Office of Radiation and Indoor Air 

National Air and Radiation

Environmental Laboratory

EPA 402-R-09-008

June 2009

www.epa.gov/narel

Radiological Laboratory Sample 

Screening Analysis Guide for

Incidents of National Significance

United States

Environmental Protection

Agency
 EPA 402-R-09-008 

www.epa.gov 

 June 2009 

Revision 0 

Radiological Laboratory 

Sample Screening Analysis Guide for 

Incidents of National Significance

U.S. Environmental Protection Agency 

Office of Air and Radiation 

Office of Radiation and Indoor Air 

National Air and Radiation Environmental Laboratory 

Montgomery, AL 36115 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance

This report was prepared for the National Air and Radiation Environmental Laboratory of the Office of 

Radiation and Indoor Air, United States Environmental Protection Agency. It was prepared by 

Environmental Management Support, Inc., of Silver Spring, Maryland, under contracts 68-W-03-038, work 

assignment 35, and EP-07-037, Work Assignments B-33 and I-33, all managed by David Garman. Mention 

of trade names or specific applications does not imply endorsement or acceptance by EPA. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

i

Preface 

The document describes methods that may be applied by personnel at a radioanalytical laboratory for sample radioactivity screening following a radiological or nuclear incident, such as that 

caused by a terrorist attack. The methods used for the screening of a large number of contaminated samples, and the decisions regarding sample processing, will change based on the radionuclides involved in the event and the incident priorities. The rapid assessment and prioritization of 

individual sample activity concentrations for analytical processing require consistent application 

of the method used for screening the different types of samples that will be generated during 

such an incident. A quality assurance program that addresses this screening process from 

instrument calibration through data reporting will also be necessary to provide defensible 

decisions and data. 

The need to ensure adequate laboratory infrastructure to support response and recovery actions 

following a major radiological incident has been recognized by a number of federal agencies. 

The Integrated Consortium of Laboratory Networks (ICLN), created in 2005 by 10 federal 

agencies1

, consists of existing laboratory networks across the federal government. The ICLN is 

designed to provide a national infrastructure with a coordinated and operational system of 

laboratory networks that provide timely, high-quality, and interpretable results for early detection 

and effective consequence management of acts of terrorism and other events requiring an 

integrated laboratory response. It also designates responsible federal agencies (RFAs) to provide 

laboratory support across response phases for chemical, biological, and radiological agents. To 

meet its RFA responsibilities for environmental samples, EPA has established the Environmental 

Response Laboratory Network (ERLN) to address chemical, biological, and radiological threats. 

For radiological agents, EPA is the RFA for monitoring, surveillance, and remediation, and will 

share responsibility for overall incident response with the U.S. Department of Energy (DOE). As 

part of the ERLN, EPA’s Office of Radiation and Indoor Air is leading an initiative to ensure 

that sufficient environmental radioanalytical capability and competency exist across a core set of 

laboratories to carry out EPA’s designated RFA responsibilities. 

EPA’s responsibilities, as outlined in the National Response Framework, include response and 

recovery actions to detect and identify radioactive substances and to coordinate federal 

radiological monitoring and assessment activities. This document was developed to provide 

guidance to those radioanalytical laboratories that will support EPA’s response and recovery 

actions following a radiological or nuclear incident of national significance (INS). 

The calibration and screening methods outlined in this document provide guidance in gross 

sample radioactivity measurements to support the laboratory’s efforts to process a large influx of 

samples rapidly. These methods are based upon the anticipated varied activity levels that 

incoming samples probably would contain if they were impacted by a radiological dispersion 

device into the atmosphere, water, or soil. 

 

1

 Departments of Agriculture, Commerce, Defense, Energy, Health and Human Services, Homeland Security, 

Interior, Justice, and State, and the U.S. Environmental Protection Agency. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

ii

The use of a planned methodology to assess radioactivity levels of samples that contain significant quantities of radioactive materials will ensure that the radioanalytical data produced will be 

of known quality and appropriate for the intended incident response decisions. This guide will 

assist laboratories in establishing measurement quality objectives (MQOs) for the screening 

instruments. This will allow laboratories to have greater confidence in screening measurements 

as they potentially will be using radionuclides for calibration that are likely to be present in such 

samples. 

As with any technical endeavor, actual radioanalytical projects may require particular methods or 

techniques to meet specific measurement quality objectives. This document cannot address a 

complete catalog of analytical methodologies or potential radionuclides nor does it intend to 

proscribe particular methodologies. Laboratories that have screening techniques using alternative 

methods or instruments in place to address the protocols identified in this guide should continue 

to use them if they support the measurement quality objectives required by the incident. Radionuclide-specific methods to support response and recovery actions following a radiological or 

nuclear INS can be found in Standardized Analytical Methods for Environmental Restoration 

Following Homeland Security Events, Revision 4.0. 

Detailed guidance on recommended radioanalytical practices may be found in the Multi-Agency 

Radiological Laboratory Analytical Protocols Manual (MARLAP) referenced in this document, 

which provides detailed radioanalytical guidance for project planners, managers, and radioanalytical personnel based on project-specific requirements. Familiarity with Chapters 2 and 3 of 

MARLAP will be of significant benefit to the users of this guide. 

This document is one in a planned series designed to present radioanalytical laboratory personnel, Incident Commanders (and their designees), and other field response personnel with key 

laboratory operational considerations and likely radioanalytical requirements, decision paths, and 

default data quality and measurement quality objectives for samples taken after a radiological or 

nuclear incident, including incidents caused by a terrorist attack. Documents currently completed 

or in preparation include: 

• Radiological Laboratory Sample Analysis Guide for Incidents of National Significance – 

Radionuclides in Water (EPA 402-R-07-007, January 2008) 

• Radiological Laboratory Sample Analysis Guide for Incidents of National Significance – 

Radionuclides in Air (EPA 402-R-09-007, June 2009) 

• Radiological Laboratory Sample Screening Analysis Guide for Incidents of National 

Significance (EPA 402-R-09-008, June 2009) 

• Method Validation Guide for Qualifying Methods Used by Radiological Laboratories 

Participating in Incident Response Activities (EPA 402-R-09-006, June 2009) 

• Guide for Radiological Laboratories for the Identification, Preparation, and Implementation 

of Core Operations for Radiological Incident Response (in preparation) 

• Guide for Radiological Laboratories for the Control of Radioactive Contamination and 

Radiation (in preparation) 

• Radiological Laboratory Sample Analysis Guide for Incidents of National Significance – 

Radionuclides in Soil (in preparation) 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

iii

Comments on this document, or suggestions for future editions, should be addressed to: 

Dr. John Griggs 

U.S. Environmental Protection Agency 

Office of Radiation and Indoor Air 

National Air and Radiation Environmental Laboratory 

540 South Morris Avenue 

Montgomery, AL 36115-2601 

(334) 270-3450 

Griggs.John@epa.gov 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

iv

Acknowledgments

This manual was developed by the National Air and Radiation Environmental Laboratory 

(NAREL) of EPA’s Office of Radiation and Indoor Air (ORIA). Dr. John Griggs was the project 

lead for this document. Several individuals provided valuable support and input to this document 

throughout its development. Special acknowledgment and appreciation are extended to Dr. Keith 

McCroan, ORIA/NAREL; Mr. Daniel Mackney for instrumental sample analysis support, 

ORIA/NAREL; Ms. Schatzi Fitz-James, Office of Emergency Management, Homeland Security 

Laboratory Response Center; and Mr. David Garman, ORIA/NAREL. We also wish to 

acknowledge the external peer reviews conducted by Lindley J. Davis and Carolyn Wong, whose 

thoughtful comments contributed greatly to the understanding and quality of the report. 

Numerous other individuals, both inside and outside of EPA, provided peer review of this 

document, and their suggestions contributed greatly to the quality and consistency of the final 

document. Technical support was provided by Dr. N. Jay Bassin, Dr. Anna Berne, Mr. David 

Burns, Dr. Carl V. Gogolak, Dr. Robert Litman, Dr. David McCurdy, and Mr. Robert Shannon 

of Environmental Management Support, Inc. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

v

Contents 

Acronyms, Abbreviations, Units, and Symbols............................................................................ vii 

Radiometric and General Unit Conversions .................................................................................. ix 

I. INTRODUCTION ................................................................................................................... 1 

A. Purpose and Objectives....................................................................................................... 4 

B. Scope of DQOs/MQOs for the Screening Process ............................................................. 4 

C. Measurement Quality Objectives: Relationship of Derived Concentrations, AAL, ADL, 

Risk Levels, and uMR to Dose ............................................................................................... 5 

II. RADIONUCLIDES................................................................................................................. 7 

III. DISCUSSION.......................................................................................................................... 8 

A. Sample Screening and Processing at the Laboratory.......................................................... 8 

Gross Activity Measurement Instruments .......................................................................... 8 

Instrument Response Characteristic Determination.......................................................... 10 

Crosstalk: Detector Responses to Radioactive Emissions................................................ 10 

Detector Background ........................................................................................................ 13 

Sample Geometry.............................................................................................................. 14

Laboratory Instruments..................................................................................................... 14 

B. Calibration of Instrumentation for Screening Analyses.................................................... 14 

Detector Type.................................................................................................................... 14 

Geometry........................................................................................................................... 17 

Crosstalk, Dead-Band, and Self-Absorption Factors........................................................ 21 

Final Instrument Calibration and Method Validation....................................................... 21 

C. Calibration of Screening Instruments when Radionuclide Identities are Known ............... 22 

D. Measurement Quality Objectives (MQOs) for the Screening Process ............................. 24 

E. Key Recommendations ..................................................................................................... 25 

Appendix I – Screening Instrumentation Initial Calibration......................................................... 26 

Appendix II – Radiological Event Screening for 241Am............................................................... 28 

Appendix III – Screening Instrumentation Response Corrected for Different Radionuclide....... 30 

Appendix IV – Additional Sources and References ..................................................................... 32 

Figures 

Figure 1 – Halogen-Quenched GM Detector Response to Gamma Radiation (A) with Shield 

Open (B) with Shield Closed ................................................................................................ 16 

Figure 2 – Gamma Energy Response for a Na(Tl) detector ......................................................... 16 

Figure 3 – Shown Without Bricks Covering Top of Shielded Geometry. NaI(Tl) Detector 

Example ................................................................................................................................ 19 

Figure 4 – An Improved Orientation for Shielding. Active Detector Area Within Shielding...... 19 

Figure 5 – Relative Size of Shielded Volume............................................................................... 20 

Figure 6 – Sample Shielding and Detector Orientations for Gross Screening of Air Particulate 

Filters Using an Alpha/Beta Pancake Detector..................................................................... 20 

Figure 7 – Survey Meter 12345 Energy Open Window Response Curve for Beta Emitters........ 23 

Figure 8 – Gamma Energy Response Curve for a NaI(Tl) Detector ............................................ 30 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

vi

Tables 

Table 1 – Radionuclides of Concern............................................................................................... 7 

Table 2 – Detectors Used for Gross Sample Screening.................................................................. 9 

Table 3 – Radionuclides Spanning the Energy-Calibration Range .............................................. 15 

Table 4 – Response and Figure of Merit for 60Co and 137Cs with Different NaI(Tl) Detector 

Configurations....................................................................................................................... 18 

Table 5 – Screening Instrument Conversion Factor Based on Sample Analysis of a 1-Liter 

Sample Geometry.................................................................................................................. 23 

Table 6 – Calibration Data for Screening Instrument Response................................................... 26 

Table 7 – Response Factors (RF) for Radionuclides with Respective Screening Equipment...... 27 

Table 8 – Gross Screening Measurement Results from Transportation Incident ......................... 29 

Table 9 – Results of Screening Measurement Using Adjusted Response .................................... 31 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

vii

Acronyms, Abbreviations, Units, and Symbols 

(Excluding chemical symbols and formulas)

" ........................alpha particle 

AAL ..................analytical action level 

ADL ..................analytical decision level 

AL .....................action level 

$ ........................beta particle 

Bq......................becquerel (1 dps) 

CERCLA...........Comprehensive Environmental Response, Compensation, and Liability Act of 

1980 (“Superfund”) 

cfm ....................cubic feet per minute 

CFR...................Code of Federal Regulations

cm......................centimeter 

cpm....................counts per minute 

d.........................day 

DAC ..................derived air concentration 

DL .....................discrimination limit 

DOE ..................U.S. Department of Energy 

DP .....................decay product(s) 

dpm ...................disintegration per minute 

dps.....................disintegration per second 

DQO..................data quality objective 

DRP...................discrete radioactive particle 

e

– .......................electron 

E$max ..................maximum energy of the beta-particle emission 

EDD ..................electronic data deliverable 

EPA...................U.S. Environmental Protection Agency 

ERLN...............Environmental Response Laboratory Network 

FOM..................figure of merit 

( ........................gamma ray 

g.........................gram 

Ge......................germanium [semiconductor] 

GM ....................Geiger-Muller detector 

GP .....................gas proportional 

GPC...................gas proportional counting [counter] 

GS .....................gamma spectrometry 

Gy......................gray 

h.........................hour 

H0 ......................null hypothesis 

H1 ......................alternate hypothesis 

HPGe.................high-purity germanium detector 

IC.......................Incident Commander [or designee] 

ICLN ...............Integrated Consortium of Laboratory Networks 

IND ...................improvised nuclear device 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

viii

INS ....................incident of national significance 

keV....................thousand electron volts 

L ........................liter 

LBGR................lower bound of the gray region 

LCS ...................laboratory control sample 

LEPD.................low-energy photon detector 

LS......................liquid scintillation 

LSC ...................liquid scintillation counter 

MARLAP..........Multi-Agency Radiological Laboratory Analytical Protocols Manual

MARSSIM ........Multi-Agency Radiation Survey and Site Investigation Manual

MCL..................maximum contaminant level 

MDC .................minimum detectable concentration 

MeV ..................million electron volts 

mg .....................milligram (10–3 g) 

min ....................minute 

mL.....................milliliter (10–3 L) 

MQO .................measurement quality objective 

mR.....................milliroengten (10–3 R) 

mrem .................millirem (10–3 rem) 

:g ......................microgram (10–6 g) 

NaI(Tl) ..............thallium-activated sodium iodide detector 

NORM ............. naturally occurring radioactive materials 

nMR.....................relative method uncertainty 

PAG...................Protective Action Guide 

pCi.....................picocurie (10–12 Ci) 

QA.....................quality assurance 

QC.....................quality control 

rad .....................radiation absorbed dose 

RDD ..................radiological dispersal device (i.e., “dirty bomb”) 

RDL...................required detection limit 

REGe.................reverse electrode germanium detector 

rem ....................roentgen equivalent man 

RFA..................responsible federal agency 

s.........................second 

SI.......................International System of Units 

SOP ...................standard operating procedure 

STS....................sample test source 

Sv ......................sievert 

t1⁄2 .......................half-life 

TAT...................turnaround time 

TEDA................triethylenediamine 

TEDE ................total effective radiation dose equivalent 

UBGR ...............upper bound of the gray region 

uMR .....................required method uncertainty 

y.........................year 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

ix

Radiometric and General Unit Conversions 

To Convert To Multiply by To Convert To Multiply by 

years (y) seconds (s) 

minutes (min) 

hours (h) 

days (d) 

3.16×107

5.26×105

8.77×103

3.65×102

s 

min 

h 

d 

y 3.17×10–8

1.90×10–6

1.14×10–4 

2.74×10–3

disintegrations per 

second (dps) Becquerels (Bq) 1 Bq dps 1 

Bq 

Bq/kg 

Bq/m3

Bq/m3

picocuries (pCi) 

pCi/g 

pCi/L 

Bq/L 

27.0 

2.70×10–2

2.70×10–2

10-3

pCi 

pCi/g 

pCi/L 

Bq/L 

Bq 

Bq/kg 

Bq/m3

Bq/m3

3.70×10–2

37.0 

37.0 

103

microcuries per 

milliliter (:Ci/mL) pCi/L 109 pCi/L :Ci/mL 10–9

disintegrations per 

minute (dpm) 

:Ci 

pCi 

4.50×10–7

4.50×10–1 pCi dpm 2.22 

cubic feet (ft3) cubic meters (m3) 2.83×10–2 m3 ft3 35.3 

gallons (gal) liters (L) 3.78 liters gallons 0.264 

gray (Gy) rad 102 rad Gy 10–2

roentgen equivalent 

man (rem) 

sievert (Sv) 10–2 Sv rem 102

Note: Traditional units are used throughout this document instead of SI units. Protective Action 

Guides (PAGs) and their derived concentrations appear in official documents in the traditional 

units and are in common usage. Conversion to SI units will be aided by the unit conversions in 

this table. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

1

I. INTRODUCTION 

Most laboratories do not routinely screen samples under conditions found during an emergency 

response situation, such as from a radiological or nuclear incident of national significance (INS). 

Many of these samples are higher in activity and need to be accurately surveyed and prioritized 

for analysis based on direction from the Incident Commander (IC).1 This document describes 

methods that may be applied by personnel at a radioanalytical laboratory for screening of 

samples for radioactivity. The specific techniques described in this guide may be used to assess 

the gross α, β, or γ activity in samples that may have been contaminated as the result of a radiological or nuclear event, such as a radiological dispersion device (RDD), improvised nuclear 

device (IND), or an intentional release of radioactive materials into the atmosphere or a body of 

water or aquifer, or to terrestrial areas via mechanical or other methods. In the event of a major 

incident that releases radioactive materials to the environment, EPA will turn to selected radioanalytical laboratories to support its response and recovery activities. In order to expedite sample 

analyses and data feedback, the laboratories will need guidance on EPA’s expectations. 

A response to a release of radioactivity to the environment likely will occur in three phases that 

are generally defined in this document as: “early” (onset of the event to about day 4), “intermediate” (about day 4 to about day 30), and “recovery” (beyond about day 30). Each phase of an 

incident response will require different and distinct radioanalytical resources to address the 

different consequences, management, priorities, and requirements of a phase. Some of the more 

important radioanalytical laboratory issues germane to an incident response consist of radionuclide identification and quantification capability, sample load capacity, sample processing turnaround time, quality of analytical data, and data transfer capability. This guide emphasizes the 

laboratory screening of samples from the end of the early phase, through the intermediate phase, 

and into the recovery phase (but does not address the screening by initial responders). 

Although not the focus of this document, during the early phase, analytical priorities need to 

address the protection of the public and field personnel due to potentially high levels of 

radioactivity and the need to provide for qualitative identification of radionuclides. During this 

phase, the Protective Action Guides (PAGs) for radiological emergencies require evacuation of a 

population if the projected short-term total effective radiation dose equivalent2

 (TEDE) exceeds 

1 rem.3

 The nominal trigger for sheltering is 1 rem over four days (projected avoided inhalation 

dose). The radioanalytical resource requirements (field or fixed laboratory) for this early phase 

may vary significantly depending on the time frame, source-term nuclide, and the extent of the 

contamination. 

During the intermediate phase, the radionuclides and matrices of concern are known qualitatively, and the quantitative levels suitable for making decisions based on action levels need to be 

 

1

 Throughout this guide, the term “Incident Commander” (or “IC”) includes his or her designee. 

2

 The sum of the effective dose equivalent (for external exposure) and the committed effective dose equivalent (for 

internal exposure). TEDE is expressed in units of sievert (Sv) or rem. 

3

 The common unit for the effective or “equivalent” dose of radiation received by a living organism, equal to the 

actual dose (in rads) multiplied by a factor representing the danger of the radiation. “Rem” stands for “roentgen 

equivalent man,” meaning that it measures the biological effects of ionizing radiation in humans. One rem is equal 

to 0.01 Sv. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

2

rapidly determined. For the intermediate phase, PAGs have been established to limit the 

projected radiation doses for different exposure periods, not to exceed 2-rem TEDE over the first 

year, 500-mrem TEDE during the second or any subsequent year, or 5 rem over the next 50 years 

(including the first and second years of the incident). In addition, radionuclide concentration 

limits for food and water as regulated by the Food and Drug Administration and EPA would be 

applicable. 

The final, or “recovery,” phase occurs as part of a radiological incident site-remediation effort. 

During this phase, when site atmospheric characterization and remediation cleanup effectiveness 

are determined, there is potential for more extensive radiochemical analyses at the lowest radionuclide concentrations. 

The analytical resources needed during any phase of a radiological event will depend on the 

radionuclide analytical action level (AAL)4 developed for the various media that may affect 

human exposure. The radionuclide AALs, which are derived radionuclide concentrations for the 

different media types based on the PAGs or risk values, may change depending upon the phase 

of the event. 

The time period of an incident where this document will find its greatest utility is early in the 

intermediate phase through the end of the recovery phase. Laboratories performing analyses 

must focus on optimizing sample analyses so that the initial qualitative aspects and concentrations related to the appropriate AALs can be determined quickly (i.e., rapid turnaround of 

sample results). Radioanalytical screening by laboratories during these phases will include 

methods for all three radioactive emissions. During the recovery phase, however, the screening 

techniques used for samples will be more focused because the radionuclides from the event are 

likely to have already been identified and chemically characterized. 

During all phases of an incident response, radioanalytical resources are needed for the gross 

radiation screening of samples for prioritization of sample processing or for information related 

to the general level of contamination, identification of the radionuclide source term, and 

quantification of the radionuclides in a variety of sample media. This document has been 

developed to provide guidance during an incident on techniques to enhance the ability to 

differentiate radioactivity in samples near action levels and optimize the calibration of the 

screening equipment used for gross sample activity measurement. Using these techniques should 

help laboratories to prioritize samples in a timely fashion based on the request of the IC. 

The process of screening samples using a survey instrument can be described in two stages. The 

first stage deals with the receipt of the bulk sample shipment and assessment of the radiation 

dose rate (mrem/h) or gross activity (cpm) from the shipment and the individual samples, prior 

to opening any samples. The main purpose in this stage is to identify any immediate radiological 

hazard to the receipt personnel and sample analysts. This screening measurement typically is 

made using an instrument that does not discriminate particle energies or assess total dose rate 

from the sample. For example, an instrument like a Geiger-Mueller (GM) detector is sensitive to 

 

4

 The term “analytical action level” (AAL) is used in this publication series as a general term denoting the radionuclide concentration at which action must be taken by incident responders. The AAL should always correspond to 

a PAG or risk-based dose. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

3

all gamma and beta particles with enough energy to pass through the container walls without 

identifying which is which. At this time, no assessment of alpha particle or low-energy beta 

particle contamination can be made. The measurement should not take more than 5 to 10 seconds 

to complete per sample. Important aspects of the outcome of this measurement are that the 

samples can be appropriately shielded and labeled for both radiation protection and prioritization 

purposes, and that the sample mass and integrity remain unchanged (this is a non-destructive, 

non-invasive test). 

The second stage of screening is more substantive in that it examines the total radionuclide 

activity for a particular type of particle emitted from the radionuclides contained within the 

sample. Ideally, if 90Sr, 14C, and 99Tc were all contained in a sample, the instrument used for 

screening would measure the total contribution as the sum of the three, even if it could not 

identify them individually. Unfortunately, the instruments used for screening are not ideal: 

detector response tends to be proportional to the characteristic energy of the radiation emitted by 

a radionuclide and the detection is also impacted by sample self-shielding. It is very important to 

ensure that a screening test will provide a conservative estimate of the total activity of the 

radionuclides present to ensure that the screen does not underestimate the total amount of a 

radionuclide present. If the identity of the radionuclides is known, a different response factor 

should be applied when measuring the medium-to-high energy beta from 90Sr/90Y than for the 

lower energy 14C in samples where mass attenuation may be significant. 

Using gas proportional counting (GPC) or liquid scintillation counting (LSC) to perform the 

screening process has several important consequences. First, when the sample container itself is 

opened, the potential exists for contaminating both the sample and the laboratory. Second, a 

portion of the sample may be sacrificed for the screening process, which may require judicious 

sub-sampling. Third, chain-of-custody must be established for open sample containers and 

aliquanting prior to actual analysis. This will prevent questions later on regarding the sample 

integrity. 

This document provides technical information and recommendations for a laboratory faced with 

screening samples received following a radiological INS. Screening samples deals with the 

detector responses to radiation and the effects of different forms of radiation on different detector 

types. Three appendices provide detailed scenarios that use the information in the technical 

section of the document. These scenarios illustrate when to change calibration and screening 

techniques based upon what is known about the sample’s radioactive contaminants and the 

instrument detection efficiency. The methods demonstrated by these scenarios are: 

• Preparation of laboratory screening equipment for an INS event; 

• Receipt of samples from an INS event with known radionuclides for which the laboratory 

screening instruments are calibrated; and 

• Receipt of samples from an INS event with known radionuclides, but the laboratory 

screening equipment must use a detection correction factor because the instruments were 

not calibrated with radionuclides present in the event samples. 

Facility personnel should use these examples as guidance to prepare the screening instruments 

that are commonly used in their laboratories to analyze gross activity in samples from an INS. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

4

A. Purpose and Objectives 

This document describes how to develop laboratory methods to perform gross radioactivity 

analysis for samples resulting from an INS. It discusses technical issues associated with 

screening measurements, provides the suggested methodologies to determine correction factors 

for these instruments, offers a consistent methodology for measuring sample gross activity 

concentrations, and provides guidance on the calibration of screening equipment commonly used 

by laboratories. 

Although the list of potential threat radionuclides is relatively short, instrument responses to the 

different particle energies may vary significantly depending upon the type of screening 

instrument used. It is important to be able to use screening instrumentation to support the overall 

laboratory process of sample prioritization and analysis that will support decisions to protect the 

health and safety of the public. 

This document provides guidance for a user to select appropriate methods for screening at 

different points in the analytical process. The critical points in the process are sample receipt, 

sample prioritization, and rapid feedback to the IC on samples exceeding action levels. 

The specific objectives for response personnel to accomplish in preparing their laboratories for 

such an event include: 

• Performance of method validation for each instrument/sample geometry combination 

used in screening; 

• Identification of consistent methods of screening for various media; 

• Screening instrument configurations that streamline the screening process; 

• Screening measurements that will aid in prioritizing samples for analyses; and 

• Methods for calibration of screening equipment that will have the widest applicability to 

those radionuclides most likely to result from an INS. 

B. Scope of DQOs/MQOs for the Screening Process 

The use of screening instrumentation to prioritize samples based on the amount of activity in an 

individual sample should be consistent for all laboratories responding to an INS. This should 

allow the processing of samples and return of results to the IC based on the measurement quality 

objectives (MQOs) of the event in the timeliest manner. During the early phase of an event when 

the identity and extent of radioactive contamination are unknown, the screening instrumentation 

should be calibrated with radionuclides that are routinely used for gross screening calibrations, 

but in a geometry that should support the best discrimination of activity levels. As the event 

progresses and the specific radionuclides are identified, either the calibration may be changed to 

reflect the known radionuclides or an interpolated correction factor for instrument response due 

to other radionuclides based on energy should be used. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

5

Other guides in this series5 identify Protective Action Guides (PAGs) as associated concentrations and AALs that are critical measurement limits. The screening instruments used in the 

laboratory to support the rapid and organized evaluation of sample priority should be calibrated 

for gross activity measurements at these critical measurement limits in order to achieve the 

established MQOs stated in the other guides. 

Samples that have the potential for considerations in a criminal investigation must be handled 

separately, and the laboratory should receive information from the Incident Commander on how 

to process these samples. 

C. Measurement Quality Objectives: Relationship of Derived Concentrations, 

AAL, ADL, Risk Levels, and uMR to Dose 

MQOs External to the Laboratory

Gross activity screening of samples is the first step to assessing whether or not a particular 

sample exceeds a PAG’s derived radionuclide concentration for the matrix that is being assessed 

for radioactive materials. PAGs establish radiation dose limits applicable to different phases of 

an incident response. The PAG (expressed as a numerical dose level) indicates a level of 

exposure at which protective action should be taken to prevent, reduce, or limit a person’s 

radiation dose during a radiological incident. The measurements that are made with screening 

instruments in the radioanalytical laboratory should be correlated to the PAGs expressed as 

concentrations (or other AALs) for each matrix defined by the incident. 

A derived concentration of a radionuclide that corresponds to a PAG or risk-based dose in a 

specific matrix can be calculated and used to facilitate the application of these radioanalytical 

action levels in the laboratory for decision-making purposes. For example, the derived air 

concentration (DAC, in units of pCi/m3

) of an individual radionuclide in air corresponds to a 

radiation dose (PAG) to a specific population. For each matrix that undergoes screening, there 

should be a derived radionuclide concentration that may be directed by regulation or selected

based on the specific incident. 

Screening instruments, when configured properly, can be used to conservatively determine if a 

sample has or has not exceeded an AAL. However, when the total gross screening activity 

exceeds an AAL, it may not be possible to determine if the AAL for an individual radionuclide is 

actually exceeded until radionuclide-specific methods are performed. In cases where it is not 

possible to determine if an individual radionuclide AAL has been exceeded, screening provides 

the laboratory with the information to prioritize samples that need to be analyzed first. The 

priority for sample analyses will be decided based upon the incident phase and the specific needs 

of the IC. For example, the order of analysis could be based on highest activity first, lowest 

activity first, gamma response first, or any such logical priority. 

 

5

 See Appendix IV for further references to how measurements are used to make decisions regarding PAGs and 

action levels. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

6

MQOs Internal to the Laboratory

The laboratory also needs the screening equipment to correlate to MQOs established in the 

laboratory and thus facilitate sample processing. The screening MQO will likely change as the 

event progresses and the known concentration of the radionuclides involved becomes more 

certain and their concentration diminishes due to radioactive decay, dilution, or dispersion. Using 

Radiological Laboratory Sample Analysis Guide for Incidents of National Significance–Radionuclides in Air as an example, four different levels are assessed over the course of an event: 2 

rem, 500 mrem, 10–4 risk, and 10–6 risk. As the event progresses towards samples being analyzed 

at the level of 10–6 risk, the method detection capability may need to improve in order to 

continually and efficiently prioritize samples. The feedback to the IC will be slowed down 

because the decreased sample activities will result in longer screening times for samples and 

longer count times for samples following analytical separations. 

The changing MQOs will have a “domino effect” on laboratory QC analyses, such as spikes, 

duplicates, laboratory control samples (LCSs), and blank samples, processed in a batch. The 

activity levels for spikes and LCSs may become lower as the event progresses, and the 

acceptance criteria for the QC samples also may change. Changes to the required measurement 

uncertainties for these QC samples will require longer counting times and also may slow down 

reporting to the IC. 

The required method uncertainty (uMR) may have default values for each radionuclide and matrix 

(other guides in this series identify these default values; see references in Appendix IV) or may 

have incident-driven values. In either case, the laboratory should be prepared to adjust these 

values when required by the incident MQOs for both the screening instruments and the 

radionuclide specific methods. The value of uMR and the acceptable error rates for Type I and 

Type II errors are used to determine the analytical decision level (ADL). The ADL is a value that 

is less than the AAL. When the ADL is exceeded, it is concluded that the AAL has also been 

exceeded, guarding against a decision error that would allow a sample exceeding the AAL to go 

undetected. The ADL concept is also used for both screening instruments and laboratory-specific 

methods. For more details on these concepts, see Appendix VI to Radiological Laboratory 

Sample Analysis Guide for Incidents of National Significance–Radionuclides in Water (EPA 

2008a). 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

7

II. RADIONUCLIDES 

The list in Table 1 is specifically for an RDD event and the major (non-inclusive) dose-related 

radionuclides that might be released during the detonation of an IND. In the case of an IND, 

numerous short- and long-lived radionuclides will be present, requiring proper identification and 

quantification. Several of the radionuclides on the list have progeny that coexist with the parents. 

Thus, if 228Th were to be found, 224Ra also would be present (although it is not listed). Several 

different radionuclides may be present even if only one RDD is used. 

TABLE 1 – Radionuclides of Concern

Alpha Emitters Beta/Gamma Emitters

Radionuclide Half-Life Emission Type Radionuclide Half-Life Emission Type

241Am 432.6 y ", (, [X-ray] 227Ac[2] 21.77 y $, (

242Cm 163 d " 141Ce[1] 32.51 d $, (

243Cm 29.1 y ", ( 144Ce[3] 284.9 d $, (

244Cm 18.10 y " 57Co[1] 271.7 d ,, (, X-ray

237Np 2.14×106

 y ", (, [(, X-ray] 60Co[1] 5.271 y $, (

210Po [1] 138.4 d " 134Cs[1] 2.065 y $, (

238Pu 87.7 y ", [(, X-ray] 137Cs[4] 30.07 y $, (

239Pu 2.41×104

 y ", [(, X-ray] 3H[1] 12.32 y $ only

240Pu 6.56×103

 y ", [(, X-ray] 125I

[1] 59.40 d ,, (, X-ray

226Ra[2] 1.60×103

 y ", ( 129I

[2] 1.57×107

 y $, (, X-ray

228Th[2] 1.912 y ", ( 131I[1] 8.021 d $, (

230Th 7.538×104

 y ", (

192Ir[1] 73.83 d $, (

232Th 1.405×1010

 y " 99Mo[2] 65.94 h $, (

234U 2.455×105

 y " 32P[1] 14.26 d $ only

235U 7.038×108

 y ", ( 103Pd[1] 16.99 d $, (

238U[3] 4.468×109

 y " 241Pu 14.29 y β, [", γ]

U-Nat[3] --- " 228Ra[2] 5.75 y $ only

 103Ru[2] 39.26 d $, (

 106Ru[2] 373.6 d

$ only, (β, γ

from progeny)

 75Se[1] 119.8 d ε, (

 89Sr[1] 50.53 d $ only

 90Sr[2] 28.79 y $ only

 99Tc[1] 2.11×105

 y $ only

Notes: 

The half-lives of the nuclides are given in years (y), days (d) or hours (h). 

[1] No radioactive progeny or progeny not analytically useful. 

[2] Radioactive progeny with short half-lives, and the progeny may be used as part of the detection 

method for the parent. 

[3] Radioactive progeny not used for quantification, only screening. 

[4] Radioactive progeny used for quantification only, not screening. 

Brackets [ ] indicate minor emission probability. If large quantities of these radionuclides are present, 

these minor emission modes may contribute significantly to any screening measurements made on the 

sample. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

8

Instruments available for screening should provide a consistent measure of sensitivity6 to allow 

detection of as many radionuclides as possible. However, some radionuclides (depending on total 

activity levels) likely will evade detection with routine screening instrumentation (solid scintillators or gas detectors). Generally, those radionuclides that decay by electron capture, positron 

emission, or very-low beta particle emission (and no gamma emission) should be analyzed with 

radiochemical-specific methods to determine their presence. The radionuclides from Table 1 in 

this group are: 3H, 99Tc, 125I, 228Ra, 241Pu, and 106Ru. However, it should be noted that if liquid 

scintillation is used as a screening technique, a measurable response to these radionuclides will 

occur. 

III. DISCUSSION 

The discussion section of the document is divided into five parts. Part A deals with sample 

screening and different instruments that are commonly used to make these measurements. This 

section also provides some insight into technical issues encountered when performing gross 

sample activity measurements when the radionuclide being measured is unknown. 

Part B deals with the calibration of screening equipment and the effects on the calibration 

process as a function of the particle type emitted by the calibration source and its energy It also 

discusses the responses of different types of detectors and provides figures demonstrating detector and sample configurations that may be advantageous for screening of samples for gross 

activity. 

Part C deals with the use of screening equipment for prioritizing samples when the radionuclide(s) present are known. 

 

Part D discusses the MQO process, and Part E provides key recommendations for the laboratory 

in establishing a screening protocol for samples resulting from radiological incidents. 

A. Sample Screening and Processing at the Laboratory 

Guidance on using both the screening instrumentation and the radiation-specific detectors for 

emergency response sample screening is discussed in this section. 

Gross Activity Measurement Instruments 

If the sample screening process at the laboratory is organized properly, it can significantly 

improve the turnaround time for results and minimize risk of the spread of contamination in the 

laboratory, as well as the chance for sample cross-contamination. 

Gross activity measurements can be made using two general types of instrument—a ratemeter or 

a scaler. 

 

6

 In this context, sensitivity refers to the ability of the screening equipment to detect different particles. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

9

The ratemeter measures the radiation emission per unit time in real time, but not all instruments 

have a summation function that would allow total decays to be measured over a defined time 

period. The overall sensitivity and ability of these instruments to discriminate radiation types are 

generally low. Although these are portable instruments that are often used for general area 

surveys, for the purposes of this guide these instruments are used in a fixed geometry relative to 

the samples. These instruments have time constants whose duration can be changed so that an 

average response to general radiation measured is more easily determined. A shorter time 

constant display has more frequent readings with the subsequent result of a “jumpy” needle or 

scale display when activity levels are close to the background level. By increasing the time 

constant, these measurements are averaged out internally and the display becomes more constant. 

This is more of a benefit to the application where the sample and the detector are in a fixed 

juxtaposition. When using a ratemeter for assessing gross radiation levels, it will be necessary for 

the laboratory to establish a protocol to determine the measurement value when meter/display 

readings are not constant (e.g., average the values of the high- and low-meter readings during a 

20-second observation). 

The scaler measures individual events and records them during a specified time period. 

Instrument outputs are generally in terms of total counts. The assessment of the gross activity 

generally takes longer with the scaler than with the rate meter, but the interpretation of the values 

obtained is somewhat more definitive. Some of these instruments have modest energy 

discrimination capabilities. However, these capabilities are severely limited when a mixture of 

radionuclides of varied decay modes is present. Laboratories should have a protocol that 

describes how to use the gross count data obtained by these types of instruments. 

Table 2 identifies general descriptions of gross activity measurement instruments and laboratory 

screening instruments that can be used for sample screening and specific emission types to which 

they are most sensitive. 

TABLE 2 – Detectors Used for Gross Sample Screening 

Type of Detector Sensitive to: 

Geiger-Mueller (GM) Detector 

[Ionizable Gas] 

Gamma (X-rays) 

Open-end GM Detector 

[Ionizable Gas] 

Beta, Gamma, X-rays 

(some high activity alpha) 

GM Pancake Style Detectors 

[Ionizable Gas] 

Beta and Gamma 

(some high activity alpha) 

Micro-R meters [NaI(Tl)] Gamma and X-rays 

Cylindrical Probe [NaI(Tl)] Gamma and X-rays 

Thin Window (Alpha Scintillator) Alpha 

Thin Window (Beta Scintillator) Beta (low response to photons) 

Dual Phosphor Detectors 

First Layer [ZnS] 

Second Layer [Organic] 

Alpha and Beta 

Portable Gamma Detectors [HPGe] Gamma (X-rays) 

Small Article Monitors [NaI(Tl)] Gamma (X-rays) 

Small Article Monitors [Organic Scintillator] Beta and Gamma (X-rays) 

Liquid Scintillation [Liquid Fluor] Beta, Alpha, and Gamma 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

10

Instrument Response Characteristic Determination

The first factor to consider when performing a sample survey is the actual response by the 

instrument to the potential radionuclides plus any decay progeny in the sample. Not only is the 

response of these instruments different for each type of radiation, but it may also vary in a 

complex way with respect to the energy of decay. A couple of examples that demonstrate these 

differences in response are: 

• The response of a NaI(Tl) micro-R meter will be different for high-energy photons 

compared to low-energy photons (i.e., it over-responds to low-energy photons). 

• A GM pancake detector will respond to both alpha and beta radiation. However, for equal 

activities of 32P (beta-emitter) and 242Cm (alpha-emitter), the instrument will yield a 

greater response (i.e., higher counts per minute) from the betas of 32P. 

• An open-end GM detector will respond to both beta and gamma radiation. However: 

o The response to 10 nCi of 89Sr (Eβmax at 1.49 MeV) will be greater than that for 10 

nCi of 99Tc (Eβmax at 0.294 MeV). 

o The response to 50 nCi of 137Cs (gamma energy 0.662 MeV) will be smaller than 

that for 50 nCi of 57Co (gamma energies at 0.136 and 0.122 MeV) because the 

lower energy gamma rays interact more favorably due to the photoelectric effect. 

These examples illustrate that the type and energy of radiation, as well as branching ratios, 

abundance values, and other physical properties of the radionuclide and the detection system are 

significant factors in assessing the total activity of a sample during the screening process using 

survey meters when the exact types of radionuclides present are unknown. Radionuclide-specific 

detection parameters are explained in detail in Knoll.7

Crosstalk: Detector Responses to Radioactive Emissions

In addition to the individual particle energy providing a different response in a particular 

detector, one type of particle may yield a response indicative of another type of particle. This is 

particularly true with gross alpha-beta detection devices that rely on pulse size to determine 

whether an individual event represents an alpha, beta, or gamma detection. 

One instance of this type of incorrect identification occurs with measurement of 241Am using a 

gas proportional detector. Although 241Am is principally an alpha-emitter, it also emits a lowenergy photon at 59 keV. A photon of this low energy may yield a response in the beta channel 

because of the high probability of secondary interaction of scattered radiation with the instrument components (including electronics, detector casing, instrument housing) via the photoelectric and Compton effects. Thus, if the total activity of the 241Am is high, an incorrect 

assumption regarding beta activity could be made. 

Care must also be used to evaluate and interpret the results with respect to possible beta-to-alpha 

and alpha-to-beta crosstalk effects when screening air filters (or other solid materials) for gross 

alpha and beta activities by instruments using gas proportional counting. The type of effect 

 

7

 Knoll, Glen F. 1979. Radiation Detection and Measurement, New York: John Wiley and Sons, Inc. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

11

depends on the instrument mode of operation, setup, voltage plateaus, and discriminator settings. 

For most modern gas proportional counting instruments, the mode of operation may include: 

 

1) Simultaneous measurements of alpha and beta activities based on a single operating 

plateau and beta-to-alpha and alpha-to-beta discriminator settings; or 

2) Independent analysis of alpha and beta plus alpha activities on two separate voltage 

plateaus. 

The second mode of operation, for most practical purposes, eliminates the beta-to-alpha crosstalk 

effect. However, the alpha response on the beta plateau must be estimated and the beta results 

adjusted accordingly. The remainder of the discussion that follows here will address the simultaneous alpha- and beta-counting mode. 

The instrument voltage discriminator setting8 should be adjusted when operating in the simultaneous alpha and beta counting mode to maximize the alpha detector efficiency and minimize the 

beta-to-alpha response crosstalk. These settings should be established using a source with matrix 

characteristics similar to the samples received from the incident response since absorption of the 

alpha particles in the matrix will decrease the alpha energy available with a proportional decrease 

in the signal voltage for processing. Typically, nominal instrument settings can be established 

that allow for an acceptable alpha counting efficiency and a beta-to-alpha crosstalk of <0.1 %. 

However, depending on the sample matrix and instrument settings, the actual crosstalk value can 

vary widely from this value. For air filter matrices, the alpha detector efficiency may be as low 

as 5 to 10%, and the beta-to-alpha crosstalk may contribute significantly to this value. 

When evaluating gross alpha and beta activity results of sample analyses for the purpose of 

sample prioritization (for subsequent radionuclide-specific analyses), it is important to consider 

the possible effect of the beta-to-alpha crosstalk on deciding if the instrument alpha results have 

been artificially increased. The beta-to-alpha crosstalk effect may be most important either 

during the initial phases of an incident (when the radionuclides of interest are unknown) or when 

the composition of the mixture of alpha- and beta-emitting radionuclides is known. For the latter 

case, the beta-to-alpha crosstalk effect should be addressed. This can be done, once the 

radionuclides have been identified, by performing instrument calibrations for crosstalk using the 

actual radionuclides of concern, and corrections can be made that are both accurate and of known 

uncertainty. 

A general observation of the AALs for those alpha- and beta-emitting radionuclides identified in 

Table 1 indicates that the AALs for the beta-emitting nuclides are at least a factor of 500 or 

greater than for the alpha-emitting nuclides. For example, the 500 mrem AALs for 90Sr and 137Cs 

are 110 and 550 pCi on the air filter for a 68 m3 air sample. For the same dose and volume 

sampled, the AALs for 241Am and 239Pu are 0.17 and 0.14 pCi. For gross screening sample 

prioritization, the AALs for the 90Sr and 241Am should be used. Note that when the actual betato-alpha crosstalk discrimination is 0.1%, the alpha response observed from 90Sr activity at the 

AAL may be > 0.1 cpm. With an alpha detector efficiency of 10%, the reported activity would be 

 

8

ASTM International (ASTM D7282-06). Standard Practice for Set-up, Calibration, and Quality Control of 

Instruments Used for Radioactivity Measurements, ANNEX X2. West Conshohocken, PA. Available for purchase 

from: www.astm.org/Standards/D7282.htm. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

12

near the gross alpha screening AAL. Therefore, when evaluating gross alpha results when the 

beta result is greater than 500 to 1,000 times the alpha result, care must be taken to avoid the 

false conclusion that the screening alpha AAL has been exceeded. When screening air filters that 

have a beta-emitting radionuclide whose AAL is greater than the 90Sr AAL, the beta-to-alpha 

crosstalk effect may be greater (depending on the beta particle energy), and the gross alpha 

screening AAL may be artificially exceeded more often when the radionuclide beta activity is 

near its own AAL. 

As an example, suppose 90Sr at the 500 mrem AAL (110 pCi/m3) had deposited on an air filter. 

The activity would be the sum of 90Sr + 90Y = 220 pCi/m3 for a 68 m3 sample9 (a total activity of 

3.32×104 dpm). The measured beta activity for a 30% efficient detector would be 

 

beta dpm = 0.3 × 3.32×104 dpm beta = 9.96×103 cpm. 

The alpha response from beta-to-alpha crosstalk would be based on the crosstalk factor, which is 

relatively small (about 0.1%). Thus, the apparent alpha activity counted would be 

cpm = 9.96×103

 × 0.001 = 9.96 cpm. 

Alpha background on a GPC will be small at ~ 0.05 cpm. Thus, with an alpha efficiency of 0.1 

(10%), the net count rate for alpha would yield a calculated alpha activity of 

alpha = (9.96-0.05) / (0.1) = 99.1 dpm = 44.7 pCi. 

This would yield a false indication of alpha activity when none is present. 

Using the same reasoning, example AALs can be applied to the evaluation of air filters with 

elevated alpha activity. The effects of alpha-to-beta crosstalk (versus beta-to-alpha crosstalk) can 

be calculated, and the potential impact on artificially exceeding the beta AALs can be determined. 

When operating a gas proportional counter in the simultaneous alpha and beta counting mode, 

the initial adjustment of the voltage discriminators is intended to minimize the beta-to-alpha 

crosstalk. Crosstalk, however, is more dependent on the specific radionuclide present in the 

sample and its physical decay and emission properties, than on the instrument discriminator 

settings. Actual alpha-to-beta crosstalk can vary from less than 3% to more than 30%, depending 

on the radionuclide and other factors. 

Alpha-to-beta crosstalk correction factors should be determined during the initial instrument 

efficiency calibrations. These factors can be useful in making corrections to the beta count rate, 

based on the alpha count rate, but only when the radionuclide present has been correctly 

identified and the instrument has been calibrated accordingly. 

 

9

 The volume of 68 m3 is used as a reference volume as described in the Radiological Laboratory Sample Analysis 

Guide for Incidents of National Significance–Radionuclides in Air (2009, In Preparation). 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

13

When performing gross screening analyses, however, where the radionuclide has not been 

identified and the instrument has not been appropriately calibrated, making a crosstalk correction 

based on the initial instrument calibration can result in significant errors in the measurement of 

the beta activity in the sample. Depending on the project MQOs and event circumstances, it may 

be preferable to make no crosstalk correction and to potentially overestimate the sample beta 

activity. Because the beta AALs are typically much higher than the alpha AALs, this overestimate should result in artificially exceeding the beta AALs only when the alpha activity is 

extremely elevated. 

In the previous example for beta-to-alpha crosstalk, the 500-mrem AAL for 90Sr is 110 pCi for a 

68 m3 air sample. An 241Am activity of 2,200 pCi would be required to yield a beta channel 

signal that would correspond to the 110 pCi activity for 90Sr, or nearly 13,000 times the 241Am 

AAL. For other alpha-emitting radionuclides, the alpha activity required to cause this beta AAL 

to be artificially exceeded could be greater than 100,000 times the AAL of that other radionuclide. 

In these unusual cases, the apparent beta activity should be confirmed by an appropriate technique, such as recounting the sample with an alpha-attenuating barrier in place and comparing 

the beta count rates from the two analyses. For screening analyses, however, these techniques 

assist only in estimating the degree of bias in the results, and do not correct for all sources of 

crosstalk. 

This effect can be illustrated by calculating the quantity of alpha activity from 241Am that would 

yield an indication of beta activity at the AAL for 90Sr. Given the 10−6

 risk AAL for 90Sr of 0.29 

pCi/m3 and an assumed sampled air volume of 68 m3: 

• A beta activity from (90Sr + 90Y) on a filter at the AAL would be approximately 88 dpm; 

• The beta counts recorded (with a detector efficiency of 30%) would be ~26 cpm beta; and 

• A normal beta background of 1 cpm yields a net beta count rate of ~25 cpm. 

Assuming 30% alpha-to-beta crosstalk and 10% counting efficiency for 241Am, the alpha activity 

required to produce alpha-to-beta crosstalk equivalent to the 90Sr AAL would be 

Alpha activity = 26/(0.3×0.1×2.22), or approximately 390 pCi. 

Thus, an activity of 241Am of 390 pCi can cause an apparent beta activity equivalent to the AAL 

of 90Sr even when there is none present. 

Detector Background

A second factor to consider during sample screening is the background. Background can be 

divided into the categories of instrument (intrinsic or electronic), environmental (laboratory 

location), and sample container/sample. These should be minimized when possible to achieve the 

best signal to background ratio for the sample. As will be shown further on in this document, 

reduction of background is one of the most important limiting factors for detection of low level 

sample activity during the screening process. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

14

Some examples of potential background concerns are: 

• Proximity of one screening instrument to another when samples or groups of samples 

contain enough activity to have an impact on a neighboring instrument. 

• Presence of radionuclides with multiple emissions that can be detected by the instrument. 

Since the level of background is crucial to the measurement, the shielding of the detector is an 

important consideration. 

Sample Geometry

The third factor that should be considered when using survey meters is the consistency of the 

sample-to-detector geometry. The method of calibration of the survey meter and the method used 

to screen samples using the survey meter should match as closely as possible to obtain the best 

estimate of absolute activity in the samples. 

Finally, sample self-absorption should be evaluated when assessing the results of sample 

screening. This effect is most critical with alpha- or beta-emitters, but for low-energy photonemitters it also will be a contributing factor to misidentification of particles. The loss of particle 

energy as it travels through the sample medium will cause it to yield a smaller ionization pulse in 

the detection device. As described earlier, this can register a false count for the wrong type of 

emitted particle. 

Each of these three factors will be considered in the sections below that address the calibration of 

screening detection equipment. 

Laboratory Instruments 

Hand-held devices are not the only types of instrumentation that can be used for performing a 

gross radiological screen on a sample. Consideration should also be given to using three 

mainstays of the radiochemical laboratory for screening analyses. Gas proportional counters 

(GPC), NaI(Tl) detectors, and liquid scintillation counters (LSC) normally are used for 

radionuclide-specific analyses, and in such applications radiochemical purity of the sample test 

source (STS) is imperative. These instruments can be used to assess total activity as well. This 

may require a modification or re-configuration of laboratory instrumentation to dedicate some 

portion of the laboratory resources to emergency response rapid screening. 

B. Calibration of Instrumentation for Screening Analyses 

Detector Type

Examples of different types of gross screening survey meters and laboratory screening instruments are summarized here: 

• Gross Alpha 

o ZnS(Ag) scintillation detector with a thin aluminum or MylarTM window
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

15

o Open-end GM detector 

o Gas-filled pancake probe with a thin window 

• Gross Beta 

o Plastic organic scintillator with a thin aluminum or Mylar window 

o Gas-filled GM detector (with slide-window allowing gamma detection in the presence 

of beta) 

• Gross Gamma 

o Gas-filled GM detector 

o Sodium iodide (NaI[Tl]) or cesium iodide (CsI[Tl]) detector (well or flat type crystal) 

with scaler for open discrimination counting 

o Micro-R meter using NaI(Tl) or CsI(Tl) detector 

o HPGe detector (may be flat or well type) set for gross counts using summation of all 

channels 

It would not be practical to maintain calibrations for each of the radionuclides, or mixtures of 

radionuclides, shown previously in Table 1. However, a straightforward process can be 

performed to relate the response of each detector to decay particle energy. While the 

measurements are not as precise as more extensive laboratory measurements, it allows increased 

accuracy for a longer list of radionuclides when making an estimate of the total activity. This can 

be accomplished by selecting at least two (but preferably three or more) radionuclides that emit 

characteristic decay particles with distinct energies that span the usable range of the instrument. 

Table 3 identifies a list of radionuclides that can be obtained as standards for calibration of 

detector energy. Their emissions and energies for calibration are also included. 

Table 3 – Radionuclides Spanning the Energy-Calibration Range 

Radionuclide 57Co 60Co 137Cs 99Tc 90Sr/90Y 230Th[1] 241Am

Emission Type γ γ γ βmax βmax α α

Energy, MeV 0.122, 0.136 1.173, 1.332 0.662 0.29 0.545, 2.28[2] 4.69 5.49 

[1] This is the primary alpha for thorium; thorium has progeny that emit alphas as well. 

[2] This energy belongs to 90Y, which is in secular equilibrium with the 90Sr. 

Next, the net instrument response for each of the radionuclides is measured in a standard 

configuration (i.e., a “geometry”: matched quantities of sample, containers, and position relative 

to the active volume of the detector). For each type of decay particle and geometry, instrument 

response should be plotted against the average decay energy10 of the particle emitted. Using 

these data, a table of response factors (i.e., efficiencies) is prepared that correlates to each of the 

radionuclides in Table 1 based on decay type and respective average decay energy. An example 

of this application can be seen in Figure 1, which shows the energy response to different energy 

gamma radiation for a halogen quenched GM detector, and in Figure 2 for a NaI(Tl) detector. 

Note the significant, relative effect that using the GM shield has on the detection of the lowerenergy versus the high-energy gamma emitters. This also can be used in a qualitative sense to 

assess the overall energy profile of the gamma emitters. 

 

10 See example in Appendix III for 192Ir. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

16

FIGURE 1 – Halogen-Quenched GM Detector Response to Gamma Radiation (A) with 

Shield Open (B) with Shield Closed 

The maximum in detector response for the commonly used NaI(Tl) detector is about 100 keV 

(see Figure 2). For a comparable sized CsI detector, the response would be more efficient 

overall, and the maximum in the efficiency curve would be at a slightly higher energy. This is 

due to the difference in physical properties of the CsI(Tl) crystal material. 

FIGURE 2 – Gamma Energy Response for a Na(Tl) detector 

The response for an alpha-beta survey meter,11 using a halogen quench fill gas and a thin mica 

window pancake probe, may have the following characteristics: 

• Efficiency (2-pi geometry): 5%–14C; 22%–90Sr/90Y; 19%–99Tc; 32%–32P; 15%–239Pu 

 

11 The response curves and characteristics for these instruments were taken from information provided by Ludlum 

Measurements, Inc., at www.ludlums.com. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

17

• Sensitivity: Typically 3,300 cpm/mR/h (137Cs gamma) 

• Energy Response: Energy dependent 

From these few examples, it can be seen that the response of a survey instrument to different 

types and energies of radiation is a complex function of not only the radiation emitted but also of 

the survey instrument used. 

Geometry

The relative geometry of sample to screening instrument and shielding can take on several 

different configurations. It is very important to ensure that the sample measurement matches the 

calibration geometry. Some of the considerations that will affect the optimal configuration of 

sample to detection device are: 

• Shielding (detector). The detection capability of the screening method will be optimized 

by shielding the detector to reduce ambient background and minimize response to 

external sources of radiation. The detector and detector shielding configurations should 

remain fixed so that the background count rate is reasonably constant. 

• Shielding (container). The sample container material can be made of glass, polyethylene, 

Teflon, or other non-reactive material. The effect that these different materials have on 

shielding the radioactive emissions from the detector varies with particle type and energy. 

Also, the thickness of the container walls can increase the average distance of the centerof-activity of the sample to the detector. Both of these sample container characteristics 

can affect the net screening result. 

• Volume/shape/density. The sample volume must be consistent with gross measurements 

made during the calibration of the screening equipment so that the relative configuration 

of sample-to-detector is maintained. Thus, it is important that the sample container be 

virtually identical to the container used for calibration purposes. Sample density (or for 

solids, the degree of compaction) has a significant effect on the potential self-shielding of 

the sample from the detector. The mass of the calibration source and the sample should 

be relatively close in value to achieve consistent configuration. 

• The figure of merit12 (FOM) for the configuration of the shielding may need to be 

optimized (i.e., a larger FOM is better). For example, it may be advantageous to have a 

relatively large shielded volume with the sample centrally located, versus a shielded 

volume that exactly fits the sample geometry. 

• Location of the sensitive detection area in the screening equipment. The manufacturer’s 

detailed diagram for the specific model of screening equipment should be available so 

that the optimum position of the detector with the sample can be achieved (See Figure 3). 

• Size and shape of the detector with respect to the sample geometry. The sample shape 

and detector juxtapositioning can have significant effects on the measurement. One 

measure of this is the FOM. 

An example illustrating the effects of the size and shape of the detector on the FOM can be seen 

in Table 4, which identifies some data taken using NaI(Tl) detectors of various sizes (none of 

 

12 FOM = [(detector efficiency)2

/background] (Mann et al., 1991). 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

18

these were well detectors). The sample container was a 1 liter plastic bottle. The data were 

recorded using a detector and shielding as shown in Figure 3. The configuration of the detector 

and shielding actually used in this case was not optimal: In the bottom orientation position, the 

detector is partially unshielded, and the flat surface of the NaI(Tl) detector can is facing the 

sample bottle. Figure 3 also shows the side orientation where again the detector is partially 

unshielded, and the curved detector cover is parallel to the sample. Also, note the actual position 

of the detector crystal in both cases. It is clear in either case, however, that detector size and 

positioning with respect to the sample will have a significant effect on the measurement 

sensitivity (based on the FOM). 

TABLE 4 – Response and Figure of Merit for 60Co and 137Cs with Different NaI(Tl) Detector 

Configurations 

Radionuclide 

Activity 

pCi/L 

Net 

cpm 

Figure of Merit 

 1"×1" 2"×2" 3"×3" 1"×1" 2"×2" 3"×3" 

Background (BO) – 2.80×102 1.65×103 2.4×103 – – – 137Cs (BO) 5.038×105

 2.22×103 1.34×104 1.3×104 6.93×10–8 4.28×10–7 2.77×10–7

60Co (BO) 3.317×104

 5.5×102 9.5×102 7.5×102 1.67×10–7 4.97×10–7 2.13×10–7 

Background (SO) – 2.7×102 1.1×103 1.75×103 – – – 137Cs (SO) 5.038×105

 5.03×103 1.66×104 1.78×104 3.69×10–7 9.87×10–7 7.13×10–7

137Cs (SO) 5.038×104 3.1×102 9.0×102 6.0×102 1.4×10–7 2.90×10–9 8.1×10–8

137Cs (SO) 1.242×104 1.5×102 3.0×102 1.2×102 5.40×10–7 5.30×10–7 5.33×10–8

60Co (SO) 3.317×104 5.8×102 2.2×103 1.5×103 1.13×10–6 3.99×10–6 1.17×10–6

Notes: 

SO = Side Orientation (see Figure 3) 

BO = Bottom Orientation (see Figure 3) 

Example Calculation: For the Cs bottom orientation (BO) and the 1"×1" detector 

 FOM = [net cpm/pCi/L]2

/[Background] = [2.22×103/5.038×105]

2

/(2.8×102) = 6.93×10–8 

The data indicate that the biggest detector volume does not always give the highest count rate, 

nor does it always yield the highest value FOM. Thus, it is imperative that the detection 

equipment used be assessed in a similar fashion to determine which screening equipment is best 

suited for each combination of matrix and geometry. Two factors to be considered in determining this are: 

• Location of the mean sample activity relative to the location of the detector, and 

• Shielding (covering) of the screening equipment. 

There are different considerations for samples that need to be screened for gamma radiation. An 

example is using a NaI(Tl) well detector. Many different sample types can be accommodated 

into this well for screening purposes. For example, a 47-mm air particulate filter may be rolled 

and inserted into a container, such that the container will fit reproducibly into the well of the 

NaI(Tl) detector, improving overall efficiency for detection. When doing this, care must be taken 

to avoid contaminating the detector. That specific geometry for calibrating this style of detector 

can be accommodated by most laboratories. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

19

FIGURE 3 – Shown Without Bricks Covering Top of Shielded Geometry. NaI(Tl) Detector Example 

Figure 4 shows another way to configure 

the detector and the sample bottle to 

achieve a better FOM for the measurement. 

In this configuration, the active area of the 

NaI(Tl) detector is inside the shielding and 

thus has a lower net background from room 

and ambient background contributions. 

Figure 5 shows two different configurations 

of shielding with respect to the detector that 

will provide different backgrounds. Note 

that the thickness of the shielding walls is 

the same but that the internal cavity in 

which the detector is held is larger in Figure 

5B. The larger volume ultimately leads to a 

better FOM since any Compton scattering 

from the shielding in 5B will impinge to a 

lesser degree on the detector than in 5A 

solely due to distance. In Figure 5B, a 

sample stand has been added to put the 

sample in the middle of the shielded 

volume, and the detector has been raised slightly to yield the same orientation as in 5A, thus 

maintaining the same detector efficiency. 

FIGURE 4 – An Improved Orientation for Shielding. 

Active Detector Area Within Shielding. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

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FIGURE 5 – Relative Size of Shielded Volume 

Figure 6 shows a configuration for a pancake-style screening instrument (could be gross alphabeta or beta-gamma). The air particulate filter is slid into place beneath the detector, which is 

maintained in a fixed position using a small stand. The presence of shielding allows reduction in 

background for the detector and for the sample, and provides a fixed geometry for consistent 

results. 

FIGURE 6 – Sample Shielding and Detector Orientations for Gross Screening of Air 

Particulate Filters Using an Alpha/Beta Pancake Detector 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

21

Crosstalk, Dead-Band, and Self-Absorption Factors

The degree of crosstalk as determined under routine instrument calibration conditions may not be 

significant. However, when the activity being measured is two and three orders of magnitude 

greater than normal sample test sources, crosstalk that was once obscured in the background may 

provide a signal that is indicative of a particle type that is absent. Thus, it is important to 

challenge the screening instrumentation with standards of high activity so that the level of crosstalk can be assessed. One such application involves GPC systems that are simultaneously 

counting gross alpha and beta activity. An assessment of crosstalk should be made in the beta 

channel response when the alpha activity is large compared to the beta activity and compared 

with the same beta activity response with no corresponding alpha activity. The inverse 

assessment should also be made. These measurements may lead the lab to apply a “dead band” 

between the lower level beta and upper level alpha discriminator settings that normally would 

not be used. This dead band would minimize the crosstalk, but would also lower the efficiency 

for both types of particles. Thus, the use of a dead band should be used judiciously to avoid 

abnormally long count times when screening time is at a premium. 

It should also be recognized that elevated activity of radionuclides that decay only by beta 

emission may result in counts above background when using a sodium iodide detector for gross 

count assessment (e.g., as when using a small article monitor). The bremstrahlung radiation, 

emitted as a result of the beta interaction with matter, yields low-energy photons that produce a 

signal in the sodium iodide detector. 

Self-absorption factors are significant for alpha- and beta-emitters. Determining how sample 

mass affects the efficiency of detection can be estimated using calibration sources and absorbing 

materials of known areal density (measured in units of mg/cm2

) placed between the sample and 

the detector. This intervening material would simulate the sample mass when the sample is not 

ideal (i.e., the sample is not “massless” and will absorb some of the contained radiation). This 

mass attenuation correction for self-absorption is similar to determining unknown beta particle 

energy using the Feather Method.13 For alpha particles, this may mean using a thin film of 

aluminized Mylar, while for betas, varied thicknesses of aluminum metal may be used. The areal 

density effect for each detector should be semi-quantitatively identified so that estimates of 

activity correction can be made when samples of observable mass are measured using detection 

techniques such as GPC. 

Final Instrument Calibration and Method Validation

Once the detectors to be used for screening have been selected and the considerations for sample 

to detector configuration and efficiency have been assessed, a method should be written. The 

 

13A technique that has been used successfully to determine the energy of beta-only emitters is to measure the range 

of the beta particles in a pure material (“Feather analysis”). The ranges of beta particles in several pure materials 

(such as aluminum) have already been established. The units of thickness are expressed as areal density, or mg/cm2

. 

A set of aluminum absorbers of varying thickness is used, and the activity versus the absorber thickness is plotted on 

a semi-log scale. The linear portion of this curve is then extrapolated to find the “zero” activity thickness. This is 

then related to the Eβmax of the beta particle, which will be characteristic for a particular radionuclide. A discussion 

of this technique can be found in Chase, G.D. and J.L. Rabinowitz (1967). Principles of Radioisotope Methodology, 

3rd Edition. Minneapolis: Chase and Burgess. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

22

method should incorporate the laboratory’s best estimate of the potential geometries and 

plausible radionuclides into the procedure. Specific instructions regarding the receiving and 

storing of the samples, recording of data, and sample aliquanting for particle-specific screening 

should be included in this method. Once the method is written, a method validation process that 

follows the Method Validation Requirements for Qualifying Methods Used by Radiological 

Laboratories Participating in Incident Response Activities (EPA, 2009b) should be followed. 

The method validation process requires the use of proficiency test samples to validate the 

detector response to achieve the MQOs established for the project or by the laboratory. Once the 

method has been validated, the procedure should be implemented routinely for sample 

processing by all staff members, which will reinforce training on the procedure. 

C. Calibration of Screening Instruments when Radionuclide Identities are Known 

Screening equipment that is calibrated for overall response to decay particles will have its 

accuracy challenged if the radionuclide in the sample to be measured has a different particle or 

energy. 

During the initial phases of an emergency, before the identity of the radionuclide(s) associated 

with the event has been established, a response factor for the screening equipment presumably 

will be based on a single radionuclide, such as 137Cs. As the radiological event progresses, the 

radionuclide(s) associated with the event should be identified. For example, if 192Ir is identified, 

the factor used to convert cpm/sample to pCi/sample should be changed so that the screening 

equipment more accurately characterizes the sample activity level, and the laboratory will be 

able to characterize the activity of the samples more accurately. This change in the response 

factor can be implemented in several ways: 

1. The laboratory has already established a response factor on the screening equipment for 

this radionuclide in this geometry. In this case, receipt instructions need to be updated to 

include the identity of the radionuclide(s) of concern. For example, consider a beta/gamma 

survey meter that has been calibrated with a 137Cs source that had a measured response 

factor for a 1-L liquid sample of 5.1×10–4 mR/h per pCi. This factor has been entered into 

the electronic database for the meter used (identified by serial number). Knowing now that 

the radionuclide of interest is 192Ir, with a response factor of 2.8×10–4 mR/h per pCi, this 

response factor should replace the 137Cs value currently present in the electronic database 

.

14 This change will identify more accurately the activity based on gross screening 

measurements. 

2. The laboratory has performed an energy response factor curve for the screening equipment 

and can interpolate the curve for the effective mean emission energy of the radionuclide 

present in the sample. For example, this method is demonstrated in Figure 7 for a 47-mm 

air particulate filter using simulated data. 

 

14 The date of this change and the reason for the change need to be noted in the instrument data files. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

23

FIGURE 7 – Survey Meter 12345 Energy Open Window Response Curve for Beta Emitters 

In Figure 7, the radionuclide energies represented are approximately one-third of Eβmax. 

Thus, as an example, the effective beta particle energy for 14C is 156 keV/3 = 55 keV.15

3. The laboratory may take a sample that is to be analyzed, and determine a conversion factor 

based on a comparison of the screening value and radionuclide-specific analysis results. In 

this case, it would likely be best to take an average conversion factor from several samples 

to ensure the most accurate representation of the factor. This is because the factor can be 

affected by non-uniform distribution in the sample. Consequently, the laboratory should 

consider the potential for significant uncertainty in this conversion factor, which may be 

estimated by the standard deviation of the individual measurements used to calculate the 

average conversion factor. 

As an example, a data table like the one below could be constructed. Note that the information 

shown is not based on actual data but is used for illustrative purposes only. 

Table 5 – Screening Instrument Conversion Factor Based on Sample Analysis of a 1-Liter 

Sample Geometry 

Sample 

Screening 

Value, 

(mR/h)/L 

RadionuclideSpecific Analysis 

Results, 137Cs 

μCi/L 

Conversion 

Factor 

[μCi]/(mR/h) 

Estimated 

Conversion Factor 

Uncertainty 

[μCi]/(mR/h) [1]

Background 2 – – – 

Sample 1 55 1,601 30.2 – 

Sample 2 78 2,005 26.4 – 

Sample 3 41 1,448 37.1 – 

Average Conversion Factor: 31 ± 5 

[1] The method used to estimate the screening equipment uncertainty must be decided upon by the laboratory. The 

column is included here so that it is clear that this should be one aspect of this process. 

 

15 It is important to note that the use of this type of curve is not necessary for alpha instruments since the alpha 

response would be mostly independent of energy. 

Beta Response Curve

0

5

10

15

20

25

30

35

0 100 200 300 400 500 600

Energy, keV

pCi/(microR/Hr)
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

24

In this example, the samples have already been screened using a micro-R meter. The samples are 

then analyzed using a radionuclide-specific method, and the values obtained are specifically for 

137Cs. The final analytical values for the samples are divided by the original exposure rate 

measurements to obtain a conversion factor for the radionuclide contained in the event-specific 

samples. The average conversion factor and the associated uncertainty estimate are rounded to 

the appropriate number of significant digits. In this case, the conversion factor would allow the 

laboratory to estimate the concentration of 137Cs in the subsequent samples, based on the microR meter screening results. This simplified example uses a single radionuclide with no ingrowth 

considerations. In cases where one or more radioactive progeny may be present, care must be 

taken to ensure that the screening conditions, especially the degree of progeny ingrowth, are 

reasonably consistent. In all cases, the counting geometry for sample screening should be as 

consistent as possible. 

During the latter phases of an event (when the radionuclide content of the samples is expected at 

the 10-4 risk level for air filters and the maximum contaminant levels for drinking water), the 

screening of lower activity samples may be performed using a different technique. For example, 

if both alpha- and beta-emitters are present, rather than using GPC to screen the samples for both 

alpha- and beta-emitters simultaneously, it may be advantageous to perform each screen 

separately and extend the count time to ensure better discrimination between those samples 

where analysis is required immediately and those that may be delayed. 

D. Measurement Quality Objectives (MQOs) for the Screening Process 

Screening of samples as they arrive significantly impacts the laboratory’s decisions about which 

samples to analyze first. The IC should have decided how the samples are to be prioritized and 

communicated this to the laboratory. The laboratory may confidently screen these samples for 

gross activity so that they can be processed in a timely fashion based on the needs of the 

incident. 

General guidance on how to establish an MQO for the required method uncertainty can be found 

in MARLAP (2004) and specifically for radionuclides in water (EPA 2008a, Appendix VI). 

Additional MQOs for screening should be established by the laboratory based on the type of 

instrumentation available. 

In order to illustrate the typical decisions and actions to be taken by a laboratory for calibration 

and gross sample screening, three examples using theoretical samples and measurement results 

are provided in Appendices I-III. These examples demonstrate an acceptable method for the 

calibration of instruments and measurement of samples, but each example is one of several 

different possible variations of calibration and measurement techniques. The examples here 

should not be construed as limiting. 

The first scenario (Appendix I) illustrates how a laboratory may prepare its screening equipment 

to be ready to receive samples from a radiological incident. The instrumentation and standards 

used are limited to what is available to the laboratory, which demonstrates how some basic 

planning can assist in being prepared for such an event. In the second scenario (Appendix II), the 

same laboratory has received samples from a radiological transportation accident and has been 

asked to rapidly assess the spread and degree of contamination. The calibration of the screening 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

25

equipment is optimized to assess the contamination levels in the samples that have been sent. 

The third scenario (Appendix III) discusses how instrument calibrations may be adjusted during 

the latter phases of an event when the radionuclide(s) identity(ies) is (are) known. In this 

instance, the screening process will be looking at lower overall activity in the samples so that recalibration with the same radionuclide will enhance the detection capability of the screening 

equipment. 

E. Key Recommendations 

Laboratories should be prepared for potential radiological events where large numbers of 

samples at much higher activity concentrations than normal arrive suddenly. To assist laboratory 

personnel in promptly receiving, prioritizing, and analyzing samples, the following is a summary 

of the key recommendations for sample screening: 

• Screening equipment should be calibrated with traceable sources that match geometries for 

anticipated emergency response samples. 

• These calibrations should have associated direct reading conversion factors for ease of 

reporting results in the appropriate units. 

• Laboratories should have written procedures (or instructions) for the process of screening 

emergency response samples. 

• Shielding for the screening equipment should be configured to maximize the signal to 

background ratio, providing the analyst with smaller uncertainties of the measurement. 

• A plan that provides for the calibration adjustment of the screening equipment based on the 

incident radionuclide(s) should be prepared for that time when the activities are much 

lower, and better discrimination between lower level activities will be required. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

26

Appendix I – Screening Instrumentation Initial Calibration 

It is assumed that laboratories will have properly calibrated their instrumentation prior to an 

event. The data provided in the following three scenarios (Appendices I–III) are used for 

illustrative purposes only. Each laboratory should consider using the general techniques modeled 

here for its laboratory-specific methods to be used for sample screening. In addition, uncertainty 

values have been omitted from these examples. For actual calibration and screening, 

uncertainties should always be included in the expression of the final results. 

Background 

ABG Laboratory, Inc., has decided to set aside certain instruments for radiological events where 

sample gross screening will be necessary. A GM pancake detector and an old 3"×3", planar, 

NaI(Tl) detector with a scaler have separate, shielded geometries for samples of 47-mm filters, a 

1-L liquid, and a 250-g solids container. The equipment is to be located near the samplereceiving area of the laboratory facility. Once the equipment is set up, the laboratory staff 

performs background counts on the instruments while waiting for the new calibration sources to 

arrive. The calibration sources are 99Tc, 90Sr, 241Am, 57Co, 230Th, and 60Co. Each source has been 

ordered for each geometry identified above and is traceable to a national standards body, such as 

the National Institute of Standards and Technology in the United States. 

Discussion 

The NaI(Tl) detector was set up to accumulate total counts in a two-minute count. The GM 

pancake detector was set up in rate mode for cpm. The following table identifies the detector 

background and response from the standards for each of the instruments. 

TABLE 6 – Calibration Data for Screening Instrument Response 

Detector 

Radionuclide 

Source 

Total 

Background*

Activity

pCi 

Air Filter, 

Net Counts 

250 g Can, 

Net Counts 

1 L Bottle, 

Net Counts 

NaI(Tl) 57Co 5,840 cpm 8.0×105 8.88×104 5.33×104 2.13×104 

 2.0×105 2.22×104 1.33×104 5.33×103 

60Co 5,840 cpm 8.0×105 1.24×105 9.59×104 3.66×104 

 2.0×105 3.11×104 2.40×104 8.88×103 

GM 

Air Filter, 

Net cpm 

250 g Can 

(open) 

Net cpm 

1 L bottle 

(closed, side 

measurement) 

Net cpm 

Alpha 241Am 0.05 cpm 40 8 0.18 0 

 10 2 0.04 0 

232Th 0.05 cpm 32 6.5 0.14 0 

 8.0 1.6 0.04 0 

Beta 99Tc 0.8 cpm 4.5×103 509 10 0.1 

 1.2×103 136 2.5 0.03 

90Sr 0.8cpm 300 133 67 20 

 80 35.5 18 5.3 *

For the sodium iodide detector, background counts were summed over the energy range of 50 to 2500 keV. For the 

GM detector, the background represents an average measurement performed at several times of the day. Each 

instrument background measurement was made using an empty sample container in the position for sample 

measurement, and the sample plus detector were shielded with 4" of lead brick. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

27

The laboratory staff has made separate calibration factors for low- and high-energy gamma-ray 

emitters. Similarly, for the 90Sr and 99Tc, the efficiency of detection of the 90Sr is much better due 

to a smaller degree of self-absorption in the sample and better penetration of the GM detector 

beta shield when used. The response factors for both the 241Am and the 232Th are the same. The 

laboratory staff has made the following response factor table for its instruments: 

TABLE 7 – Response Factors (RF) for Radionuclides with Respective Screening 

Equipment 

Radionuclide 

Energy, 

keV 

Abundance

Factor* 

Air Filter, 

pCi/cpm 

Open Tuna 

Can, pCi/cpm 

Bottle, 

pCi/cpm 

NaI(Tl) Detector, Gamma 57Co 121, 135 1.003 17.5 29.9 74.8 

60Co 1,173; 1,332 2.0 6.45 8.34 22.5 

GM Detector, Alpha 241Am 5,449; 5,440 1.0 5.01 2.25×102 1.5×105 

GM Detector, Beta 99Tc, 210 1.0 8.83 4.50×102 4.50×104 

90Sr 546; 2,280 2.0 1.13 2.25 7.5 

*The abundance factor is the number of particles that are produced per decay of the radionuclide and can be detected 

by the detector listed. The value for 60Co is 2.0 since it yields two gamma rays for each decay (the gamma rays are 

in full coincidence). For 90Sr, the value is 2.0 since it is in secular equilibrium with its progeny 90Y, also a betaemitter. 

The response factors in the table are calculated as follows: 

Thus, for the air filter geometry on the NaI(Tl) detector for 60Co: 

(net cpm) AbundanceFactor

Source pCi RF ⋅ =

6.45 pCi/cpm (1.24 10 counts/2 min) 2.0

8 10 RF 5

5

Co-60 = × ⋅

× =
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

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Appendix II – Radiological Event Screening for 241Am 

Background 

The date is November 15, and steady winds from the northwest at about 20-25 mph are expected 

through tomorrow. A truck is carrying used 99mTc generators16 and 241Am smoke detectors (as 

the bulk of its shipment, but other radioactive waste materials of smaller volume were on board). 

The truck overturns and slides into a rock embankment, bursting into flames along a small twolane highway between towns, and burns down to the tires. Air sampling equipment has been 

stationed in several locations in both towns and along several roadsides. Air samples are 

expected to arrive at the laboratory by 1800 hours this evening (it is currently 1300 hours). 

Additionally, several hundred soil and crop samples are expected over the next week so that the 

plume can be tracked. 

The IC has requested that the highest activity samples be identified and analyzed first so that the 

recovery phase can focus on: 

• Determining how much material has become airborne, and 

• Cleaning up high activity areas first to remove the bulk of the source term. 

Discussion 

ABG Laboratory, Inc., has been contacted and told to expect the samples shortly. It will be using 

the calibrations it has made for its screening equipment to accommodate the influx of samples. 

Table 8 identifies the sample activity measured for each of the matrices received at ~1800 hours. 

Knowing the truck’s cargo makes use of the calibration factors straightforward. The air 

particulate filters have been transmitted in glassine envelopes, and the soil samples were stored 

in solids (tuna) can geometry with a removable lid. The laboratory has verified that these 

geometries match the geometries it used for its gross screening calibration of the instruments. 

The spreadsheet it is using has the following equations for the analysis: 

• Air Filters 

o Gross Alpha Activity = (meter reading, cpm – 0.05, cpm)×(5.01pCi/cpm) 

o Gross Beta activity = (meter reading, cpm – 0.8, cpm)×(8.83 pCi/cpm) 

o Gross Gamma Activity17 = (Total counts – 5,840 cpm)x(17.5 pCi/cpm) 

• Solids Can 

o Gross Alpha Activity18 = (meter reading, cpm – 0.05, cpm)×(225pCi/cpm) 

o Gross Beta activity = (meter reading, cpm – 0.8, cpm)×(450 pCi/cpm) 

o Gross Gamma Activity = (Total counts – 5,840 cpm)×( 8.34 pCi/cpm) 

 

16 Although the 99mTc (t1⁄2 = 6 hours) and its 99Mo (t1⁄2 = 66 hours) precursor have decayed, the progeny 99Tc has a 

half-life of 2.1×105 y, and will thus be present in the environmental samples exposed during the accident. 17 Note that the energy of 241Am (59 keV) is somewhat lower than that of 57Co (122 and 135 keV) and will be 

significantly affected by the aluminum shielding on the NaI(Tl) detector. 

18 The laboratory homogenized the samples by shaking prior to opening and performing the gross screen. The values 

will be affected due to sample self-shielding. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

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TABLE 8 – Gross Screening Measurement Results from Transportation Incident 

Sample 

Alpha 

GM 

Detector, 

cpm 

α Gross 

Screening 

Estimate, pCi 

Beta 

Open Window 

Probe, cpm 

β Gross 

Screening 

Estimate, pCi 

NaI(Tl) 

Detector, 

cpm 

γ Gross 

Screening 

Estimate, pCi 

Air Filter-1 4.7 23.3 1.77 8.57 5,750 -1,580 

Air Filter-2 0.085 0.175 5.82 44.3 5,900 1,050 

Air Filter-3 0.10 0.25 2.88 18.4 5,880 700 

Soil-1 0.550 113 1.46 297 6,050 1,750 

Soil-2 0.16 24.8 3.9 1,395 8,120 19,000 

Soil-3 0.07 4.5 0.7 -45 6,000 1,330 

 

The laboratory reports back to the IC that the sample results, bolded above, have the highest 

concentrations based on gross screening results, and the analyses for 241Am and 99Tc are in 

progress. The laboratory supervisor queues the samples according to activity. The highestactivity samples are to be analyzed first. The supervisor also notifies the separations chemists 

about the levels of activity they will find in these samples. 

The laboratory protocol has established a limit of 100 pCi per aliquant. Normally, the sample 

size processed is 2.0 g. However, for Soil-2, there is 250 g of sample, and in order to be less than 

100 pCi, only 1.0–1.3 g of sample will be aliquanted for this analysis.19 

 

19 The gross gamma estimate for the entire sample is 19,000 pCi. This gives about 19,000/250g = 76 pCi/g. Taking a 

2-g sample would result in 152 pCi, exceeding the laboratory limit. An aliquant of 1.3 g yields 98 pCi. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

30

NaI Response Curve for Gamma Rays

(Air Particulate Filter)

0.00E+00

2.00E+03

4.00E+03

6.00E+03

8.00E+03

1.00E+04

1.20E+04

1.40E+04

1.60E+04

1.80E+04

2.00E+04

0 200 400 600 800 1000 1200 1400 1600 1800 2000

keV

pCi/cpm

Response pCi/cpm

Appendix III – Screening Instrumentation Response Corrected for Different 

Radionuclide 

Background 

A suspected terrorist event involving explosive devices has occurred. Several different 

radioactive materials suppliers have reported thefts of large quantities of radionuclides in the past 

three months. The missing radionuclides were 210Po and 192Ir. Preliminary evidence from the 

scene of the incident identified the presence of radioactive materials. It is suspected that the 

materials that were reported missing are related to this event. 

Radiochemistry Analysts of America has been contacted to screen, then analyze about 200 

samples a day for 192Ir and 210Po, and any other radionuclides that may be present. The samples 

will be air particulate filters (47 mm) and soil (~0.200 kg). It is Day 1 at 1100 hours, and the first 

sample shipment will arrive at 0600 hours on Day 2. The IC has indicated that the sample 

priority is to analyze those samples with the highest activity first. The laboratory has neither a 

210Po nor a 192Ir source/standard. 

Discussion 

The laboratory has selected a NaI(Tl) well detector to screen the air particulate filter samples for 

the 192Ir. Its current calibration factor used 60Co, but it has a response curve based on energy as 

shown below.20

 

The average energy21 of the 192Ir is approximately 390 keV. This corresponds to a factor of 

1.30×104 pCi/(net cpm) as estimated using the curve in Figure 8. A similar curve was made for 

the solid geometry and a response factor of 6.1×103 pCi/(net cpm) for 192Ir was estimated. 

 

20 Calibration points for the curve were 88, 320, 662, 1115, and 1836 keV. The standards were counted for 5 minutes 

each in a shielded geometry. The standards used were individual radionuclides (i.e., not a mixed gamma ray source). 21 Ir-192 has several different gamma rays. The average energy per decay event is approximately 390 keV based on 

the sum of the gamma ray abundances multiplied by their respective energies. 

FIGURE 8 – Gamma Energy Response Curve for a NaI(Tl) Detector 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

31

The laboratory staff is using a GM pancake-style detector for alpha screening of the samples. 

The corresponding response factors for alpha particles are: 

Air filter Solid, 200 g

5.01 2.25×102 

The following day, several hundred samples are received, and the screening process begins. An 

example dataset is shown below: 

TABLE 9 – Results of Screening Measurement Using Adjusted Response 

Sample ID 

Air 

Filter 1 

Air 

Filter 2 

Air 

Filter 3 

Soil 

Sample 1 

Soil 

Sample 2 Background 

NaI(Tl), cpm 4,630 4,550 4,480 6,100 4,700 4,500 

GM Detector, cpm 2.80 1.56 0.23 0.13 0.14 0.12 

Screening Results 

Gross gamma, pCi 1.7×106 6.5×105 -2.6×105 9.8×107 1.2×106 – 

Gross alpha, pCi 13.43 7.21 0.55 2.25 4.5 – 

Based on the results of these screening measurements, air filter 1 and soil 1 have the highest 

activities and should be analyzed first for 192Ir and 210Po. 
Radiological Laboratory Sample Screening Analysis Guide for Incidents of National Significance 

32

Appendix IV – Additional Sources and References 

U.S. Environmental Protection Agency (EPA). 1992. Manual of Protective Action Guides and 

Protective Actions for Nuclear Incidents. Washington, DC. EPA 400-R-92-001, May. 

Available at: www.epa.gov/rpdweb00/rert/pags.html. 

U.S. Environmental Protection Agency (EPA). 2008a. Radiological Laboratory Sample Analysis 

Guide for Incidents of National Significance–Radionuclides in Water. Revision 0. Office of 

Air and Radiation, Washington, DC. EPA 402-R-07-007, January. Available at: www.epa.

gov/narel/recent_info.html.

U.S. Environmental Protection Agency (EPA). 2008b. Standardized Analytical Methods for 

Environmental Restoration Following Homeland Security Events, Revision 4.0. Office of 

Research and Development, Washington, DC. EPA/600/R-04/126D, September. Available 

at: www.epa.gov/ordnhsrc/sam.html. 

U.S. Environmental Protection Agency (EPA). 2009a. Radiological Laboratory Sample Analysis 

Guide for Incidents of National Significance–Radionuclides in Air. Revision 0. Office of Air 

and Radiation, Washington, DC. EPA 402-R-09-007, June. Available at: www.epa.gov/narel/

recent_info.html. 

U.S. Environmental Protection Agency (EPA). 2009b. Method Validation Guide for Radiological Laboratories Participating in Incident Response Activities. Revision 0. Office of Air and 

Radiation, Washington, DC. EPA 402-R-09-006, June. Available at: www.epa.gov/narel/

recent_info.html. 

U.S. Environmental Protection Agency (EPA). (In preparation). Guide for Radiochemical 

Laboratories for the Identification, Preparation, and Implementation of Core Operations 

Unique to Radiological Incident Response. Revision 0. Office of Air and Radiation, Washington, DC. 

Mann, W.B., A. Rytz, and A. Spernol (1991). Radioactivity Measurements: Principles and 

Practices. Pergamon Press, p. 65 

Multi-Agency Radiological Laboratory Analytical Protocols Manual (MARLAP). 2004. EPA 

402-B-04-001A, July. Volume I, Chapters 6, 7, 20, Glossary; Volume II and Volume III, 

Appendix G. Available at: www.epa.gov/radiation/marlap.