this test method are given in Refs (2-6).
122 124
603 keV and 1691 keV. Measurement of both Sb and Sb
3.1.7.1 Discussion—Another term in common usage is re-
provides an additional verification of the method’s accuracy.
search reactor. Reactor conditions which may make the reactor
This standard employs the most sensitive Sb gamma-ray, but
unsuitable for this test method (for example, very low neutron
the same methods and equations apply equally to all gamma-
fluence rates or high operating temperature) are sample and
rays of both isotopes.
reactor dependent. Such conditions should be considered prior
to use of this test method.
5. Significance and Use
3.1.8 standard uncertainty—measurement uncertainty of the
results of a measurement expressed as a standard deviation
5.1 High levels of antimony are commonly used in flame
(GUM, see 2.3). retardant formulations for various materials. NAA is a test
method that can be useful for verifying these levels and, for
4. Summary of Test Method
other materials, NAA can also be useful in establishing the
4.1 The test method can be applied directly to solid samples amount of low level contamination, if any, with high sensitivity
such as plastic pellets or cylindrical pieces of cable insulation. and high precision.
E3063 − 24
5.2 Neutron activation analysis provides a rapid, highly and force the analyst to count the sample farther from the
sensitive, nondestructive procedure for antimony determina- detector or to wait until the amount of radioactivity decreases.
tion in a wide range of matrices. This test method is indepen- This reduces the sensitivity for the detection of antimony. Also,
dent of the chemical form of the antimony. high-energy gamma-rays which scatter in the detector by the
Compton process and deposit only part of their energy in the
5.3 This test method can be used for quality and process
detector may produce counts in the gamma-ray spectrum near
control in the petrochemical and other manufacturing
564.2 keV. These counts under the Sb 564.2 keV gamma-ray
industries, and for research purposes in a broad spectrum of
peak make it more difficult to determine the peak area and
applications.
increase the uncertainty of the peak area due to counting
statistics.
6. Detection Limit and Range of Application
7.2 A specific potentially interfering radionuclide is As
6.1 Using a research nuclear reactor and germanium semi-
which emits a weak gamma-ray at 563.2 keV, almost the same
conductor detector, the estimated detection limit for antimony
energy as the gamma-ray of Sb. However, this interference
in plastics is 1 ng/g (9). This detection limit may be reduced by
only becomes significant when there is as much arsenic as
using a larger sample, a reactor with higher neutron fluence
antimony in the sample and it can easily be corrected. Knowing
rates, higher counting efficiency detectors, and longer irradia-
that the ratio of the areas of the As peaks at 563.2 keV and
tion and counting times. However, under the conditions of this
559.3 keV is a constant for a given germanium detector and a
test method, the main factor determining the detection limit is
given counting geometry, approximately 0.025, one can correct
the amount of interfering elements in the sample.
the interference for each sample using the observed area of the
6.1.1 The detection limit of 1 ng/g provided in this test
As peak at 559.3 keV.
method presumes clean materials. In materials containing high
amounts of interfering elements, the detection limit may be
7.3 There are a number of other potentially interfering
higher. This detection limit of 1 ng/g implies that, for samples radionuclides which emit gamma-rays near the Sb energy of
actually containing 1 ng/g (not known by the analyst), there is
564.2 keV. They are listed in Table 1. In most cases the
a 50 % chance that the analysis will result in a peak area elements producing these nuclides will be present in the sample
corresponding to greater than 1 ng/g and it will be judged that
in low quantities and the interfering gamma-rays will have a
antimony was detected and a quantitative result will be given. negligible effect on the 564.2 keV peak area.
6.1.2 For this same sample, there is also a 50 % chance that 7.3.1 However, for materials with expected low antimony
the analysis will result in a peak area corresponding to less than content and which may contain these interfering elements in
1 ng/g and it will be judged that antimony was not detected and higher quantities, it is prudent to verify the spectrum for the
a result of “not detected” will be given. For samples containing presence of the associated gamma-ray. If the associated
no antimony (or less than 0.1 ng/g) there is a 2.5 % probability gamma-ray peak is detected, then the interference should be
that the result of the analysis will be greater than 1 ng/g and a corrected using the area of the associated gamma-ray peak. The
quantitative result will be given (false positive). quantity to subtract from the area of the Sb peak at 564.2
keV is the area of the associated peak multiplied by the
6.2 Near the detection limit, the uncertainty in the measured
intensity ratio and multiplied by the ratio of detection efficien-
antimony mass fraction is 0.5 ng/g. This standard uncertainty is
cies at the interfering gamma-ray energy and the associated
caused mainly by statistical fluctuations in the Compton
gamma-ray energy.
background under the small antimony peak at 564.2 keV.
7.3.2 Ac is a naturally occurring radionuclide in the
6.3 With a detection limit of 1 ng/g, the limit of quantitation 232
Th decay series. It is found in the materials of the floor and
(for 10 % uncertainty) is 5 ng/g. This means that, for samples
walls surrounding the detector and will give a peak in the
containing 5 ng/g antimony or more, it is possible to produce
spectrum at 562.9 keV if the detector is not sufficiently
an analysis result with 10 % standard uncertainty or less.
shielded from background radiation.
6.4 At levels above 10 mg/g (1 %), nonlinear effects in the
7.4 An important aspect of this analysis method that must be
relation between observed peak area and antimony concentra-
controlled is the geometry during both irradiation and count-
tion shall be considered and the application of corrections for
ing. Fluence rates will vary between standard and samples
saturation effects such as neutron self-shielding shall be per-
TABLE 1 Potentially Interfering Radionuclides
mitted.
6.4.1 For samples with high antimony content, neutron Interfering gamma- Associated
Nuclide Intensity
ray gamma-ray
Element Half-life
self-shielding correction may use a procedure such as that of
Produced Ratio
Energy Intensity
Ref (10) which takes into account sample size, observed
As As 26.3 h 563.2 1.20 559.3 45.0 0.0267
antimony content, and the ratio of thermal to epithermal
117m
Cd Cd 3.4 h 564.4 14.7 1066.0 23.1 0.6364
fluence rates of the reactor irradiation site used. 134
Cs Cs 2.06 563.2 8.38 604.7 97.6 0.0859
years
Nd Pm 28.4 h 564.9 0.35 340.1 22.0 0.0159
7. Interferences and Necessary Corrections
Eu Eu 13.4 564.0 0.467 1408.0 20.8 0.0225
7.1 All radionuclides which emit high-energy gamma-rays years
152m
Eu Eu 9.34 h 562.9 0.226 841.6 14.6 0.0155
may potentially interfere with the detection of the 564.2 keV
Th Ac back- 562.9 1.01 911.2 29.0 0.0348
gamma-ray of Sb. When these gamma-rays are detected in
ground
large numbers, the high count rate may saturate the detector
E3063 − 24
which are at different positions in the irradiation container. 7.8 The antimony content of the high-purity polyethylene
These variations are easily corrected using flux monitors to sample vials or bags used for irradiation is usually very low
measure the fluence rate gradients or from the knowledge of and it is usually not necessary to transfer the sample to a fresh
variations that are reproducible. container between irradiation and counting, except when the
7.4.1 Similarly, the positioning of the sample at the detector antimony content is expected to be in the nanogram per gram
is critical and must be accurately reproducible. For example, if range.
the sample is considered to be a point source located 6 mm
8. Apparatus
from a germanium detector, a 1 mm change in position of the
sample along the detector axis was found to result in a 5 %
8.1 Research Nuclear Reactor—The operational character-
change in detector efficiency.
istics of common research reactor types are given in (2-6).
7.4.2 Since efficiency is defined as the fraction of gamma
These and larger research nuclear reactors are all suitable for
rays emitted from the source that interact with the detector, it
the measurement of antimony by neutron activation analysis.
is evident that a change in efficiency would result in an equal
Larger reactors produce higher neutron fluence rates.
percentage change in measured activity and in apparent anti-
8.1.1 However, since even the smaller reactors can produce
mony concentration. Such very close sample-detector geom-
adequate sample activities, the use of larger reactors does not
etries are only used for samples with very low antimony
result in better sensitivity for the detection of antimony. With
content and the effect of sample-detector distance variations is
smaller reactors, the tendency is to use larger samples. The
greatly reduced when samples are counted farther from the
larger samples are more representative of the material to be
detector.
analyzed but they may lead to higher neutron self-shielding
corrections. Larger reactors may have a very well thermalized
7.5 Since Sb emits high-energy gamma rays, determina-
irradiation site which essentially eliminates the need for an
tions are usually not significantly affected by variations in
epithermal neutron self-shielding correction.
gamma-ray self-absorption in the sample. Corrections for
8.1.2 Smaller reactors, developed mainly for neutron acti-
gamma-ray attenuation during counting are usually negligible,
vation analysis, tend to be more flexible for neutron activation
except for very large samples and those of very high density
analysis and more readily available. Larger reactors may be
such as heavy metal matrices.
unavailable for long periods of time for refueling or for higher
7.6 The effect of neutron self-shielding may be significant at
priority experiments and their neutron spectra and power levels
high antimony concentrations. Antimony is activated by two
are not under the control of the neutron activation analysis
types of neutrons: thermal (energies below 0.5 eV), and
personnel.
epithermal (energies above 0.5 eV). As thermal and epithermal
8.1.3 Some smaller reactors have stable and highly repro-
neutrons penetrate into the sample, some are absorbed and the
ducible neutron fluence rates at the irradiation sites, which may
center of the sample receives a lower fluence rate than the
eliminate the need for the repeated use of fluence rate monitors.
outside of the sample, and possibly significantly lower than the
If a laboratory chooses not to use fluence rate monitors and to
flux monitor or the standard.
rely on the reproducibility of the neutron fluence rate, it should
7.6.1 As an example of the importance of addressing
have in place a quality assurance program that ensures this
self-shielding, for a cylindrical sample with a diameter of
reproducibility.
3 mm and length of 10 mm and 10 mg/g or 1 % antimony, the
8.2 Pneumatic Sample Transfer System—Samples are usu-
average reduction in fluence rate over the volume of the sample
is 0.007 % for thermal neutrons and 2.1 % for epithermal ally transferred to and from the reactor irradiation position with
a pneumatic system operating with compressed air. However,
neutrons (10). Each facility should characterize their own
self-shielding conditions considering the neutron spectrum at with the fairly long half-life of Sb, 2.72 days, even slower
manual insertion and removal of the samples from the reactor
the specific research nuclear reactor and the size and antimony
is acceptable.
content of the sample.
7.6.2 If neutron self-shielding correction is required in order
8.3 Counting Equipment:
to achieve an acceptable accuracy and detection limit, neutron
8.3.1 Sample Changer—For the measurement of antimony
self-shielding correction may use a procedure such as that of
in batches of samples, a mechanical sample changer is desir-
Ref (9) which takes into account sample size, observed
able. It may be a robotic arm or a turntable. It places the
antimony content, and the ratio of thermal to epithermal
samples one by one at the counting position on the germanium
fluence rates of the reactor irradiation site used.
detector.
7.7 Thermal neutron self-shielding may be significant in 8.3.2 Gamma Detector—A high-resolution germanium
samples of PVC because PVC contains 56.7 % chlorine by semiconductor detector is used. The resolution is usually
weight and chlorine is a strong absorber of thermal neutrons. specified as full-width-half-maximum at 1332 keV; 1.6 keV to
For a cylindrical sample of polyvinyl chloride (PVC) with a 2.0 keV is typical. The resolution should be approximately 1.5
diameter of 3 mm and a length of 10 mm analyzed in a keV at the antimony energy of 564.2 keV. The largest volume
Slowpoke reactor, the average reduction in thermal neutron high-efficiency detector (200 cm ) does not offer much advan-
fluence rate over the volume of the sample is approximately tage over a smaller volume detector (50 cm ). All high-purity
9 % (10). For PVC samples, thermal neutron self-shielding germanium (HPGe) detectors are very stable over long periods
should be corrected. A procedure such as that described in Ref of time and the counting conditions are highly reproducible.
(10) can be used to perform this correction. The standards and samples are usually held at the same
E3063 − 24
distance from the detector using Plexiglas supports. The weighed and encapsulated quickly. Some organic liquids,
laboratory should have in place a quality assurance program especially some fossil fuel byproducts, may diffuse through the
that ensures the reproducibility of the detection efficiency for walls of polyethylene irradiation vials, resulting in sample
the counting geometry used. weight loss if not analyzed promptly after packaging.
8.3.3 Gamma-ray Spectrometer—A gamma-ray spectrom-
10.3 Ideally, the geometry of the sample and that of the
eter is a system comprising a detector which detects individual
comparator standard (see 3.1.1) should be similar in order to
gamma-rays and converts their energy into an electronic pulse
avoid positioning differences during counting. In practice, this
whose voltage is proportional to the energy deposited in the
is usually not possible. There are at least small differences in
detector, and a multichannel pulse-height analyzer containing
size, shape, and density. A physical or empirical model is used
an analog to digital convertor (ADC) which measures the pulse
to correct for these differences.
heights, assigns a digital value, and stores the individual counts
in the channels of a gamma-ray spectrum according to the
11. Calibration and Standardization
digital values assigned.
11.1 Ideally, the comparator standard would be a stable,
8.3.3.1 Modern spectrometer
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation: E3063 − 17 E3063 − 24
Standard Test Method for
Antimony Content Using Neutron Activation Analysis (NAA)
This standard is issued under the fixed designation E3063; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope Scope*
1.1 This test method covers the measurement of antimony concentration in plastics or other hydrocarbon or organic matrix by
using neutron activation analysis (NAA). The sample is activated by irradiation with neutrons from a research reactor and the
subsequently emitted gamma-rays are detected with a germanium semiconductor detector. The same system may be used to
determine antimony concentrations ranging from 1 ng/g to 10 000 μg/g with the lower end of the range limited by numerous
interferences and the upper limit established by the demonstrated practical application of NAA.
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size and shape
during irradiation and counting to a standard sample of known antimony content using very simple sample preparation. Several
variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive
description of the principles of neutron activation analysis using reactor neutrons (1).
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of
regulatory limitations prior to use. Specific precautions are given in Section 9.
1.5 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E177 Practice for Use of the Terms Precision and Bias in ASTM Test Methods
E691 Practice for Conducting an Interlaboratory Study to Determine the Precision of a Test Method
2.2 U.S. Government Document:
Code of Federal Regulations, Title 10, Part 20
This test method is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05
on Nuclear Radiation Metrology.
Current edition approved Nov. 1, 2017Jan. 1, 2024. Published November 2017February 2024. Originally approved in 2016. Last previous edition approved in 20162017
as E3063E3063 – 17.-16. DOI: 10.1520/E3063-17.10.1520/E3063-24.
The boldface numbers in parentheses refer to a list of references at the end of this standard.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Available from the Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402.
*A Summary of Changes section appears at the end of this standard
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E3063 − 24
2.3 Joint Committee for Guides in Metrology (JCGM) Reports:
JCGM 100:2008, GUM 19951995, with minor corrections , with minor corrections, Evaluation of measurement data—Guide to
the expression of uncertainty in measurement
3. Terminology
3.1 Definitions: See also Terminology E170.
3.1.1 comparator standard—a reference standard of known antimony content whose specific activation and counting sensitivity
−1
(counts (mg of antimony) ) may be used to quantify the antimony content of a sample irradiated and counted under the same
conditions. Often, a comparator standard is selected to have a matrix composition, physical size, density, and shape very similar
to the corresponding parameters of the sample to be analyzed. Differences in size, density, shape, and matrix composition between
sample and standard may be corrected for using physical or empirical models.
3.1.2 gamma-ray spectrometer—a system comprising a detector which detects individual gamma-rays and converts their energy
into an electronic pulse whose voltage is proportional to the energy deposited in the detector, and a multichannel pulse-height
analyzer which measures the pulse heights, assigns a digital value, and stores the individual counts in the channels of a gamma-ray
spectrum according to the digital values assigned.
3.1.3 intensity—the probability of emission of a gamma-ray of a given energy per decay. Another commonly used term is gamma
abundance.
3.1.4 monitor—any type of detector or comparison reference material that can be used to produce a response proportional to the
neutron fluence rate in the irradiation position, or to the radionuclide decay events recorded by the sample detector.
3.1.4.1 Discussion—
An aluminum wire with 1 mg/g Au content is often used as a fluence rate monitor. Iron wires are used as well. It is important to
distinguish that the monitor is not a standard used to scale the antimony content of the samples to be measured, but rather is used
to normalize the analysis system among samples irradiated simultaneously at different positions in the polyethylene irradiation
sample container or among successive analytical passes within the procedure. When using reactors with highly reproducible
fluence rate, such as those with 1 % variation over long periods of time, monitors may not be necessary for every irradiation.
3.1.5 neutron fluence rate—the fluence rate (see definition in Terminology E170) of neutrons. In this test method it refers to the
value at the site in the reactor where sample and comparator standard are irradiated.
3.1.6 pneumatic transfer system——aa system used to transport the sample to the irradiation site in the reactor and then to a
sample receiver.
3.1.6.1 Discussion—
It may also be used to transport the sample directly to the counting station where the activity of the sample is measured. For the
measurement of antimony, where a long decay time between irradiation and counting is usually required, the samples are manually
transferred from the receiver to the germanium semiconductor detector or to a mechanical sample changer which transports them
one by one at the appropriate time to the counting position at the detector.
3.1.7 research nuclear reactor, n—a nuclear reactor that uses the fission of uranium to operate at a well-controlled power level
and produces neutrons that can be used for experiments and for neutron activation analysis. The operational characteristics of
reactor types believed to be applicable to this test method are given in Refs (2-6).
3.1.7.1 Discussion—
Another term in common usage is research reactor. Reactor conditions which may make the reactor unsuitable for this test method
(for example, very low neutron fluence rates or high operating temperature) are sample and reactor dependent. Such conditions
should be considered prior to use of this test method.
3.1.8 standard uncertainty—measurement uncertainty of the results of a measurement expressed as a standard deviation (GUM,
see 2.3).
4. Summary of Test Method
4.1 The test method can be applied directly to solid samples such as plastic pellets or cylindrical pieces of cable insulation. The
Document produced by Working Groups of the Joint Committee for Guides in Metrology (JCGM). Available free of charge at BIPM website (http://www.bipm.org).
E3063 − 24
weighed sample to be analyzed is placed in a polyethylene container for transfer from the sample-loading sample loading port to
the irradiation site in the reactor. Several samples, standards, and monitors may be irradiated simultaneously provided that the
self-shielding effects of multiple samples on each other are well understood (7). After irradiation for a pre-selected time, the
samples are returned to the sample receiver. After an appropriate decay period to allow the decay of short-lived radio-isotopes,
typically 24 h, the samples are manually unpacked and transferred from the receiver to the germanium semiconductor detector or
to a mechanical sample changer which transports them one by one at the appropriate time to the counting position at the detector.
The signals from the detector are sent to a multichannel pulse-height analyzer which measures the energies of the individual
gamma-rays and places them in a gamma-ray spectrum. The spectrum has peaks at the characteristic energies of the elements
present in the sample. The spectrum for each sample is stored for subsequent analysis.
4.2 The amount of total antimony (all chemical forms) in the sample is proportional to the corrected and normalized peak area
and is quantified by use of the corrected and normalized peak area of the comparator standard(s).
121 122
4.3 When antimony is irradiated with neutrons, the atoms of the isotope Sb capture neutrons and are converted to Sb which
is radioactive with a half-life of 2.72 days. Sb decays by emitting a beta-ray and gamma-rays of several possible energies. From
Ref (8), the main gamma-ray, at 564.2 keV, is emitted in 70.67 % of decays. The amount of total antimony (all chemical forms)
in the sample is proportional to the corrected number of counts in the peak at 564.2 keV. The area of the peak at 564.2 keV is
corrected for counts beneath the peak due to Compton scattered gamma-rays and for pulse losses (dead-time) in the combined
detector-multichannel pulse-height analyzer system. All modern multichannel pulse-height analyzers accurately correct for pulse
losses up to their maximum useable count-rates.
4.3.1 The detector must have good energy resolution because the peak at 564.2 keV must be well separated from nearby peaks
82 76
such as those from Br at 554.3 keV and As at 559.1 keV.
121 122 123 124
4.4 In addition to Sb capturing neutrons to produce Sb, the Sb isotope captures neutrons to produce Sb, with a 60-day
122 124
half-life. This isotope has strong gamma lines at 603 keV and 1691 keV. Measurement of both Sb and Sb provides an
additional verification of the method’s accuracy. This standard employs the most sensitive Sb gamma-ray, but the same methods
and equations apply equally to all gamma-rays of both isotopes.
5. Significance and Use
5.1 High levels of antimony are commonly used in flame retardant formulations for various materials. NAA is a test method that
can be useful for verifying these levels and, for other materials, NAA can also be useful in establishing the amount of low level
contamination, if any, with high sensitivity and high precision.
5.2 Neutron activation analysis provides a rapid, highly sensitive, nondestructive procedure for antimony determination in a wide
range of matrices. This test method is independent of the chemical form of the antimony.
5.3 This test method can be used for quality and process control in the petrochemical and other manufacturing industries, and for
research purposes in a broad spectrum of applications.
6. Detection Limit and Range of Application
6.1 Using a research nuclear reactor and germanium semiconductor detector, the estimated detection limit for antimony in plastics
is 1 ng/g (9). This detection limit may be reduced by using a larger sample, a reactor with higher neutron fluence rates, higher
counting efficiency detectors, and longer irradiation and counting times. However, under the conditions of this test method, the
main factor determining the detection limit is the amount of interfering elements in the sample.
6.1.1 The detection limit of 1 ng/g provided in this test method presumes clean materials. In materials containing high amounts
of interfering elements, the detection limit may be higher. This detection limit of 1 ng/g implies that, for samples actually
containing 1 ng/g (not known by the analyst), there is a 50 % chance that the analysis will result in a peak area corresponding to
greater than 1 ng/g and it will be judged that antimony was detected and a quantitative result will be given.
6.1.2 For this same sample, there is also a 50 % chance that the analysis will result in a peak area corresponding to less than 1
ng/g and it will be judged that antimony was not detected and a result of “not detected” will be given. For samples containing no
antimony (or less than 0.1 ng/g) there is a 2.5 % probability that the result of the analysis will be greater than 1 ng/g and a
quantitative result will be given (false positive).
E3063 − 24
6.2 Near the detection limit, the uncertainty in the measured antimony mass fraction is 0.5 ng/g. This standard uncertainty is
caused mainly by statistical fluctuations in the Compton background under the small antimony peak at 564.2 keV.
6.3 With a detection limit of 1 ng/g, the limit of quantitation (for 10 % uncertainty) is 5 ng/g. This means that, for samples
containing 5 ng/g antimony or more, it is possible to produce an analysis result with 10 % standard uncertainty or less.
6.4 At levels above 10 mg/g (1 %), non-linearnonlinear effects in the relation between observed peak area and antimony
concentration shall be considered and the application of corrections for saturation effects such as neutron self-shielding shall be
permitted.
6.4.1 For samples with high antimony content, neutron self-shielding correction may use a procedure such as that of Ref (10)
which takes into account sample size, observed antimony content, and the ratio of thermal to epithermal fluence rates of the reactor
irradiation site used.
7. Interferences and Necessary Corrections
7.1 All radionuclides which emit high energy high-energy gamma-rays may potentially interfere with the detection of the 564.2
keV gamma-ray of Sb. When these gamma-rays are detected in large numbers, the high count-rate count rate may saturate the
detector and force the analyst to count the sample farther from the detector or to wait until the amount of radioactivity decreases.
This reduces the sensitivity for the detection of antimony. Also, high energy high-energy gamma-rays which scatter in the detector
by the Compton process and deposit only part of their energy in the detector may produce counts in the gamma-ray spectrum near
564.2 keV. These counts under the Sb 564.2 keV gamma-ray peak make it more difficult to determine the peak area and increase
the uncertainty of the peak area due to counting statistics.
7.2 A specific potentially interfering radionuclide is As which emits a weak gamma-ray at 563.2 keV, almost the same energy
as the gamma-ray of Sb. However, this interference only becomes significant when there is as much arsenic as antimony in the
sample and it can easily be corrected. Knowing that the ratio of the areas of the As peaks at 563.2 keV and 559.3 keV is a constant
for a given germanium detector and a given counting geometry, approximately 0.025, one can correct the interference for each
sample using the observed area of the As peak at 559.3 keV.
7.3 There are a number of other potentially interfering radionuclides which emit gamma-rays near the Sb energy of 564.2 keV.
They are listed in Table 1. In most cases the elements producing these nuclides will be present in the sample in low quantities and
the interfering gamma-rays will have a negligible effect on the 564.2 keV peak area.
7.3.1 However, for materials with expected low antimony content and which may contain these interfering elements in higher
quantities, it is prudent to verify the spectrum for the presence of the associated gamma-ray. If the associated gamma-ray peak is
detected, then the interference should be corrected using the area of the associated gamma-ray peak. The quantity to subtract from
the area of the Sb peak at 564.2 keV is the area of the associated peak multiplied by the intensity ratio and multiplied by the
ratio of detection efficiencies at the interfering gamma-ray energy and the associated gamma-ray energy.
228 232
7.3.2 Ac is a naturally occurring radionuclide in the Th decay series. It is found in the materials of the floor and walls
surrounding the detector and will give a peak in the spectrum at 562.9 keV if the detector is not sufficiently shielded from
background radiation.
TABLE 1 Potentially Interfering Radionuclides
Interfering gamma- Associated
Nuclide Intensity
ray gamma-ray
Element Half-life
Produced Ratio
Energy Intensity
As As 26.3 h 563.2 1.20 559.3 45.0 0.0267
117m
Cd Cd 3.4 h 564.4 14.7 1066.0 23.1 0.6364
Cs Cs 2.06 563.2 8.38 604.7 97.6 0.0859
years
Nd Pm 28.4 h 564.9 0.35 340.1 22.0 0.0159
Eu Eu 13.4 564.0 0.467 1408.0 20.8 0.0225
years
152m
Eu Eu 9.34 h 562.9 0.226 841.6 14.6 0.0155
Th Ac back- 562.9 1.01 911.2 29.0 0.0348
ground
E3063 − 24
7.4 An important aspect of this analysis method that must be controlled is the geometry during both irradiation and counting.
Fluence rates will vary between standard and samples which are at different positions in the irradiation container. These variations
are easily corrected using flux monitors to measure the fluence rate gradients or from the knowledge of variations that are
reproducible.
7.4.1 Similarly, the positioning of the sample at the detector is critical and must be accurately reproducible. For example, if the
sample is considered to be a point source located 6 mm from a germanium detector, a 1 mm change in position of the sample along
the detector axis was found to result in a 5 % change in detector efficiency.
7.4.2 Since efficiency is defined as the fraction of gamma rays emitted from the source that interact with the detector, it is evident
that a change in efficiency would result in an equal percentage change in measured activity and in apparent antimony concentration.
Such very close sample-detector geometries are only used for samples with very low antimony content and the effect of
sample-detector distance variations is greatly reduced when samples are counted farther from the detector.
7.5 Since Sb emits high-energy gamma rays, determinations are usually not significantly affected by variations in gamma-ray
self-absorption in the sample. Corrections for gamma-ray attenuation during counting are usually negligible, except for very large
samples and those of very high density such as heavy metal matrices.
7.6 The effect of neutron self-shielding may be significant at high antimony concentrations. Antimony is activated by two types
of neutrons: thermal (energies below 0.5 eV), and epithermal (energies above 0.5 eV). As thermal and epithermal neutrons
penetrate into the sample, some are absorbed and the center of the sample receives a lower fluence rate than the outside of the
sample, and possibly significantly lower than the flux monitor or the standard.
7.6.1 As an example of the importance of addressing self-shielding, for a 3-mm diameter, 10-mm long cylindrical sample with
10 cylindrical sample with a diameter of 3 mm and length of 10 mm and 10 mg/g or 1 % antimony, the average reduction in fluence
rate over the volume of the sample is 0.007 % for thermal neutrons and 2.1 % for epithermal neutrons (10). Each facility should
characterize their own self-shielding conditions considering the neutron spectrum at the specific research nuclear reactor and the
size and antimony content of the sample.
7.6.2 If neutron self-shielding correction is required in order to achieve an acceptable accuracy and detection limit, neutron
self-shielding correction may use a procedure such as that of Ref (9) which takes into account sample size, observed antimony
content, and the ratio of thermal to epithermal fluence rates of the reactor irradiation site used.
7.7 Thermal neutron self-shielding may be significant in samples of PVC because PVC contains 56.7 % chlorine by weight and
chlorine is a strong absorber of thermal neutrons. For a 3-mm diameter, 10-mm long cylindrical sample of polyvinyl chloride
(PVC) with a diameter of 3 mm and a length of 10 mm analyzed in a Slowpoke reactor, the average reduction in thermal neutron
fluence rate over the volume of the sample is approximately 9 % 9 % (10). For PVC samples, thermal neutron self-shielding should
be corrected. A procedure such as that described in Ref (10) can be used to perform this correction.
7.8 The antimony content of the high-purity polyethylene sample vials or bags used for irradiation is usually very low and it is
usually not necessary to transfer the sample to a fresh container between irradiation and counting, except when the antimony
content is expected to be in the ng/g nanogram per gram range.
8. Apparatus
8.1 Research Nuclear Reactor—The operational characteristics of common research reactor types are given in (2-6). These and
larger research nuclear reactors are all suitable for the measurement of antimony by neutron activation analysis. Larger reactors
produce higher neutron fluence rates.
8.1.1 However, since even the smaller reactors can produce adequate sample activities, the use of larger reactors does not result
in better sensitivity for the detection of antimony. With smaller reactors, the tendency is to use larger samples. The larger samples
are more representative of the material to be analyzed but they may lead to higher neutron self-shielding corrections. Larger
reactors may have a very well thermalized irradiation site which essentially eliminates the need for an epithermal neutron
self-shielding correction.
8.1.2 Smaller reactors, developed mainly for neutron activation analysis, tend to be more flexible for neutron activation analysis
E3063 − 24
and more readily available. Larger reactors may be unavailable for long periods of time for refueling or for higher priority
experiments and their neutron spectra and power levels are not under the control of the neutron activation analysis personnel.
8.1.3 Some smaller reactors have stable and highly reproducible neutron fluence rates at the irradiation sites, which may eliminate
the need for the repeated use of fluence rate monitors. If a laboratory chooses not to use fluence rate monitors and to rely on the
reproducibility of the neutron fluence rate, it should have in place a quality assurance program that ensures this reproducibility.
8.2 Pneumatic Sample Transfer System—Samples are usually transferred to and from the reactor irradiation position with a
pneumatic system operating with compressed air. However, with the fairly long half-life of Sb, 2.72 days, even slower manual
insertion and removal of the samples from the reactor is acceptable.
8.3 Counting Equipment:
8.3.1 Sample Changer—For the measurement of antimony in batches of samples, a mechanical sample changer is desirable. It may
be a robotic arm or a turn-table.turntable. It places the samples one by one at the counting position on the germanium detector.
8.3.2 Gamma Detector—A high-resolution germanium semiconductor detector is used. The resolution is usually specified as
full-width-half-maximum at 1332 keV; 1.6 keV to 2.0 keV is typical. The resolution should be approximately 1.5 keV at the
antimony energy of 564.2 keV. The largest volume-high efficiencyvolume high-efficiency detector (200 cm ) does not offer much
advantage over a smaller volume detector (50 cm ). All high-purity germanium (HPGe) detectors are very stable over long periods
of time and the counting conditions are highly reproducible. The standards and samples are usually held at the same distance from
the detector using Plexiglas supports. The laboratory should have in place a quality assurance program that ensures the
reproducibility of the detection efficiency for the counting geometry used.
8.3.3 Gamma-ray Spectrometer—A Gamma-ray Spectrometergamma-ray spectrometer is a system comprising a detector which
detects individual gamma-rays and converts their energy into an electronic pulse whose voltage is proportional to the energy
deposited in the detector, and a multichannel pulse-height analyzer containing an analog to digital convertor (ADC) which
measures the pulse heights, assigns a digital value, and stores the individual counts in the channels of a gamma-ray spectrum
according to the digital values assigned.
8.3.3.1 Modern spectrometers have typically 8192 channels, so that the details of the spectrum from a high-resolution detector are
clearly visible. Older systems, comprising a pulse amplifier and an analog to digital converter (ADC) in a multichannel
pulse-height analyzer, have now been largely replaced by a Digital Spectrometer,digital spectrometer, which digitizes the pulses
coming directly from the detector preamplifier.
8.3.3.2 The digital spectrometer is usually more stable than the older systems with a pulse amplifier because the amplifier was
often the cause of gain shifts and drift. The digital spectrometer uses a personal computer for the human interface. All
spectrometers lose some counts when two gamma-rays arrive at the detector almost simultaneously. Losses increase with
increasing sample activity and increasing count rate and they must be corrected. The digital spectrometers have excellent
counting-loss counting loss correction systems and are faster than the older systems with amplifier and ADC:ADC; typically they
have three times lower counting losses with the same sample activity.
8.3.4 Gamma-ray Spectrometer Software—The software controls the data acquisition parameters, displays the gamma-ray
spectrum on the computer monitor, and stores the spectra in files.
9. Hazards
9.1 Precautions—Staff handling radioactive material and working near a nuclear reactor need to be properly trained. Follow all
radiation safety regulations and monitor radiation
...