ASTM D7784-20

Standard Practice for the Rapid Assessment of Gamma-ray Emitting Radionuclides in Environmental Media by Gamma Spectrometry

SIGNIFICANCE AND USE
5.1 This practice was developed for the rapid determination of gamma-emitting radionuclides in environmental media. The results of the test may be used to determine if the activity of these radionuclides in the sample exceeds the action level for the relevant incident or emergency response. The detection limits will be dependent on sample size, counting configuration, and the detector system in use.  
5.2 In most cases, a sample container which is large in diameter and short in height relative to the detector will provide the best gamma-ray detection efficiency. For samples of water or other low-Z materials (for example, vegetation), the re-entrant or Marinelli-style beaker may yield the best gamma-ray detection efficiency.  
5.3 The density of the sample material and physical parameters of the sample container (for example, diameter, height, material) may have significant consequences for the accuracy of the sample analysis as compared to the calibration. For this reason, the ideal calibration material and container (often referred to as ‘geometry’) will be exactly the same as the samples to be analyzed. Differences in sample container or sample matrix may introduce significant errors in detector response, especially at low gamma-ray energies. Every effort should be made to account for these differences if the exact calibration geometry is not available.  
5.4 This practice establishes an empirical gamma-ray spectrometer calibration using standards traceable to the SI via a national metrology institute (NMI) such as the National Institute of Standards and Technology (NIST) in the United States and the National Physical Laboratory (NPL) in the United Kingdom in a specific geometry selected to ensure that the container, density, and composition of the standard matches that of the samples as closely as possible. However, in some cases it may be beneficial to modify such initial calibrations using mathematical modeling or extrapolations to an alternate geometry. Use...
SCOPE
1.1 This practice covers the quantification of radionuclides in environmental media (for example, water, soil, vegetation, food) by means of simple preparation and counting with a high-resolution gamma ray detector. Because the practice is designed for rapid analysis, extensive efforts to ensure homogeneity or ideal sample counting conditions are not taken.  
1.2 The values stated in SI units are to be regarded as standard. The values given in parentheses after SI units are provided for information only and are not considered standard.  
1.3 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.  
1.4 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: 14-Dec-2020
• ICS: 17.240 - Radiation measurements
• Technical Committee: D19 - Water
• Drafting Committee: D19.04 - Methods of Radiochemical Analysis

Relations

• Refers: ASTM D1129-13(2020)e2 - Standard Terminology Relating to Water
• Effective Date: 01-May-2020
• Refers: ASTM D7902-20 - Standard Terminology for Radiochemical Analyses
• Effective Date: 01-May-2020
• Refers: ASTM D7902-18 - Standard Terminology for Radiochemical Analyses
• Effective Date: 01-Feb-2018
• Refers: ASTM C998-17 - Standard Practice for Sampling Surface Soil for Radionuclides
• Effective Date: 01-Jun-2017
• Refers: ASTM D7902-16 - Standard Terminology for Radiochemical Analyses
• Effective Date: 01-Feb-2016
• Refers: ASTM D7902-14e1 - Standard Terminology for Radiochemical Analyses
• Effective Date: 15-Jan-2014
• Refers: ASTM D7902-14 - Standard Terminology for Radiochemical Analyses
• Effective Date: 15-Jan-2014
• Refers: ASTM D3370-10 - Standard Practices for Sampling Water from Closed Conduits
• Effective Date: 01-Dec-2010
• Refers: ASTM C998-05(2010)e1 - Standard Practice for Sampling Surface Soil for Radionuclides
• Effective Date: 01-Jun-2010
• Refers: ASTM D1129-10 - Standard Terminology Relating to Water
• Effective Date: 01-Mar-2010
• Refers: ASTM D3370-08 - Standard Practices for Sampling Water from Closed Conduits
• Effective Date: 01-Oct-2008
• Refers: ASTM D3370-07 - Standard Practices for Sampling Water from Closed Conduits
• Effective Date: 01-Dec-2007
• Refers: ASTM D3649-06 - Standard Practice for High-Resolution Gamma-Ray Spectrometry of Water
• Effective Date: 15-Dec-2006
• Refers: ASTM D7282-06 - Standard Practice for Set-up, Calibration, and Quality Control of Instruments Used for Radioactivity Measurements
• Effective Date: 15-Dec-2006
• Refers: ASTM D1129-06a - Standard Terminology Relating to Water
• Effective Date: 01-Sep-2006

Overview

ASTM D7784-20: Standard Practice for the Rapid Assessment of Gamma-ray Emitting Radionuclides in Environmental Media by Gamma Spectrometry provides a streamlined methodology for quantifying gamma-emitting radionuclides in various environmental samples. Developed by ASTM, this practice is essential for rapid screening during radiological emergencies or environmental assessments. It is widely used where time-sensitive measurement of radioactivity is critical, such as incident response, environmental monitoring, and regulatory compliance.

Gamma spectrometry enables the detection and quantification of radionuclides present in materials like soil, water, vegetation, and food, using high-resolution detectors. This practice emphasizes speed, practical sample handling, and empirical calibration to support emergency response and decision-making.

Key Topics

• Rapid Measurement: Designed for quick determination of gamma-emitting radionuclides without demanding extensive sample preparation or homogenization, making it suitable for field applications and emergency scenarios.
• Gamma Spectrometry: Utilizes high-purity germanium detectors for high-resolution analysis, providing accurate identification and quantification across a wide energy range (approx. 20 keV to 2200 keV).
• Calibration: Establishes empirical calibration that matches the geometry and composition of calibration standards to actual samples. SI traceability is ensured through national metrology institutes such as NIST (USA) or NPL (UK).
• Container and Sample Considerations: Highlights the significant impact of sample density and container geometry on measurement accuracy. For low-Z materials like water or vegetation, Marinelli-style containers are often preferred for optimal efficiency.
• Quality Control and Uncertainty Management: Integrates routine quality control with laboratory control samples, blanks, duplicates, and independent reference materials to assure data validity and assess measurement uncertainty.
• Interference Management: Addresses the possibility of spectral interferences and self-absorption, especially for radionuclides emitting low-energy gamma rays (<100 keV).

Applications

• Environmental Monitoring: Useful for routine or emergency assessment of water, soil, and biological samples to determine the presence and activity of gamma-ray emitting radionuclides, aiding in compliance with environmental regulations.
• Emergency Response: Enables rapid screening to determine if observed radioactivity surpasses action levels following a radiological incident, such as radiological dispersal devices or nuclear events.
• Regulatory Compliance: Supports laboratories and agencies in providing defensible results when assessing radioactive contamination in environmental media.
• Public Health and Safety: Assists government and health organizations in incident response planning and the rapid assessment of potential threats to the public from radionuclide contamination.

Related Standards

• ASTM C998 - Practice for Sampling Surface Soil for Radionuclides
• ASTM D1129 - Terminology Relating to Water
• ASTM D3370 - Practices for Sampling Water from Flowing Process Streams
• ASTM D3648 - Practices for the Measurement of Radioactivity
• ASTM D3649 - Practice for High-Resolution Gamma-Ray Spectrometry of Water
• ASTM D7282 - Practice for Set-up, Calibration, and Quality Control of Instruments Used for Radioactivity Measurements
• ASTM D7902 - Terminology for Radiochemical Analyses

Reference data and calibration sources used in this practice can also be found from the National Nuclear Data Center (Brookhaven National Laboratory) and the Bureau International des Poids et Mesures (BIPM).

---

Keywords: gamma spectrometry, ASTM D7784-20, radionuclide assessment, environmental radioactivity, rapid screening, emergency response, radiological analysis, calibration, quality control, spectral interference.

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: D7784 − 20
Standard Practice for the
Rapid Assessment of Gamma-ray Emitting Radionuclides in
Environmental Media by Gamma Spectrometry
This standard is issued under the fixed designation D7784; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope D7902Terminology for Radiochemical Analyses
2.2 Other Documents:
1.1 This practice covers the quantification of radionuclides
BIPM-5Decay Data Evaluation Project (DDEP)
in environmental media (for example, water, soil, vegetation,
NUDAT2
food) by means of simple preparation and counting with a
high-resolution gamma ray detector. Because the practice is
3. Terminology
designed for rapid analysis, extensive efforts to ensure homo-
3.1 Definitions:
geneity or ideal sample counting conditions are not taken.
3.1.1 For definitions of terms used in this standard, refer to
1.2 The values stated in SI units are to be regarded as
Terminologies D1129 and D7902.
standard. The values given in parentheses after SI units are
providedforinformationonlyandarenotconsideredstandard.
4. Summary of Practice
1.3 This standard does not purport to address all of the
4.1 Following sample collection, sample material is placed
safety concerns, if any, associated with its use. It is the
in a suitable container for analysis by a gamma spectrometry
responsibility of the user of this standard to establish appro-
system. A suitable container is defined as a container which
priate safety, health, and environmental practices and deter-
willbothholdthesampleinafixedgeometryandforwhichthe
mine the applicability of regulatory limitations prior to use.
gamma spectrometry system has been calibrated. For solid
1.4 This international standard was developed in accor-
samples, the samples may be ground, sieved, or otherwise
dance with internationally recognized principles on standard-
preparedforthepurposeofvolumereduction,homogenization,
ization established in the Decision on Principles for the
or conformance to the calibration standard, as desired.
Development of International Standards, Guides and Recom-
5. Significance and Use
mendations issued by the World Trade Organization Technical
Barriers to Trade (TBT) Committee.
5.1 This practice was developed for the rapid determination
of gamma-emitting radionuclides in environmental media.The
2. Referenced Documents
results of the test may be used to determine if the activity of
2.1 ASTM Standards:
these radionuclides in the sample exceeds the action level for
C998Practice for Sampling Surface Soil for Radionuclides
the relevant incident or emergency response. The detection
D1129Terminology Relating to Water
limits will be dependent on sample size, counting
D3370Practices for Sampling Water from Flowing Process
configuration, and the detector system in use.
Streams
5.2 In most cases, a sample container which is large in
D3649PracticeforHigh-ResolutionGamma-RaySpectrom-
diameter and short in height relative to the detector will
etry of Water
provide the best gamma-ray detection efficiency. For samples
D7282Practice for Set-up, Calibration, and Quality Control
ofwaterorotherlow-Zmaterials(forexample,vegetation),the
of Instruments Used for Radioactivity Measurements
re-entrantorMarinelli-stylebeakermayyieldthebestgamma-
ray detection efficiency.
This practice is under the jurisdiction ofASTM Committee D19 on Water and
5.3 The density of the sample material and physical param-
is the direct responsibility of Subcommittee D19.04 on Methods of Radiochemical
eters of the sample container (for example, diameter, height,
Analysis.
material) may have significant consequences for the accuracy
Current edition approved Dec. 15, 2020. Published January 2021. Originally
of the sample analysis as compared to the calibration. For this
approved in 2012. Last previous edition approved in 2012 as D7784 – 12. DOI:
10.1520/D7784-20.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM Available from BIPM, Sèvres Cedex, France, https://www.bipm.org.
Standards volume information, refer to the standard’s Document Summary page on Available from National Nuclear Data Center at Brookhaven National
the ASTM website. Laboratory, W Princeton Ave, Yaphank, NY 11980, http://www.nndc.bnl.gov.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
D7784 − 20
TABLE 1 Example of Most Likely Radionuclides for
reason, the ideal calibration material and container (often
Emergency Response
referred to as ‘geometry’) will be exactly the same as the
Gamma
samples to be analyzed. Differences in sample container or
Nuclide Gamma Fraction Half-Life (d)
Energy (keV)
sample matrix may introduce significant errors in detector
Ac-227 100 3.17E-04 7.96E+03
response, especially at low gamma-ray energies. Every effort Ac-227 83.96 2.21E-04 7.96E+03
Ag-110m 657.75 9.47E-01 2.50E+02
should be made to account for these differences if the exact
Ag-110m 884.67 7.29E-01 2.50E+02
calibration geometry is not available.
Am-241 59.54 3.63E-01 1.58E+05
Am-242m 49.3 1.90E-03 5.55E+04
5.4 This practice establishes an empirical gamma-ray spec-
Am-243 74.67 6.60E-01 2.70E+06
trometer calibration using standards traceable to the SI via a
Au-198 411.80 9.55E-01 2.70E+00
Au-198 70.82 1.38E-02 2.70E+00
national metrology institute (NMI) such as the National Insti-
Ba-133 30.97 6.29E-01 3.91E+03
tute of Standards and Technology (NIST) in the United States
Ba-133 355.86 6.23E-01 3.91E+03
and the National Physical Laboratory (NPL) in the United Ba-137m 661.62 9.00E-01 1.77E-03
Ba-137m 32.19 3.82E-02 1.77E-03
Kingdom in a specific geometry selected to ensure that the
Ba-140 537.38 1.99E-01 1.28E+01
container, density, and composition of the standard matches
Ba-140 29.96 1.43E-01 1.28E+01
that of the samples as closely as possible. However, in some
Bi-207 569.67 9.80E-01 1.39E+04
Bi-207 1063.62 7.70E-01 1.39E+04
cases it may be beneficial to modify such initial calibrations
Cd-109 24.95 1.43E-01 4.53E+02
using mathematical modeling or extrapolations to an alternate
Cd-113m 263.7 6.00E-05 5.33E+03
geometry. Use of such a model may be acceptable, depending
Cd-113m 23.17 6.00E-05 5.33E+03
Ce-141 145.45 4.80E-01 3.24E+01
on the measurement quality objectives of the analysis process,
Ce-141 36.03 8.88E-02 3.24E+01
and provided that appropriate compensation to uncertainty
Ce-143 293.3 4.34E-01 1.40E+00
estimates are included. The use of such calibration models is Ce-143 36.03 3.23E-01 1.40E+00
Ce-144 133.53 1.08E-01 2.84E+02
best supported by the successful analysis of a method valida-
Ce-144 36.03 4.80E-02 2.84E+02
tion reference material (MVRM).
Cf-252 43.4 1.30E-04 8.99E+02
Cm-242 44.03 3.25E-04 1.63E+02
5.5 This practice addresses the analysis of numerous
Cm-243 103.75 2.08E-01 1.04E+04
gamma-emitting radionuclides in environmental media. This
Cm-244 42.82 2.55E-04 6.61E+03
Cm-245 103.76 2.30E-01 3.11E+06
practice should be applicable to non-environmental media (for
Co-58 810.75 9.95E-01 7.08E+01
example, urine, debris, or rubble) that have similar physical
Co-58 511 3.00E-01 7.08E+01
properties. The key determination of similar physical proper-
Co-60 1332.51 1.00E+00 1.93E+03
Co-60 1173.23 9.99E-01 1.93E+03
ties is the ability to demonstrate that the gamma spectrometry
Co-60 2158.7 8.00E-06 1.93E+03
system response to the sample configuration is suitably similar
Cr-51 320.07 9.83E-02 2.77E+01
to that for which the system is calibrated.
Cs-134 604.66 9.76E-01 7.53E+02
Cs-134 795.76 8.54E-01 7.53E+02
5.6 For the analysis of radionuclides with low gamma-ray
Cs-136 818.5 1.00E+00 1.30E+01
emission energies (<100 keV), self-absorption of the gamma- Cs-136 1048.07 8.00E-01 1.30E+01
Cs-137 661.62 8.46E-01 1.10E+04
rays in the sample matrix can have a significant adverse effect
Cs-137 32.19 3.70E-02 1.10E+04
on detection and quantification. The user should verify that
Eu-152 40.12 3.00E-01 4.64E+03
instrument calibrations appropriately account for any self- Eu-152 121.78 2.92E-01 4.64E+03
Eu-154 123.1 4.05E-01 3.11E+03
absorption that may result from the sample matrix.
Eu-154 1274.8 3.55E-01 3.11E+03
Eu-155 86.45 3.27E-01 1.81E+03
5.7 Commonly available energy and efficiency calibration
Eu-155 105.31 2.18E-01 1.81E+03
standards cover the energy range of approximately 60 keV to
Fe-59 1099.22 5.65E-01 4.51E+01
1836keV.Resultsobtainedusinggamma-raypeaksoutsidethe
Fe-59 1291.56 4.32E-01 4.51E+01
Gd-153 41.54 6.00E-01 2.42E+02
efficiencycalibratedenergyrangewillhavegreateruncertainty
Gd-153 40.9 3.20E-01 2.42E+02
not accounted for in the uncertainty calculations of this
Hf-181 482.16 8.60E-01 4.25E+01
practice. Great care should be taken to review the efficiency Hf-181 133.05 4.30E-01 4.25E+01
Hg-203 279.17 8.15E-01 4.66E+01
calibration values and the shape of the efficiency curve outside
Hg-203 72.87 6.40E-02 4.66E+01
thisrange.Forgreateraccuracyintheanalysisofradionuclides
Ho-166m 184.41 7.39E-01 4.38E+05
whosegamma-rayenergiesareoutsidethisrange,acalibration
Ho-166m 810.31 5.97E-01 4.38E+05
I-125 27.47 7.30E-01 6.01E+01
standard which includes radionuclide(s) whose gamma-ray
I-125 27.2 3.92E-01 6.01E+01
energies span the energy range of radionuclides of interest is
I-129 29.78 3.60E-01 5.73E+09
advised. I-129 29.46 1.90E-01 5.73E+09
I-131 364.48 8.12E-01 8.04E+00
I-131 636.97 7.27E-02 8.04E+00
6. Interferences
I-131 284.29 6.06E-02 8.04E+00
I-131 80.18 2.62E-02 8.04E+00
6.1 A list of some gamma-ray emitting radionuclides with
I-131 29.78 2.59E-02 8.04E+00
relevantdataisprovided,forinformationonly,inTable1.This
I-132 667.69 9.87E-01 9.92E-02
I-132 772.61 7.62E-01 9.92E-02
listincludesradionuclideswhichmaybeofinteresttoagencies
In-114m 24.21 2.00E-01 4.95E+01
respondingtoalargescaleradiologicalevent.Throughinspec-
In-114m 189.9 1.77E-01 4.95E+01
tion of the list, it becomes apparent that there are numerous
Ir-192 316.49 8.70E-01 7.40E+01
opportunities for interferences based on the gamma energy
D7784 − 20
TABLE 1 Continued TABLE 1 Continued
Gamma Gamma
Nuclide Gamma Fraction Half-Life (d) Nuclide Gamma Fraction Half-Life (d)
Energy (keV) Energy (keV)
Ir-192 468.06 5.18E-01 7.40E+01 Tb-160 298.57 2.74E-01 7.21E+01
K-40 1460.75 1.07E-01 4.68E+11 Tc-99 89.6 6.50E-06 7.82E+07
La-140 1596.2 9.55E-01 1.68E+00 Te-127 417.9 9.93E-03 3.90E-01
La-140 487.03 4.30E-01 1.68E+00 Te-127 360.3 1.35E-03 3.90E-01
Mn-54 834.81 1.00E+00 3.12E+02 Te-129 27.77 1.64E-01 4.83E-02
Mo-99 140.51 9.09E-01 2.76E+00 Te-129 459.5 7.14E-02 4.83E-02
Mo-99 739.47 1.30E-01 2.76E+00 Te-129m 27.47 1.53E-01 3.36E+01
Na-22 511 1.80E+00 9.50E+02 Te-129m 27.2 7.80E-02 3.36E+01
Na-22 1274.54 9.99E-01 9.50E+02 Te-131m 773.67 3.81E-01 1.25E+00
Nb-94 871.1 1.00E+00 7.42E+06 Te-131m 852.21 2.06E-01 1.25E+00
Nb-94 702.5 1.00E+00 7.42E+06 Te-132 228.16 8.85E-01 3.25E+00
Nb-95 765.82 9.90E-01 3.52E+01 Te-132 28.5 5.40E-01 3.25E+00
Nd-147 91.1 2.83E-01 1.11E+01 Th-227 236 1.12E-01 1.85E+01
Nd-147 38.72 2.30E-01 1.11E+01 Th-227 50.2 8.50E-02 1.85E+01
Nd-147 531 1.35E-01 1.11E+01 Th-227 256.25 6.80E-02 1.85E+01
Np-237 86.49 1.31E-01 7.82E+08 Ti-44 78.4 9.47E-01 1.73E+04
Np-237 29.38 9.80E-02 7.82E+08 Ti-44 67.8 8.77E-01 1.73E+04
Np-237 95.87 2.96E-02 7.82E+08 Tl-204 70.82 7.40E-03 1.38E+03
Np-239 103.7 2.40E-01 2.36E+00 Tl-204 68.89 4.00E-03 1.38E+03
Np-239 106.13 2.27E-01 2.36E+00 Tm-170 84.26 1.00E-01 1.29E+02
Pa–234m 1001.03 5.90E-03 8.13E-04 Tm-170 52.39 6.80E-02 1.29E+02
Pa-234m 766.6 2.07E-03 8.13E-04 Tm-170 51.35 3.60E-02 1.29E+02
Pb-210 46.52 4.00E-02 7.45E+03 U-235 185.72 5.40E-01 2.57E+11
Pm-145 37.36 3.86E-01 6.47E+03 U-235 143.76 1.05E-01 2.57E+11
Pm-145 36.85 2.11E-01 6.47E+03 U-235 163.35 4.70E-02 2.57E+11
Pm-147 121.2 4.00E-05 9.58E+02 U-238 48 7.50E-04 1.72E+12
Pm-149 285.9 3.10E-02 2.21E+00 V-48 983.5 1.00E+00 1.61E+01
Pm-149 859.4 1.00E-03 2.21E+00 V-48 1311.6 9.80E-01 1.61E+01
Pm-151 340.08 2.24E-01 1.18E+00 V-48 511 9.80E-01 1.61E+01
Pm-151 40.12 1.66E-01 1.18E+00 W-187 685.74 2.92E-01 9.96E-01
Po-210 803 1.10E-05 1.38E+02 W-187 479.57 2.34E-01 9.96E-01
Pr-144 696.49 1.49E-02 1.20E-02 Y-90 1760.7 1.15E-04 2.67E+00
Pr-144 2185.61 7.70E-03 1.20E-02 Y-91 1204.9 3.00E-03 5.85E+01
Pu-236 47.6 6.90E-04 1.04E+03 Yb-169 50.74 7.81E-01 3.07E+01
Pu-236 109 1.20E-04 1.04E+03 Yb-169 63.12 4.50E-01 3.07E+01
Pu-238 43.45 3.80E-04 3.21E+04 Yb-169 49.77 4.22E-01 3.07E+01
Pu-238 99.86 7.24E-05 3.21E+04 Zn-65 1115.52 5.08E-01 2.44E+02
Pu-239 51.62 2.08E-04 8.81E+06 Zn-65 511 2.83E-02 2.44E+02
Pu-239 129.28 6.20E-05 8.81E+06 Zr-95 756.72 5.48E-01 6.44E+01
Pu-240 45.24 4.50E-04 2.39E+06 Zr-95 724.18 4.42E-01 6.44E+01
Pu-240 104.23 7.00E-05 2.39E+06
Pu-241 98.44 2.20E-05 5.54E+03
Pu-241 94.66 1.20E-05 5.54E+03
Pu-241 111 8.40E-06 5.54E+03
Pu-242 44.7 3.60E-02 1.41E+08
emissions. For this reason, it is important that the determina-
Pu-242 103.5 7.80E-03 1.41E+08
Ra-226 185.99 3.28E-02 5.84E+05
tionofthepresenceofagivenradionuclidebesupportedbyall
Ra-226 83.78 3.10E-03 5.84E+05
available evidence (for example, additional gamma-ray emis-
Rb-86 1076.63 8.76E-02 1.86E+01
sions).
Rh-106 511.8 2.06E-01 3.46E-04
Rh-106 621.8 9.81E-02 3.46E-04
6.2 The data provided in Table 1, Table 2, and Table 3 are
Ru-103 497.08 8.64E-01 3.94E+01
Ru-103 610.33 5.30E-02 3.94E+01 not mandatory and are provided for information only. The
Sb-124 602.71 9.81E-01 6.02E+01
composition of the nuclide library used by the laboratory
Sb-124 1691.04 5.00E-01 6.02E+01
shouldbematchedtotheanalyticalneedandthedatashouldbe
Sb-126 695.1 9.97E-01 1.25E+01
Sb-126 666.2 9.97E-01 1.25E+01 validated using a current reference source such as BIPM-5 and
Sb-127 685.5 3.57E-01 3.85E+00
Sb-127 473 2.50E-01 3.85E+00
Sc-46 1120.52 1.00E+00 8.39E+01
Sc-46 889.26 1.00E+00 8.39E+01
TABLE 2 Example of Most Likely Radionuclides for
Se-75 264.65 5.86E-01 1.20E+02
Emergency Response Subsequent To an Incident Involving a
Se-75 136 5.60E-01 1.20E+02
Radiological Dispersal Device
Sn-113 391.71 6.42E-01 1.15E+02
Sn-113 24.21 3.90E-01 1.15E+02
Alpha Emitters Beta/Gamma Emitters
Sn-123 1089 6.00E-03 1.29E+02
Am-241 Ra-226 Ac-227 P-32
Sn-123 1032 4.00E-04 1.29E+02
Cm-242 Th-228 Bi-210 Pd-103
Sn-125 1066.6 9.00E-02 9.62E+00
Cm-243 Th-230 Bi-212 Pb-210
Sn-125 915.5 4.25E-02 9.62E+00
Cm-244 Th-232 Bi-214 Pb-212
Sn-126 87.57 3.75E-01 3.65E+07
Np-237 U-234 Co-57 Pb-214
Sn-126 26.11 1.89E-01 3.65E+07
Po-210 U-235 Co-60 Pu-241
Sr-89 909.2 9.50E-04 5.05E+01
Pu-238 U-238 I-125 Ra-228
Ta-182 67.75 4.13E-01 1.15E+02
Pu-239 U-Nat I-129 Se-75
Ta-182 1121.28 3.50E-01 1.15E+02
Pu-240 Ir-192
Tb-160 876.37 3.00E-01 7.21E+01
D7784 − 20
TABLE 3 Example of Most Likely Radionuclides for
9. Calibration of High-resolution Gamma-ray
Emergency Response Subsequent To an Incident Involving an
Spectroscopy System
Improvised Nuclear Device
9.1 Accumulate an energy spectrum using a calibration
Alpha Emitters Beta/Gamma Emitters
Am-241 Ba-140/ La-140 Nd-147/Pr-147 Sb-125
standard (8.2) traceable to a national standards body, in the
U-234 Ce-141 Eu-155 Sr-89
geometrical position representing that of the samples to be
U-235 Ce-143/Pr-143 H-3 Sr-90/Y-90
analyzed.Accumulate sufficient net counts (total counts minus
U-238 Ce-144/Pr-144 I-131/Xe-131 Tc-99
Pu-238 Cs-134 I-133 Te-132/I-32
the Compton) in each full-energy gamma-ray peak of interest
Pu-239 Cs-137 Np-239 Zr-95/Nb-95
to obtain a one-sigma relative counting uncertainty of ≤1%.
Pu-240 Eu-154 Ru-103/Rh-103 Zr-97/Nb-97
Mo-99/ Tc-99m Ru-106/Rh-106
9.2 Using the gamma emission data from the calibration
standard and the peak location data from the calibration
Activation Products
Ag-110m Cr-51 Mn-54
spectrum establish the energy per channel relationship (energy
Co-60 Fe-59 Na-24
calibration) as:
En 5Offset1~Ch 3 Slope! (1)
where:
NUDAT2. Other sources of nuclear data may be used at the En = peak energy (keV),
user’s discretion. In all cases, the source should be clearly Offset = energy offset for the energy calibration equation
documented. (keV),
Ch = peak location channel number, and
6.3 Several of the radionuclides listed in Table 1, Table 2,
Slope = energy calibration equation slope (keV/channel).
and Table 3 have X-ray emissions which may interfere with
NOTE 1—Most modern spectroscopy software packages perform this
gamma-ray emissions, particularly below approximately 40
calculation,andmayincludehigher-orderpolynomialtermstoaccountfor
keV. It is the responsibility of the laboratory to ensure that
minor non-linearity in the energy calibration.
X-ray and gamma-ray interferences are accounte
...

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: D7784 − 12 D7784 − 20
Standard Practice for the
Rapid Assessment of Gamma-ray Emitting Radionuclides in
Environmental Media by Gamma Spectrometry
This standard is issued under the fixed designation D7784; 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
1.1 This practice covers the quantification of radionuclides in environmental media (e.g., (for example, water, soil, vegetation,
food) by means of simple preparation and counting with a high-resolution gamma ray detector. Because the practice is designed
for rapid analysis, extensive efforts to ensure homogeneity or ideal sample counting conditions are not taken.
1.2 The values stated in SI units are to be regarded as standard. The values given in parentheses after SI units are provided for
information purposes only.only and are not considered standard.
1.3 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.
1.4 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:
C998 Practice for Sampling Surface Soil for Radionuclides
D1129 Terminology Relating to Water
D3370 Practices for Sampling Water from Flowing Process Streams
D3648 Practices for the Measurement of Radioactivity
D3649 Practice for High-Resolution Gamma-Ray Spectrometry of Water
D7282 Practice for Set-up, Calibration, and Quality Control of Instruments Used for Radioactivity Measurements
D7902 Terminology for Radiochemical Analyses
2.2 Other Documents:
PCNUDAT data filesBIPM-5 National Nuclear Data Center, Brookhaven National Decay Data Evaluation Project (DDEP)Lab,
Upton, NY, USA
NUDAT2
This practice is under the jurisdiction of ASTM Committee D19 on Water and is the direct responsibility of Subcommittee D19.04 on Methods of Radiochemical Analysis.
Current edition approved Nov. 1, 2012Dec. 15, 2020. Published November 2012January 2021. Originally approved in 2012. Last previous edition approved in 2012 as
D7784 – 12. DOI: 10.1520/D7784-12.10.1520/D7784-20.
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’sstandard’s Document Summary page on the ASTM website.
Available from BIPM, Sèvres Cedex, France, https://www.bipm.org.
Available from National Nuclear Data Center at Brookhaven National Laboratory, W Princeton Ave, Yaphank, NY 11980, http://www.nndc.bnl.gov.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
D7784 − 20
3. Terminology
3.1 Definitions—Definitions: for definitions of terms used in this practice, refer to Terminology D1129.
3.1.1 For definitions of terms used in this standard, refer to Terminologies D1129 and D7902.
4. Summary of Practice
4.1 Following sample collection, sample material is placed in a suitable container for analysis by a gamma spectrometry system.
A suitable container is defined as a container which will both hold the sample in a fixed geometry and for which the gamma
spectrometry system has been calibrated. For solid samples, the samples may be ground, sieved, or otherwise prepared for the
purpose of volume reduction, homogenization, or conformance to the calibration standard, as desired.
5. Significance and Use
5.1 This practice was developed for the rapid determination of gamma-emitting radionuclides in environmental media. The results
of the test may be used to determine if the activity of these radionuclides in the sample exceeds the action level for the relevant
incident or emergency response. The detection limits will be dependent on sample size, counting configuration, and the detector
system in use.
5.2 In most cases, a sample container which is large in diameter and short in height relative to the detector will provide the best
gamma-ray detection efficiency. For samples of water or other low-Z materials (e.g., (for example, vegetation), the re-entrant or
Marinelli-style beaker may yield the best gamma-ray detection efficiency.
5.3 The density of the sample material and physical parameters of the sample container (e.g., (for example, diameter, height,
material) may have significant consequences for the accuracy of the sample analysis as compared to the calibration. For this reason,
the ideal calibration material and container (often referred to as ‘geometry’) will be exactly the same as the samples to be analyzed.
Differences in sample container or sample matrix may introduce significant errors in detector response, especially at low
gamma-ray energies. Every effort should be made to account for these differences if the exact calibration geometry is not available.
5.4 This methodpractice establishes an empirical gamma-ray spectrometer calibration using standards traceable to a national
standardizing body in the SI via a national metrology institute (NMI) such as the National Institute of Standards and Technology
(NIST) in the United States and the National Physical Laboratory (NPL) in the United Kingdom in a specific geometry selected
to ensure that the container, density, and composition of the standard matches that of the samples as closely as possible. However,
in some cases it may be beneficial to modify such initial calibrations using mathematical modeling or extrapolations to an alternate
geometry. Use of such a model may be acceptable, depending on the measurement quality objectives of the analysis process, and
provided that appropriate compensation to uncertainty estimates are included. The use of such calibration models is best supported
by the successful analysis of a method validation reference material (MVRM).
5.5 This practice addresses the analysis of numerous gamma-emitting radionuclides in environmental media. This practice should
be applicable to non-environmental media (for example, urine, debris, or rubble) that have similar physical properties. The key
determination of “similarsimilar physical properties”properties is the ability to demonstrate that the gamma spectrometry system
response to the sample configuration is suitably similar to that for which the system is calibrated.
5.6 For the analysis of radionuclides with low gamma-ray emission energies (<100 keV), self-absorption of the gamma-rays in
the sample matrix can have a significant adverse effect on detection and quantification. The user should verify that instrument
calibrations appropriately account for any self-absorption that may result from the sample matrix.
5.7 Commonly available energy and efficiency calibration standards cover the energy range of approximately 60 keV to 1836 keV.
Results obtained using gamma-ray peaks outside the efficiency calibrated energy range will have greater uncertainty not accounted
for in the uncertainty calculations of this practice. Great care should be taken to review the efficiency calibration values and the
shape of the efficiency curve outside this range. For greater accuracy in the analysis of radionuclides whose gamma-ray energies
are outside this range, a calibration standard which includes radionuclide(s) whose gamma-ray energies span the energy range of
radionuclides of interest is advised.
D7784 − 20
6. Interferences
6.1 A list of some gamma-ray emitting radionuclides with relevant data is provided, for information only, in Table 1. This list
includes radionuclides which may be of interest to agencies responding to a large scale radiological event. Through inspection of
the list, it becomes apparent that there are numerous opportunities for interferences based on the gamma energy emissions. For this
reason, it is important that the determination of the presence of a given radionuclide be supported by all available evidence (e.g.,
(for example, additional gamma-ray emissions).
6.2 The data provided in Table 1, Table 2, and Table 3 are not mandatory and are provided for information only. The composition
of the nuclide library used by the laboratory should be matched to the analytical need and the data should be validated using a
current reference source (e.g., Laboratoire National Henri Becquerel, http://www.nucleide.org/DDEP_WG/DDEPdata.htm, or
NuDAT data files, National Nuclear Data Center, Brookhaven National Lab, Upton, NY, USA)such as BIPM-5 and NUDAT2.
Other sources of nuclear data may be used at the user’s discretion. In all cases, the source should be clearly documented.
6.3 Several of the radionuclides listed in Table 1, Table 2, and Table 3 have x-rayX-ray emissions which may interfere with
gamma-ray emissions, particularly below approximately 40 keV. It is the responsibility of the laboratory to ensure that x-rayX-ray
and gamma-ray interferences are accounted for in the analytical process.
7. Apparatus
7.1 Analytical Balance, readable to 0.1 g.
7.2 Sample Container—a container suitable for holding the sample material to be analyzed. The container may be of any suitable
configuration, but should be reproducible in its dimensions and capacity. This should be the same container design for which the
counting system is calibrated. An ideal container is smaller in diameter than the detector to be used for analysis (7.3)(7.3) and
should be as short in the vertical dimension as is practical. A re-entrant beaker (e.g., (for example, Marinelli-style) may be used
to improve the counting efficiency for low Z-value materials. The container should be durable and sealable to prevent content loss
during handling.
7.3 Gamma-Ray Spectrometry System—high resolution high purity germanium gamma spectrometer with an energy range of
approximately 20 keV to 2200 keV (see Practice D3649). Note: The useable energy range of the gamma spectrometer will be
determined by the efficiency calibration. Further guidance on the use of high purity germanium systems may be found in Practice
D3649.
8. Reagents and Materials
8.1 Radioactive Purity—Radioactive purity should be such that the measured radioactivity of blank samples does not compromise
the applicable measurement quality objectives.
8.2 Calibration standard—Known amounts of specific radionuclides whose gamma-ray emission energies cover a wide energy
range should be used for calibration, provided that they have gamma ray energies covering the energy range of the radionuclides
of interest. The known activity of the radionuclides should be traceable to a national standardizing body such as NIST in the
USA.NMI.
9. Calibration of High-resolution Gamma-ray Spectroscopy System
9.1 Accumulate an energy spectrum using a calibration standard (8.2) traceable to a national standards body, in the geometrical
position representing that of the samples to be analyzed. Accumulate sufficient net counts (total counts minus the Compton) in each
full-energy gamma-ray peak of interest to obtain a one-sigma relative counting uncertainty of ≤1%.≤1 %.
9.2 Using the gamma emission data from the calibration standard and the peak location data from the calibration spectrum
establish the energy per channel relationship (energy calibration) as:
En 5 Offset1 C h 3 S l o p e (1)
~ !
D7784 − 20
TABLE 1 Continued
TABLE 1 Example of most likely radionuclidesMost Likely
Radionuclides for
Gamma
Nuclide Gamma Fraction Half-Life (d)
emergency responseEmergency Response Energy (keV)
Ir-192 316.49 8.70E-01 7.40E+01
Gamma
Nuclide Gamma Fraction Half-Life (d)
Ir-192 468.06 5.18E-01 7.40E+01
Energy (keV)
K-40 1460.75 1.07E-01 4.68E+11
Ac-227 100 3.17E-04 7.96E+03
La-140 1596.2 9.55E-01 1.68E+00
Ac-227 83.96 2.21E-04 7.96E+03
La-140 487.03 4.30E-01 1.68E+00
Ag-110m 657.75 9.47E-01 2.50E+02
Mn-54 834.81 1.00E+00 3.12E+02
Ag-110m 884.67 7.29E-01 2.50E+02
Mo-99 140.51 9.09E-01 2.76E+00
Am-241 59.54 3.63E-01 1.58E+05
Mo-99 739.47 1.30E-01 2.76E+00
Am-242m 49.3 1.90E-03 5.55E+04
Na-22 511 1.80E+00 9.50E+02
Am-243 74.67 6.60E-01 2.70E+06
Na-22 1274.54 9.99E-01 9.50E+02
Au-198 411.80 9.55E-01 2.70E+00
Nb-94 871.1 1.00E+00 7.42E+06
Au-198 70.82 1.38E-02 2.70E+00
Nb-94 702.5 1.00E+00 7.42E+06
Ba-133 30.97 6.29E-01 3.91E+03
Nb-95 765.82 9.90E-01 3.52E+01
Ba-133 355.86 6.23E-01 3.91E+03
Nd-147 91.1 2.83E-01 1.11E+01
Ba-137m 661.62 9.00E-01 1.77E-03
Nd-147 38.72 2.30E-01 1.11E+01
Ba-137m 32.19 3.82E-02 1.77E-03
Nd-147 531 1.35E-01 1.11E+01
Ba-140 537.38 1.99E-01 1.28E+01
Np-237 86.49 1.31E-01 7.82E+08
Ba-140 29.96 1.43E-01 1.28E+01
Np-237 29.38 9.80E-02 7.82E+08
Bi-207 569.67 9.80E-01 1.39E+04
Np-237 95.87 2.96E-02 7.82E+08
Bi-207 1063.62 7.70E-01 1.39E+04
Np-239 103.7 2.40E-01 2.36E+00
Cd-109 24.95 1.43E-01 4.53E+02
Np-239 106.13 2.27E-01 2.36E+00
Cd-113m 263.7 6.00E-05 5.33E+03
Pa–234m 1001.03 5.90E-03 8.13E-04
Cd-113m 23.17 6.00E-05 5.33E+03
Pa-234m 766.6 2.07E-03 8.13E-04
Ce-141 145.45 4.80E-01 3.24E+01
Pb-210 46.52 4.00E-02 7.45E+03
Ce-141 36.03 8.88E-02 3.24E+01
Pm-145 37.36 3.86E-01 6.47E+03
Ce-143 293.3 4.34E-01 1.40E+00
Pm-145 36.85 2.11E-01 6.47E+03
Ce-143 36.03 3.23E-01 1.40E+00
Pm-147 121.2 4.00E-05 9.58E+02
Ce-144 133.53 1.08E-01 2.84E+02
Pm-149 285.9 3.10E-02 2.21E+00
Ce-144 36.03 4.80E-02 2.84E+02
Pm-149 859.4 1.00E-03 2.21E+00
Cf-252 43.4 1.30E-04 8.99E+02
Pm-151 340.08 2.24E-01 1.18E+00
Cm-242 44.03 3.25E-04 1.63E+02
Pm-151 40.12 1.66E-01 1.18E+00
Cm-243 103.75 2.08E-01 1.04E+04
Po-210 803 1.10E-05 1.38E+02
Cm-244 42.82 2.55E-04 6.61E+03
Pr-144 696.49 1.49E-02 1.20E-02
Cm-245 103.76 2.30E-01 3.11E+06
Pr-144 2185.61 7.70E-03 1.20E-02
Co-58 810.75 9.95E-01 7.08E+01
Pu-236 47.6 6.90E-04 1.04E+03
Co-58 511 3.00E-01 7.08E+01
Pu-236 109 1.20E-04 1.04E+03
Co-60 1332.51 1.00E+00 1.93E+03
Pu-238 43.45 3.80E-04 3.21E+04
Co-60 1173.23 9.99E-01 1.93E+03
Pu-238 99.86 7.24E-05 3.21E+04
Co-60 2158.7 8.00E-06 1.93E+03
Pu-239 51.62 2.08E-04 8.81E+06
Cr-51 320.07 9.83E-02 2.77E+01
Pu-239 129.28 6.20E-05 8.81E+06
Cs-134 604.66 9.76E-01 7.53E+02
Pu-240 45.24 4.50E-04 2.39E+06
Cs-134 795.76 8.54E-01 7.53E+02
Pu-240 104.23 7.00E-05 2.39E+06
Cs-136 818.5 1.00E+00 1.30E+01
Pu-241 98.44 2.20E-05 5.54E+03
Cs-136 1048.07 8.00E-01 1.30E+01
Pu-241 94.66 1.20E-05 5.54E+03
Cs-137 661.62 8.46E-01 1.10E+04
Pu-241 111 8.40E-06 5.54E+03
Cs-137 32.19 3.70E-02 1.10E+04
Pu-242 44.7 3.60E-02 1.41E+08
Eu-152 40.12 3.00E-01 4.64E+03
Pu-242 103.5 7.80E-03 1.41E+08
Eu-152 121.78 2.92E-01 4.64E+03
Ra-226 185.99 3.28E-02 5.84E+05
Eu-154 123.1 4.05E-01 3.11E+03
Ra-226 83.78 3.10E-03 5.84E+05
Eu-154 1274.8 3.55E-01 3.11E+03
Rb-86 1076.63 8.76E-02 1.86E+01
Eu-155 86.45 3.27E-01 1.81E+03
Rh-106 511.8 2.06E-01 3.46E-04
Eu-155 105.31 2.18E-01 1.81E+03
Rh-106 621.8 9.81E-02 3.46E-04
Fe-59 1099.22 5.65E-01 4.51E+01
Ru-103 497.08 8.64E-01 3.94E+01
Fe-59 1291.56 4.32E-01 4.51E+01
Ru-103 610.33 5.30E-02 3.94E+01
Gd-153 41.54 6.00E-01 2.42E+02
Sb-124 602.71 9.81E-01 6.02E+01
Gd-153 40.9 3.20E-01 2.42E+02
Sb-124 1691.04 5.00E-01 6.02E+01
Hf-181 482.16 8.60E-01 4.25E+01
Sb-126 695.1 9.97E-01 1.25E+01
Hf-181 133.05 4.30E-01 4.25E+01
Sb-126 666.2 9.97E-01 1.25E+01
Hg-203 279.17 8.15E-01 4.66E+01
Sb-127 685.5 3.57E-01 3.85E+00
Hg-203 72.87 6.40E-02 4.66E+01
Sb-127 473 2.50E-01 3.85E+00
Ho-166m 184.41 7.39E-01 4.38E+05
Sc-46 1120.52 1.00E+00 8.39E+01
Ho-166m 810.31 5.97E-01 4.38E+05
Sc-46 889.26 1.00E+00 8.39E+01
I-125 27.47 7.30E-01 6.01E+01
Se-75 264.65 5.86E-01 1.20E+02
I-125 27.2 3.92E-01 6.01E+01
Se-75 136 5.60E-01 1.20E+02
I-129 29.78 3.60E-01 5.73E+09
Sn-113 391.71 6.42E-01 1.15E+02
I-129 29.46 1.90E-01 5.73E+09
Sn-113 24.21 3.90E-01 1.15E+02
I-131 364.48 8.12E-01 8.04E+00
Sn-123 1089 6.00E-03 1.29E+02
I-131 636.97 7.27E-02 8.04E+00
Sn-123 1032 4.00E-04 1.29E+02
I-131 284.29 6.06E-02 8.04E+00
Sn-125 1066.6 9.00E-02 9.62E+00
I-131 80.18 2.62E-02 8.04E+00
Sn-125 915.5 4.25E-02 9.62E+00
I-131 29.78 2.59E-02 8.04E+00
Sn-126 87.57 3.75E-01 3.65E+07
I-132 667.69 9.87E-01 9.92E-02
Sn-126 26.11 1.89E-01 3.65E+07
I-132 772.61 7.62E-01 9.92E-02
Sr-89 909.2 9.50E-04 5.05E+01
In-114m 24.21 2.00E-01 4.95E+01
Ta-182 67.75 4.13E-01 1.15E+02
In-114m 189.9 1.77E-01 4.95E+01
Ta-182 1121.28 3.50E-01 1.15E+02
D7784 − 20
TABLE 1 Continued
Gamma
Nuclide Gamma Fraction Half-Life (d)
Energy (keV)
Tb-160 876.37 3.00E-01 7.21E+01
Tb-160 298.57 2.74E-01 7.21E+01
Tc-99 89.6 6.50E-06 7.82E+07
Te-127 417.9 9.93E-03 3.90E-01
Te-127 360.3 1.35E-03 3.90E-01
Te-129 27.77 1.64E-01 4.83E-02
Te-129 459.5 7.14E-02 4.83E-02
Te-129m 27.47 1.53E-01 3.36E+01
Te-129m 27.2 7.80E-02 3.36E+01
Te-131m 773.67 3.81E-01 1.25E+00
Te-131m 852.21 2.06E-01 1.25E+00
Te-132 228.16 8.85E-01 3.25E+00
Te-132 28.5 5.40E-01 3.25E+00
Th-227 236 1.12E-01 1.85E+01
Th-227 50.2 8.50E-02 1.85E+01
Th-227 256.25 6.80E-02 1.85E+01
Ti-44 78.4 9.47E-01 1.73E+04
Ti-44 67.8 8.77E-01 1.73E+04
Tl-204 70.82 7.40E-03 1.38E+03
Tl-204 68.89 4.00E-03 1.38E+03
Tm-170 84.26 1.00E-01 1.29E+02
Tm-170 52.39 6.80E-02 1.29E+02
Tm-170 51.35 3.60E-02 1.29E+02
U-235 185.72 5.40E-01 2.57E+11
U-235 143.76 1.05E-01 2.57E+11
U-235 163.35 4.70E-02 2.57E+11
U-238 48 7.50E-04 1.72E+12
V-48 983.5 1.00E+00 1.61E+01
V-48 1311.6 9.80E-01 1.61E+01
V-48 511 9.80E-01 1.61E+01
W-187 685.74 2.92E-01 9.96E-01
W-187 479.57 2.34E-01 9.96E-01
Y-90 1760.7 1.15E-04 2.67E+00
Y-91 1204.9 3.00E-03 5.85E+01
Yb-169 50.74 7.81E-01 3.07E+01
Yb-169 63.12 4.50E-01 3.07E+01
Yb-169 49.77 4.22E-01 3.07E+01
Zn-65 1115.52 5.08E-01 2.44E+02
Zn-65 511 2.83E-02 2.44E+02
Zr-95 756.72 5.48E-01 6.44E+01
Zr-95 724.18 4.42E-01 6.44E+01
TABLE 2 Example of most likely radionuclidesMost Likely
Radionuclides for
emergency response subsequent to an incident
involvingEmergency Response Subsequent To an Incident
Involving a
radiological dispersal deviceRadiological Dispersal Device
Alpha Emitters Beta/Gamma Emitters
Am-241 Ra-226 Ac-227 P-32
Cm-242 Th-228 Bi-210 Pd-103
Cm-243 Th-230 Bi-212 Pb-210
Cm-244 Th-232 Bi-214 Pb-212
Np-237 U-234 Co-57 Pb-214
Po-210 U-235 Co-60 Pu-241
Pu-238 U-238 I-125 Ra-228
Pu-239 U-Nat I-129 Se-75
Pu-240 Ir-192
where:
En = peak energy (keV),
Offset = energy offset for the energy calibration equation (keV),
Ch = peak location channel number, and
Slope = energy calibration equation slope (keV/channel).
En 5 Offset1 Ch 3 Slope (1)
~ !
D7784 − 20
TABLE 3 Example of most likely radionuclidesMost Likely
Radionuclides for
emergency response subsequent to an incident
involvingEmergency Response Subsequent To an Incident
Involving an
improvised nuclear deviceImprovised Nuclear Device
Alpha Emitters Beta/Gamma Emitters
Am-241 Ba-140/ La-140 Nd-147/Pr-147 Sb-125
U-234 Ce-141 Eu-155 Sr-89
U-235 Ce-143/Pr-143 H-3 Sr-90/Y-90
U-238 Ce-144/Pr-144 I-131/Xe-131 Tc-99
Pu-238 Cs-134 I-133 Te-132/I-32
Pu-239 Cs-137 Np-239 Zr-95/Nb-95
Pu-240 Eu-154 Ru-103/Rh-103 Zr-97/Nb-97
Mo-99/ Tc-99m Ru-106/Rh-106
Activation Products
Ag-110m Cr-51 Mn-54
Co-60 Fe-59 Na-24
where:
En = peak energy (keV),
Offset = energy offset for the energy calibration equation (keV),
Ch = peak location channel number, and
Slope = energy calibration equation slope (keV/channel).
NOTE 1—Most modern spectroscopy software packages perform this calculation, and may include higher-order polynomial terms to account for minor
non-linearity in the energy calibration.
9.3 Using the gamma emission data from the calibration standard and the peak resolution data from the calibration spectrum
establish the resolution versus energy relationship (resolution calibration) as:
FWHM 5 Offset1~E n 3 S l o p e! (2)
where:
FWHM = Full Width of the peak at one-Half the Maximum counts in the centroid channel (keV),
Offset = width offset for the resolution calibration equation (keV),
En = peak energy (keV), and
Slope = resolution calibration equation slope (keV/keV).
FWHM 5 Offset1~En 3 Slope! (2)
where:
FWHM = full width of the peak at one-half the maximum counts in the centroid channel (keV),
Offset = width offset for the resolution calibration equation (keV),
En = peak energy
...