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DOE Review Release 09/20/2016

Site Profile for Pacific Northwest National 

Laboratory

ORAUT-TKBS-0027 Rev. 02

Effective Date: 09/12/2016

Supersedes: ORAUT-TKBS-0027-1 Rev 01, 

ORAUT-TKBS-0027-2 Rev 01, 

ORAUT-TKBS-0027-3 Rev 01, 

ORAUT-TKBS-0027-4 Rev 01, 

ORAUT-TKBS-0027-5 Rev 01, 

ORAUT-TKBS-0027-6 Rev 01

Subject Expert(s): Robert E. Burns Jr.

Document Owner 

Approval: Signature on File

Robert E. Burns Jr., Document Owner

Approval Date: 08/23/2016

Concurrence: Signature on File

Christopher M. Tornes, Objective 1 Representative

Concurrence Date: 08/25/2016

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Representative

Concurrence Date: 08/25/2016

Concurrence: Signature on File

James P. Griffin, Deputy Project Manager

Concurrence Date: 08/24/2016

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 09/12 /2016

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 2 of 57

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

09/12/2016 02 Revision initiated to modify the scope of the PNNL site profile to 

reflect the decision to make no distinction between PNNL and 

Hanford until the year 2005. Combined the six technical basis 

documents into one site profile document for the period from 2005 to 

present. Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Robert E. Burns, 

Jr.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 3 of 57

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

1.0 Introduction .............................................................................................................................. 8

1.1 Purpose........................................................................................................................ 9

1.2 Scope........................................................................................................................... 9

2.0 Site Description...................................................................................................................... 10

2.1 Introduction................................................................................................................. 10

2.1.1 Purpose ....................................................................................................... 10

2.1.2 Scope........................................................................................................... 10

2.1.3 Background.................................................................................................. 10

2.2 Radiological Facilities on the PNNL Main Campus ..................................................... 10

2.2.1 Environmental Molecular Sciences Laboratory............................................. 12

2.2.2 Life Sciences Laboratory II........................................................................... 12

2.2.3 Physical Sciences Facility ............................................................................ 13

2.2.3.1 Building 3410 Material Sciences and Technology 

Laboratory................................................................................. 13

2.2.3.2 Building 3420 Radiation Detection Laboratory........................... 13

2.2.3.3 Building 3425 Ultra-Low Background Counting Laboratory........ 14

2.2.3.4 Building 3430 Ultra-Trace Analysis Building .............................. 14

2.2.3.5 Building 3440 Large Detector Laboratory .................................. 14

2.2.4 Research Technology Laboratory ................................................................ 15

2.3 Operations with Specific Radionuclides at Hanford and PNNL ................................... 15

2.3.1 Tritium.......................................................................................................... 15

2.3.2 Cesium......................................................................................................... 16

2.3.3 Strontium ..................................................................................................... 16

2.3.4 Uranium ....................................................................................................... 17

2.3.5 Plutonium..................................................................................................... 17

2.3.6 Americium.................................................................................................... 18

2.3.7 Trace Radionuclides and Radionuclide Mixtures.......................................... 18

2.4 PNNL Radiological Protection Program...................................................................... 19

2.4.1 PNNL Air Monitoring Program...................................................................... 21

2.4.2 PNNL Area Dosimetry Program ................................................................... 22

3.0 Occupational Medical Dose.................................................................................................... 25

4.0 Occupational Environmental Dose ......................................................................................... 26

4.1 Introduction................................................................................................................. 26

4.1.1 Purpose ....................................................................................................... 26

4.1.2 Scope........................................................................................................... 26

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations ........................ 26

4.3 External Exposure from Onsite Ambient Sources ....................................................... 27

5.0 Internal Dosimetry.................................................................................................................. 29

5.1 Introduction................................................................................................................. 29

5.1.1 Purpose ....................................................................................................... 29

5.1.2 Scope........................................................................................................... 29

5.2 PNNL Internal Dosimetry Program ............................................................................. 29

5.2.1 Routine Bioassay ......................................................................................... 30
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5.2.2 Baseline and Termination Bioassay ............................................................. 31

5.2.3 Special Bioassay.......................................................................................... 32

5.3 PNNL Internal Monitoring Methods and Sensitivities .................................................. 33

5.3.1 In Vitro Analyses .......................................................................................... 33

5.3.2 In Vivo Counting........................................................................................... 35

5.3.3 Bioassay Measurement Sensitivities ............................................................ 36

5.3.3.1 In Vitro Analyses ....................................................................... 36

5.3.3.2 In Vivo Analyses........................................................................ 36

5.4 Internal Dosimetry Records ........................................................................................ 37

6.0 Occupational External Dose................................................................................................... 45

6.1 Introduction................................................................................................................. 45

6.1.1 Purpose ....................................................................................................... 45

6.1.2 Scope........................................................................................................... 45

6.2 Dose Reconstruction Parameters ............................................................................... 45

6.2.1 Administrative Practices............................................................................... 45

6.2.2 Site Dosimetry Technology .......................................................................... 46

6.2.2.1 Whole-Body Dosimeters............................................................ 46

6.2.2.2 Extremity Dosimetry .................................................................. 47

6.3 Dose Calculation and Reporting ................................................................................. 47

6.3.1 Dose Algorithm ............................................................................................ 47

6.3.2 Exposure Energy Spectra ............................................................................ 48

6.3.3 Missed Dose ................................................................................................ 49

7.0 Attributions and Annotations .................................................................................................. 49

References ........................................................................................................................................ 50

Glossary ............................................................................................................................................ 54
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LIST OF TABLES

TABLE TITLE PAGE

2-1 PNNL administrative dose limits as of April 2013 ................................................................... 20

2-2 Follow-up actions triggered by elevated DAC-hr estimates .................................................... 22

4-1 Annual doses from radioactive airborne emissions reported by Hanford and PNNL............... 27

4-2 Annual dose rates for the PNNL main campus and Hanford site locations ............................. 28

5-1 PNNL bioassay category codes and intervals ........................................................................ 31

5-2 Bioassay analysis codes and the corresponding bioassay categories.................................... 31

5-3 Bioassay analyses applicable to each bioassay category....................................................... 32

5-4 MDLs for routine processing of samples ................................................................................ 32

5-5 MDLs for priority processing of samples................................................................................. 33

5-6 MDLs for expedited processing of samples............................................................................ 33

5-7 MDLs for emergency processing of samples.......................................................................... 34

5-8 Contractual MDLs for processing of gamma spectroscopy samples....................................... 37

5-9 Nominal MDAs for the Hanford whole-body counters ............................................................. 37

5-10 Nominal MDA values for the Hanford planar HPGe detectors in various configurations ......... 37

5-11 Sentinel analysis codes and reported isotopes....................................................................... 38

5-12 Sentinel analytical laboratory codes ....................................................................................... 38

5-13 Sentinel bioassay frequency codes ........................................................................................ 39

5-14 Sentinel sample type codes ................................................................................................... 39

5-15 Sentinel bioassay measurement reason codes ...................................................................... 39

5-16 Sentinel bioassay sample kit codes and sample date conventions......................................... 40

5-17 Sentinel sample status codes................................................................................................. 41

5-18 Sentinel bioassay sample processing codes .......................................................................... 41

5-19 Sentinel no-sample codes...................................................................................................... 41

5-20 Sentinel unit codes.................................................................................................................41

5-21 Sentinel person codes............................................................................................................41

5-22 Sentinel in vivo analysis request codes.................................................................................. 42

5-23 Sentinel in vivo schedule (measurement) type codes............................................................. 42

5-24 Sentinel in vivo detector codes............................................................................................... 42

5-25 Sentinel body location codes for in vivo counts ...................................................................... 43

5-26 Sentinel in vivo no-result codes.............................................................................................. 43

5-27 INTERTRAC mode of intake codes........................................................................................ 44

5-28 INTERTRAC evaluation reason codes ................................................................................... 44

5-29 INTERTRAC source of intake codes ...................................................................................... 44

5-30 INTERTRAC miscellaneous codes......................................................................................... 44

6-1 Front and rear filtration thickness for the InLight Model 2 and Model 2T dosimeter 

elements ................................................................................................................................47

6-2 LLD values for the InLight Model 2T dosimeter ...................................................................... 49

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 PNNL site map.......................................................................................................................11
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ACRONYMS AND ABBREVIATIONS

AWE Atomic Weapons Employer

Bq becquerel

CED committed effective dose

CFR Code of Federal Regulations

cm centimeter

CRL Capabilities Replacement Laboratory

d day

DAC derived air concentration

DDE deep dose equivalent

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EMSL Environmental Molecular Sciences Laboratory

EPA U.S. Environmental Agency

eV electron-volt

F fast (absorption type)

FFTF Fast Flux Test Facility

ft foot

FUA Facility Usage Agreement

gal gallon

HEPA high-efficiency particulate air

HEU highly enriched uranium

HPGe hyperpure germanium

hr hour

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

km kilometer

L liter

LLD lower limit of detection

LSL-II Life Sciences Laboratory II

m meter

M moderate (absorption type)

MDA minimum detectable activity

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mL milliliter
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mo month

mrem millirem

n neutron

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

p proton

PNNL Pacific Northwest National Laboratory

PSF Plutonium Storage Facility

PUREX plutonium-uranium extraction

RTL Research Technology Laboratory

RU recycled uranium

RWP radiological work permit

S slow (absorption type)

SDE shallow dose equivalent

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TMA Thermo-Analytical Incorporated

U.S.C. United States Code

WESF Waste Encapsulation and Storage Facility

wk week

yr year

α alpha particle

μCi microcurie

μg microgram

§ section or sections
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1.0 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historical background information and guidance to assist in the preparation of 

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

including the grounds upon which such building, structure, or premise is located ... in which 

operations are, or have been, conducted by, or on behalf of, the [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons
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1.1 PURPOSE

This site profile for the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, 

describes aspects of the PNNL site and historical activities and practices pertinent to dose 

reconstruction under EEOICPA.

1.2 SCOPE

DOE and DOL have determined that, for dose reconstruction under EEOICPA, PNNL was not a 

separate facility from the Hanford Site until 2005. Information for performing dose reconstruction for 

claims that include PNNL employment before 2005 should therefore be obtained from the appropriate 

Project documentation for the Hanford Site. 

In addition, from 2005 forward, internal and external radiation doses received within the 300 Area of 

the Hanford site are considered Hanford exposures, regardless if they were received in PNNL 

facilities in 300 Area. Nonetheless, any occupationally derived radiation dose received at Hanford 

from 2005 forward is valid for inclusion in a dose reconstruction for an energy employee who worked 

at PNNL. 

Section 2.0 describes the PNNL site and facilities. It also discusses the radiological source terms 

present on the Hanford site as they pertain to the PNNL internal monitoring program. Section 2.0 also 

discusses the radiation protection program including air monitoring and area dosimetry. Section 3.0 

addresses routine occupational medical X-rays. Ambient levels of radionuclides in the PNNL 

environment are addressed in Section 4.0 along with estimated external ambient dose rates. 

Sections 5.0 and 6.0 address the technical issues in relation to measurement of internal and external 

dose, respectively. Attributions and annotations, indicated by bracketed callouts and used to identify 

the source, justification, or clarification of the associated information, are presented in Section 7.0. 
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2.0 SITE DESCRIPTION

2.1 INTRODUCTION

2.1.1 Purpose

This section briefly describes the physical environment of the PNNL site and the site activities and 

processes from 2005 until the present.

2.1.2 Scope

This section describes facilities on the PNNL main campus that house, or may house, operations with 

sealed radioactive sources, dispersible radioactive materials, or radiation generating devices. The 

major radiological source terms associated with the Hanford site are also addressed. Some PNNL 

workers routinely work on the Hanford Site, and these individuals are more likely to be monitored for 

occupational radiation exposure because of the relative magnitude of the radiological source terms at 

Hanford in relation to those at the PNNL main campus. Radiological source terms at Hanford, 

including PNNL facilities physically on the Hanford Site, are therefore the driver for much of the PNNL 

personnel monitoring practices. This section also describes the PNNL radiation protection program, 

including air monitoring and area dosimetry.

2.1.3 Background

PNNL is a multiprogram research institution in North Richland, Washington, just south of the Hanford 

Site. It conducts research in science, national security, energy, and environment in support of DOE 

and national strategic goals. It is operated by the Battelle Memorial Institute (Battelle) for the DOE 

Office of Science. Battelle has been the sole prime contractor for the management and operation of 

PNNL since 1965. For dose reconstruction under EEOICPA, PNNL and Hanford are considered 

indistinct through 2004. Information for claims that include PNNL employment before 2005 should 

therefore be obtained from the Hanford site profile. This PNNL site profile describes pertinent 

operations at PNNL from 2005 until the present.

In December 2003 DOE chartered the Pacific Northwest Site Office within the Office of Science to 

oversee operation of the PNNL. PNNL itself was established in 1965. PNNL is a project-based 

research institution. It does not engage in production operations and is not funded through large 

appropriations. It performs around 2,000 projects a year for DOE (including the National Nuclear 

Security Administration) and other agencies.

Figure 2-1 is a map of PNNL showing the core campus, the Hanford 300 Area to the north, and the 

2400 Stevens Building to the south. 

2.2 RADIOLOGICAL FACILITIES ON THE PNNL MAIN CAMPUS

The principal radiological facilities at PNNL are the Environmental Molecular Sciences Laboratory 

(EMSL), the Physical Sciences Facility (PSF), the Research Technology Laboratory (RTL), and the 

Life Sciences Laboratory II (LSL-II). The PSF and the RTL consist of multiple buildings. PNNL also 

conducts radiological work in Building 325 (the Radiochemical Processing Laboratory) and other 

facilities in the Hanford 300 Area. Information about Hanford facilities should be obtained from the 

Hanford site profile. 

There have been no known, large-scale radiological incidents at PNNL from 2005 to the present. 

Information pertaining to radiological incidents that only impacted a limited number of individuals 

should be available in a worker’s claim records. 
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Figure 2-1. PNNL site map (part 1). 
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Figure 2-1. PNNL site map (part 2). 

2.2.1 Environmental Molecular Sciences Laboratory

The EMSL (Building 3020) was constructed in 1997. It is a single building encompassing 224,000 

gross square feet. It functions as a national scientific user facility providing experimental and 

computational resources to researchers from the academic community, industry, and other 

government laboratories (Snyder, Barnett, and Bisping 2014; BMI 2012a). 

The EMSL consists primarily of offices, an auditorium, laboratories, and maintenance and storage 

areas. The laboratory spaces in EMSL consist of wet laboratories and computer laboratories. The 

wet laboratory areas were designed for work with hazardous chemical, radioactive, or biological 

materials. Experiments in the EMSL can involve complex systems and equipment and research 

animals (Snyder, Barnett, and Bisping 2014). 

Research in the EMSL focuses on integrating computational and experimental capabilities. Its 

radiological operations are limited to sealed source use and authorized work with volumetrically 

released and nondispersible materials (Snyder, Barnett, and Bisping 2014). 

2.2.2 Life Sciences Laboratory II

The LSL-II Building consists primarily of two laboratory floors with attached mechanical and electrical 

service rooms at the north and south ends. Other facilities that support LSL-II include the Chemical 

Storage and Transfer Facility (also known as LSL-A) and the Portable Chemical Storage Facility (also 

known as LSL-B). 

Research in this facility includes applied research, prototype development and testing, and system 

validation for engineered structural materials. Mechanical design, automation, computational 

mechanics, and advanced materials characterization activities are also conducted in LSL-II. Some 

electronic technology development and wet chemical work are performed (Snyder, Barnett, and 

Bisping 2014). 

The Facility Usage Agreement (FUA; BMI 2012b) for the LSL-II facility requires that the total quantity 

of radioactive material in the building be maintained below the threshold for DOE Hazard Category 3 

(as defined in DOE 1997). Some workspaces within LSL-II have ventilation systems with highefficiency particulate air (HEPA) filters, and radioactive air emissions monitoring is performed using a 

radioactive materials inventory tracking system (BMI 2012b). 

Type I and Type II sealed radioactive sources are allowed in LSL-II, and the FUA implies that 

uncertified sealed radioactive sources are allowed (BMI 2012b). 
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Unsealed radioactive material is allowed in LSL-II but is limited to the existing inventory plus holdup. 

Special nuclear and fissionable materials are not allowed (BMI 2012b). 

2.2.3 Physical Sciences Facility

The PSF was an element of the Capabilities Replacement Laboratory (CRL) initiative at PNNL to 

provide facilities and capabilities to replace those it would be losing due to the substantial clean-up 

activities being completed in the Hanford 300 Area. Altogether, the CRL project consisted of 

construction of the PSF, the Biological Sciences Facility, the Computational Sciences Facility, the Life 

Sciences Facility, and modifications to Hanford 300 Area Buildings 318, 325, 331, and 350. 

The PSF is not a single facility, but is rather a complex of the following buildings and facilities 

constructed in 2010:

• 3410 Material Science and Technology Laboratory, 

• 3420 Radiation Detection Laboratory, 

• 3425 Ultra-Low Background Laboratory, 

• 3430 Ultra-Trace Detection Laboratory, 

• 3440 Large Detector Laboratory, 

• 3455 Trailer A, 

• 3465 Trailer B, and

• Laboratory Support Warehouse (LSW). 

Trailers A and B (Buildings 3455 and 3465) and the LSW do not house radiological operations (BMI 

2011a, 2011b). 

2.2.3.1 Building 3410 Material Sciences and Technology Laboratory

Building 3410 is a two-story building constructed in 2010. It consists of office space, conference 

rooms, and common areas. The rear portion of the first floor is laboratory space. The rear portion of 

the second floor is mechanical space. The Central Utility Plant is attached to the north end via a 

common corridor. This facility includes craft shops and a receiving dock in addition to process 

systems (chilled water, compressed air, etc.) (BMI 2000). 

Type I and Type II sealed radioactive sources are allowed in Building 3410, as are unsealed 

radioactive materials (within air permit limitations). Fissionable materials are allowed, but special 

nuclear materials are not. The FUA for Building 3410 requires that the total radioactive material 

inventory not exceed the threshold for DOE Hazard Category 3 (as defined in DOE 1997). However, 

the PSF has administrative limits that are less than the Category 3 threshold. Building 3410 has a 

continuously operating radioactive air confirmatory sampling system (BMI 2012c). 

Building 3410 houses research capabilities associated with performance and life of materials in hightemperature, high-radiation, and corrosive environments such as those in next-generation 

technologies and applications in the areas of energy, construction, and transportation. Activities 

include work with metals, ceramics, polymeric materials, composites, and specialized coatings and 

surface treatments (BMI 2014). 

2.2.3.2 Building 3420 Radiation Detection Laboratory

Building 3420 contains office and laboratory space that supports research in ultra-low background 

radiation detection. The research involves a wide variety of radionuclide measurement technologies 

and capabilities including state-of-the-art analytical chemistry, radiation physics, light detection, 
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particle detection, chromatography, scintillation materials, sorbents and “smart” materials, and field 

deployable instrumentation (BMI 2014). 

Type I, Type II, and uncertified sealed radioactive sources are permitted in Building 3420, as are 

unsealed radioactive materials (within air permit limitations). Fissionable and special nuclear 

materials are allowed, with the latter being subject to safeguards category IV limits (BMI 2011c). 

Two radioactive air permits are applicable to Building 3420. One covers laboratory spaces that are 

connected to a HEPA-filtered exhaust system, and the other covers laboratory and support spaces 

that are tied to the unfiltered exhaust system (BMI 2011c). 

2.2.3.3 Building 3425 Ultra-Low Background Counting Laboratory

Building 3425 is primarily a below-grade structure that houses research in trace chemical and 

radionuclide detection. Also known as the Underground Laboratory or the Deep Lab, it is as much as 

40 ft underground to minimize background radiation. The facility supports the development and 

advancement of radiation detection technologies in conjunction with research in Building 3420. 

Additional activities in Building 3425 include radiation physics experiments, development of ultra-low 

radioactivity materials, and other fundamental sciences studies (BMI 2014). 

Type I, Type II, and uncertified sealed radioactive sources are permitted in Building 3425, as are 

unsealed radioactive materials (within air permit limitations). Fissionable materials are allowed, but

special nuclear materials are not. The FUA for Building 3425 requires that the total radioactive 

material inventory not exceed the threshold for DOE Hazard Category 3 (as defined in DOE 1997). 

However, the PSF has administrative limits that are less than the Category 3 threshold. Building 3425 

does not have an airborne effluent monitoring system (BMI 2010a). 

2.2.3.4 Building 3430 Ultra-Trace Analysis Building

Building 3430 houses research in ultra-trace and low-level detection and characterization of 

radionuclides for nuclear forensics applications. Its capabilities include highly sensitive analytical 

systems such as mass spectrometers, optical microscopes, and electron microscopes to provide 

isotopic analyses and ultra-low-level radionuclide detection in a wide variety of sample matrices. 

Type I and Type II sealed radioactive sources are allowed in Building 3410, as are unsealed 

radioactive materials. Fissionable radioactive materials are allowed, but amounts cannot exceed 

exempt quantities as defined in the PNNL Criticality Safety Subject Area. Special nuclear materials 

are allowed as long as the total does not exceed safeguards Category IV quantities. 

Authorized radiation generating devices are allowed in Building 3430. The FUA for Building 3430 

calls for a continuously running confirmatory radioactive airborne effluent monitoring system 

(BMI 2010b). 

2.2.3.5 Building 3440 Large Detector Laboratory

Building 3440 houses work on the development and testing of large radiation detection systems along 

with testing of high-energy radiography systems (BMI 2013a). 

Type I, Type II, and uncertified sealed radioactive sources are permitted in Building 3440. Fissionable 

and special nuclear materials are allowed, with the latter being subject to safeguards Category IV 

limits. Unsealed radioactive materials are not allowed in Building 3440. It has no sampling or 

monitoring system for radioactive airborne effluents (BMI 2013a). 
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2.2.4 Research Technology Laboratory

The RTL is a complex of buildings of which the main facility is Building RTL-520. RTL-520 provides 

laboratory, office, and storage space in support of a variety of research and development activities. 

Research includes chemical toxicology, environmental health physics, dosimetry, atmospheric 

science modeling, and soil and groundwater contamination studies. Coating and coating 

technologies, laser and electrochemical machining, and electrodeposition research is performed as 

well as research into the solid-liquid interface of geologic materials (BMI 2014). 

Building RTL-530 is a small (136 ft2

) concrete block and brick storage area just west of RTL-520 for 

temporary storage of radioactive materials. RTL-530 includes a lead-shielded pit underneath its 

concrete floor for storing highly radioactive materials. 

Other buildings associated with the RTL include: 

• RTL 510 Chemical and Flammable Storage, 

• RTL 550 Technical Services, 

• RTL 560 Utility Building, 

• RTL 570 Autoclave Center, 

• RTL 580 Crafts Shop, and

• RTL 590 Warehouse. 

The FUA for the RTL requires that the total quantity of radioactive material be maintained below the 

threshold for DOE Hazard Category 3 (as defined in DOE 1997). Some workspaces in RTL-520 have 

HEPA-filtered ventilation systems, and confirmatory air sampling for radioactive materials is 

performed at eight emissions points (BMI 2012d). 

Type I and Type II sealed radioactive sources are allowed in RTL-520 but not in the other RTL 

facilities without specific authorization (BMI 2012d). Unsealed radioactive materials, special nuclear, 

fissionable, and high-toxicity radioactive materials are allowed in RTL-520 and RTL-530 (BMI 2012d). 

Unsealed radioactive, high-toxicity radioactive, special nuclear, and fissionable materials are not 

allowed in other RTL facilities (BMI 2012d). 

2.3 OPERATIONS WITH SPECIFIC RADIONUCLIDES AT HANFORD AND PNNL

PNNL workers would most likely encounter dispersible radioactive materials while performing duties 

on the Hanford Site. The methods of the PNNL internal monitoring program therefore emphasize 

source terms associated with facilities and operations in the Hanford 300 Area and other Hanford 

locations. The following sections discuss the principal radionuclides and source terms present at 

Hanford and PNNL. 

2.3.1 Tritium

Tritium work at PNNL has included radioluminescent lights developed by the laboratory and used as a 

tracer or labeling compound for biological research projects. Predominant forms of tritium have been 

tritium oxide (tritiated water), tritium gas, and stable metal tritides in at least one project (notably 

zirconium tritide).

Organically bound tritium is not atypical of PNNL facilities; it is usually found in ingested foods, 

machinery oil from tritium operations, laboratory compounds labeled for research purposes, or wastes 

from other DOE facilities. 
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Tritiated particulate aerosols result from tritium being absorbed and retained on metal surfaces such 

as getters for tritium collection and storage devices (typically these are in the form of metal tritides) 

and as a residual contaminant in tritium production facilities. An additional possibility for tritiated 

particulates is tritiated metal oxides (e.g., rust and dust). A limited source of zirconium tritide metal 

filings was identified in 325 Building at Hanford.

The basic tool for monitoring intakes of soluble and insoluble tritium particulates is personal air 

sampling. Derived air concentrations (DACs) for soluble and insoluble tritiated particulates have been 

promulgated in 10 CFR Part 835 to aid in monitoring of workers and the workplace.

2.3.2 Cesium

Along with 90Sr, 137Cs is a predominant fission product isotope in source terms across Hanford. 

Historically, there have been 137Cs production operations as well, but because of the lack of new 

production and radioactive decay it is no longer a significant radionuclide in Hanford source terms.

In addition to its presence in mixtures, 137Cs has existed in relatively pure form at the Waste 

Fractionation Facility (B Plant, 221-B Building) and the Waste Encapsulation and Storage Facility 

(WESF, 225-B Building) in the Hanford 200 Area. Encapsulation programs at WESF have been 

terminated, but cesium-bearing capsules and cesium-contaminated equipment are still stored in the 

facility.

PNNL has found cesium to be more dispersible than strontium and that 137Cs will likely constitute the 

major component of most intakes involving 137Cs and 90Sr. PNNL therefore recommends that 

additional bioassay appropriate for other radionuclides of concern be considered in situations where 

no radionuclide ratio information exists but it is suspected that 90Sr or other radionuclides might be 

present along with 137Cs.

At PNNL, bioassay for cesium is performed using whole-body counting. A 3-minute count on the NaI 

preview counter provides a minimum detectable activity (MDA) of 1.3 nCi for 137Cs. A 10-minute 

count on the coaxial germanium scanning detector provides an MDA for 137Cs of about 1 nCi.

Because of its relative abundance in radiological source terms at Hanford, PNNL often uses 137Cs as 

an indicator for other nuclides, notably 90Sr and plutonium. PNNL recommends annual in vivo 

measurements for workers who are potentially exposed to radionuclide mixtures that contain 137Cs.

2.3.3 Strontium

The isotopes of dominant concern for strontium internal dosimetry are 90Sr and its decay product 90Y. 

These nuclides can be found in almost any Hanford or PNNL facility that deals with fission products or 

fission product waste mixtures. Most facilities that have strontium can also be expected to have other 

fission products present, notably 137Cs, and it is a common practice at PNNL to use 137Cs as an 

indicator of potential 90Sr. However, some chemical processes at Hanford have separated cesium 

from strontium, and relatively pure 90Sr could be associated with laboratories, waste separation 

facilities, and waste storage tank sludges. In particular, cesium and strontium separation operations 

were historically performed in the B Plant (221-B) and the WESF (225-B) in the Hanford 200 Area.

The presence of 89Sr in Hanford source terms is no longer a concern, unless a new source, such as 

material received from off site, is introduced.

All inhalation and ingestion intakes of radiostrontium at PNNL are considered to be absorption type F. 

This includes fluoride forms of strontium potentially present at B Plant and WESF from the waste 
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fractionation and encapsulation operations that were performed there into the mid-1980s. Strontium 

titanate has not been used at PNNL. 

A detailed study of the environmental background levels of 90Sr in the urine of unexposed Hanford 

workers was performed in 2008 (Antonio and Rivard 2008). The study found that an environmental 

screening level of 1 dpm/d would imply that about 1 out of 100 workers not occupationally exposed 

might exceed it. For routine 90Sr urinalyses, the PNNL internal dosimetry program uses a 10-mrem 

investigation screening level of 4 dpm/d consistent with guidance in the American National Standards 

Institute standard ANSI/HPS N13.39 (HPS 2001). 

2.3.4 Uranium

Uranium at Hanford or PNNL can be present from a number of possible sources, including residual 

slightly enriched recycled uranium (RU) from the fuel cycle of the production reactors, depleted 

uranium (DU) from research and development projects, and various individual isotopes or mixtures 

associated with short-term research-scale projects. Highly enriched uranium (HEU) and natural 

uranium (NU) were used at Hanford in its early history, so residual contamination or buried waste 

containing HEU or NU could be encountered during cleanup operations.

Uranium compounds involved in historical production operations at Hanford ranged from highly 

soluble uranyl nitrate, to somewhat less soluble uranium trioxide, to relatively insoluble uranium 

oxides. Uranium hexafluoride and uranyl fluoride were not handled at Hanford.

Residual uranium contamination at Hanford generally represents the composition of the bulk of the 

processing over the years, which ranged from depleted to slightly enriched, based on the weight 

percent of 235U. However, other enrichments might be encountered during cleanup of old facilities. 

For instance, a nondestructive analysis of gloveboxes in the 231-Z Building in 2008 found some 

gloveboxes with enrichments ranging from 1% to 18% (Hilliard 2008). These gloveboxes were 

originally used in the 308 Building. Some irradiated fuel from the Fast Flux Test Facility (FFTF) is

stored in the 200 Area, but it is well contained and should not be a source of contamination. FFTF 

fuel had various mixtures of NU and plutonium so, if a contamination event were to occur, it would be 

necessary to get ratios of uranium and plutonium isotopes from the contamination itself. Uranium-233

was separated and handled during the thorium extraction campaigns at the Plutonium-Uranium 

Extraction (PUREX) Plant in 1965, 1966, and 1970. HEU fuel slugs were used in some Hanford 

production reactors from 1949 to 1958. These compositions might also be encountered during 

clean-up operations. See the Hanford site profile for information on isotopic ratios and content of 

other constituents present in uranium source terms at Hanford. 

PNNL considers single, acute intakes of uranium to be the most likely scenario, although chronic 

exposure to low levels of uranium contamination in soil is possible during excavation operations. The 

exposure period for such operations is typically days or weeks, and intakes are monitored via air 

sampling.

2.3.5 Plutonium

Plutonium can be found in nearly all Hanford facilities, including waste handling facilities and tank 

farms. Plutonium can also be found at Hanford and PNNL as a result of PNNL research projects that 

involved spent fuel from nuclear power plants, breeder reactor applications, and radioisotope 

applications. Many of these research projects were in 300 Area facilities. Analytical chemistry 

laboratories might contain plutonium standard solutions.
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Pure isotopes of plutonium are seldom encountered at Hanford or PNNL facilities. Plutonium mixtures 

at Hanford and PNNL generally fall into one of two categories:

• Weapons-grade plutonium having a nominal 240Pu content of 6% by mass, and

• Fuel-grade plutonium having a nominal 240Pu content of 12% by mass.

A third category, commercial reactor grade, also exists. This material originates from nuclear power 

reactor fuel cycles and contains higher amounts of 240Pu (e.g., 25% by weight) than the weapons- or 

fuel-grade mixtures. Commercial reactor grade plutonium would be associated with research projects 

involving commercial reactor fuels, such as the Nuclear Waste Vitrification Project in the 1970s. 

The Hanford site profile contains isotopic mixture information for weapons-grade, fuel-grade, and 

commercial reactor grade plutonium as a function of age. The Hanford site profile should also be 

reviewed for information on the appropriate absorption type to assign for a given case.

2.3.6 Americium

Americium is found at Hanford as the ingrown 241Am progeny of 241Pu in a plutonium mixture or as a 

separated 241Am isotope that exists singly or in combination with other separated isotopes in waste 

mixtures or laboratory facilities. In addition, 243Am has been used in some PNNL facilities as a 

research isotope. Both ingrown and separated 241Am were historically present at the Hanford 

Plutonium Finishing Plant (Building 234-5Z) and associated facilities.

Separated 241Am is also a trace contaminant in many of the 200 Area tank farm waste mixtures. It is 

not likely that tank farm waste would contain separated 241Am as a pure isotope; it can be anticipated 

that it would be accompanied by much larger quantities of fission product activity. However, there 

have been isolated discoveries of relatively pure 241Am in tank farm related facilities (Maclellan 2013). 

Pure isotopes of 241Am and 243Am can be found in analytical laboratory standard solutions and in pure 

isotope research applications.

2.3.7 Trace Radionuclides and Radionuclide Mixtures

Radionuclide mixtures at Hanford can be found in reactor facilities, former chemical separations and 

processing facilities, waste management facilities, and laboratories. The radionuclides most 

commonly encountered in mixtures are 137Cs, 90Sr, and plutonium. PNNL bioassay capabilities 

therefore consider assessment of intakes for mixtures of 137Cs and 90Sr, 137Cs and plutonium, and 90Sr 

and plutonium. 

PNNL internal dosimetry methods for radionuclide mixtures make use of established radionuclide 

ratios in the case of well-characterized sources, but in other cases the detection of a nuclide that 

might be a component of a mixture is used to trigger supplemental bioassay for other, potentially 

significant, radionuclides. 

Mixtures of 137Cs and 90Sr

Mixtures of 137Cs and 90Sr are particularly common in the Hanford 200 Area tank farms. Given the 

wide variety of waste treatment methods that have been applied at Hanford over the years, a 1-to-1 

ratio of 137Cs to 90Sr (reflecting their intrinsic fission yields from uranium fuels) should not be assumed. 

PNNL asserts that an annual whole-body count using the preview counter meets their goal of a 

minimum detectable dose of 100 mrem committed effective dose (CED) for 137Cs:90Sr activity ratios of 

up to approximately 1:30. If the coaxial germanium counter is used for whole-body counts the 

137Cs:90Sr activity ratio corresponding to 100 mrem CED improves to 1:40. PNNL states that workers 
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with the potential for exposure to mixtures of 137Cs and 90Sr where the ratio of 90Sr to 137Cs exceeds 

30 (nominally) should be considered for supplemental bioassay for strontium. 

Mixtures of 137Cs and Plutonium

Mixtures of 137Cs and plutonium can be found in Hanford facilities associated with spent fuel 

management and in waste streams from such facilities. Examples include the spent fuel basins, fuel 

processing hot cells, and waste tank sludges. Mixtures of 137Cs and plutonium are likely to be 

predominantly 137Cs on an activity basis, but a wide range of 137Cs:plutonium ratios is possible, from 

1,000:1 to 1:1. PNNL states that determining the absorption of the plutonium for such mixtures is 

problematic and acknowledges that the minimum detectable dose associated with lower 

137

Cs:plutonium activity ratios (i.e., mixtures with higher plutonium content) and annual in vivo 

counting for 137Cs can be large, especially for type S plutonium. PNNL therefore recommends 

supplemental plutonium bioassay for workers with the potential for exposure to mixtures of 137Cs and 

plutonium where the 137Cs:plutonium activity ratio is approximately 30 or less. 

Mixtures of 90Sr and Plutonium 

Mixtures of 90Sr and plutonium are considered most likely to occur in tank farm facilities at Hanford, 

probably associated with contamination from waste tank sludges. For its assessments of minimum 

detectable dose, PNNL assumes the 90Sr is type F and the plutonium is type M, consistent with the 

nitrate nature of most waste tank contents. PNNL acknowledges the possibility that dry plutonium 

contamination that is exposed to dry air and a normal building or outdoor temperature environment 

can undergo gradual oxidation over time and approach a type S material, but asserts that the use of 

type M dose coefficients for plutonium provides a conservative approach to the estimations of 

minimum detectable dose based on using 90Sr as a tracer radionuclide.

PNNL recommends that workers be considered for supplemental bioassay for plutonium (in addition 

to strontium) if there is potential for exposure to mixtures where the 90Sr:plutonium activity ratio is 

about 100 or less. 

2.4 PNNL RADIOLOGICAL PROTECTION PROGRAM

PNNL implements the occupational radiation protection requirements of 10 CFR Part 835 and 

applicable DOE directives through a radiation protection program designed after the DOE 

Radiological Control Manual (DOE 1992). The current program is detailed in the PNNL Radiation 

Control Program Description published in April 2013 (BMI 2013b). Radiological support is provided to

PNNL line managers and workers by the Radiation Protection Division. PNNL radiation protection 

personnel are expected to receive day-to-day priorities from facility managers. 

Consistent with 10 CFR Part 835, general employees who have not completed appropriate training, 

examinations, and performance demonstrations are not permitted unescorted access to any 

radiological area at PNNL. Radiological workers are the only individuals expected to receive greater 

than 100 mrem in a year. The majority of monitored workers at PNNL do not receive measurable 

doses. 

In general, all radiological work at PNNL is project-based. Radiological work permits (RWPs)

therefore serve as the principal means for informing workers of area radiological conditions and entry 

requirements and to provide a mechanism for relating worker exposure to specific work activities. 

PNNL mandates the use of RWPs to control the following activities:

• Entry into radiological areas, 

• Handling of materials with removable contamination that exceed established limits, 
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• Work in localized benchtop areas, laboratory fume hoods, sample sinks, and containment 

devices that have the potential to generate contamination in areas that are otherwise free of 

contamination, 

• Work that disturbs the soil in soil contamination areas, and

• Work that involves digging in underground radioactive material areas.

The PNNL radiation control program description states RWPs should include the following information 

unless the information is already included in related work documents (BMI 2013b): 

• Dosimetry requirements including any bioassay requirements, 

• Prejob briefing requirements as applicable, 

• Training requirements for entry, 

• Protective clothing and respiratory protection requirements, 

• Radiation protection coverage requirements and stay time controls as applicable, 

• Limiting radiological conditions that might void the RWP, 

• Special dose or contamination reduction considerations, and

• Special personnel frisking considerations.

PNNL uses three different types of RWPs. Laboratory-wide RWPs are used for entry and repetitive 

work in areas with low-hazard radiological conditions. Owner-specific RWPs are used for entry and 

repetitive work in areas with low-moderate hazard radiological conditions. Job-specific RWPs are 

used for more complex work and for entry into high-hazard areas. 

Laboratory-wide and owner-specific RWPs can be used to control routine or repetitive activities, such 

as tours and inspections or minor work activities outside of areas with engineered controls (e.g., fume 

hood or glovebox), in areas with well-characterized and stable radiological conditions, or for work 

inside areas with engineered controls that contain moderate hazard radiological conditions. Jobspecific RWPs are recommended to control nonroutine operations or work in areas with changing 

radiological conditions. 

PNNL has an ALARA (as low as is reasonably achievable) committee, one of the functions of which is 

to determine the applicable administrative dose limits for PNNL and reassess them annually based on 

an evaluation of historical and projected radiation exposures, workload, and mission. A tiered system 

of limits is established under which approval from higher levels of management is needed to exceed a 

given limit. The four tiers of PNNL administrative dose limits as of April 2013 are given in Table 2-1. 

Table 2-1. PNNL administrative dose limits as of April 2013 (mrem). 

Tier Total effective dose Skin/extremity Lens of eye Any organ

1 500 15,000 4,500 15,000

2 1,000 22,500 6,750 22,500

3 1,500 30,000 9,000 30,000

4 2,000 Not applicable Not applicable Not applicable

An administrative limit on lifetime total effective dose is also in place. It is equal to an individual’s age 

times 1,000 mrem. 
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2.4.1 PNNL Air Monitoring Program

Consistent with 10 CFR Part 835, the PNNL radiation protection program mandates that monitoring 

for airborne radioactivity be performed under the following circumstances (Stoetzel 2012):

• An individual is likely to receive an exposure of 40 or more DAC-hr in a year, or

• As necessary to characterize the airborne radioactivity hazard where use of respiratory 

protective devices has been prescribed.

In addition, the PNNL radiation protection program requires that real-time (or continuous) air 

monitoring be performed as necessary to detect and provide warning of airborne radioactivity 

concentrations that warrant immediate action to terminate inhalation of radioactive material 

(Stoetzel 2012).

Workplace air sampling and monitoring at PNNL consist of fixed-location air sampling, job-specific 

sampling, and real-time air monitoring.

Fixed-location air samplers consist of a sample head and a particulate filter connected to a vacuum 

source, which is often a building vacuum system. If needed, portable, giraffe-type air samplers are 

used for fixed-location monitoring in remote areas not served by a building vacuum system. Fixedlocation air sampling is required in occupied areas where a worker is likely to receive an annual 

exposure of 40 DAC-hr or more under normal work conditions, or for entry into posted airborne 

radioactivity areas.

Job-specific air sampling involves the collection of prejob, during-job, and postjob air samples, as 

appropriate. Prejob samples can be used to establish radon progeny concentrations or verify prejob 

conditions or postings. During-job sampling is used as needed to estimate worker DAC-hr exposures 

and to verify respiratory protection requirements. Postjob sampling can be used to verify appropriate 

conditions for down-posting (i.e. reducing the posted radiological hazards and entry requirements). 

Job-specific air sampling is required during work in posted airborne radioactivity areas, during breach 

of contaminated systems or components (e.g., gloveboxes or hot cells), and during whole-body 

entries into high-contamination areas or work on equipment with high levels of removable 

contamination. Job-specific air sampling is performed using low-volume, high-volume, or lapel air 

samplers, as appropriate. Kinetic impactors might also be used for particle size studies.

Real-time air monitoring is required where there is potential for concentrations of airborne radioactivity 

that would warrant immediate action to terminate inhalation of radioactive material. Typical locations 

include near gloveboxes that contain highly radiotoxic materials; in storage rooms used to store 

potentially dispersible, highly radiotoxic materials; near doorways to hot cell airlocks; and near hot cell 

transfer mechanisms or ports. Real-time air monitoring at PNNL is performed using alpha, beta, or 

tritium continuous air monitors, as appropriate.

DAC-hr tracking is performed for all jobs requiring during-job air sampling. The need for DAC-hr 

tracking is also evaluated for fixed-location air sample results that exceed 2% of a DAC. Table 2-2

summarizes follow-up actions triggered by DAC-hr estimates for single intakes and for cumulative 

exposure over a calendar year. PNNL does not track DAC-hr estimates that are less than 1 DAC-hr.

When performing dose evaluations triggered by DAC-hr estimates, the PNNL internal dosimetry staff 

makes a determination about the representativeness of the job-specific air sampling results. All lapel 

monitoring results are considered representative. Likewise, samples collected within 1 ft of the 

individual’s nose and mouth are considered representative unless they are invalidated by cross 

contamination or other factors. Samples not in the immediate vicinity (within 1 ft ) of the worker’s 

nose and mouth, but collected from a location between the worker(s) and the release point, are 
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considered representative on the basis the result should be conservative. If necessary, airflow testing 

can be performed in cases when air samples do not meet these representativeness criteria to 

determine if a sample can be considered representative or if correction factors are needed.

These same representativeness criteria are applied regardless if dose is being assigned from jobspecific air sampling results or from fixed-location (continuous) air monitoring. In some instances 

elevated continuous air monitoring results can be caused by a job in that local area that required jobspecific air sampling. If so, then the DAC-hr estimates from the job-specific monitoring are used in 

lieu of those from the continuous air monitors.

2.4.2 PNNL Area Dosimetry Program

In January 1993 PNNL established an area monitoring dosimeter program in accordance with Article 

514 of the DOE Radiological Control Manual (DOE 1994). The purpose of the program was to 

Table 2-2. Follow-up actions triggered by elevated DAC-hr estimates. 

Type of exposure

DAC-hr 

estimate Follow-up actions

Acute (single intake) 1 to <10 Assign dose directly from DAC-hr estimate or conduct special 

bioassay. The DAC-hr estimate is entered into the DAC-hr tracking 

system if bioassay is not performed.

Acute (single intake) 10 to ≤40 Further analyze the air sample to determine the isotopic mix and better 

define the potential exposure. Conduct special bioassay and dose 

evaluation as deemed appropriate by PNNL internal dosimetry staff. 

The DAC-hr estimate is entered into the DAC-hr tracking database if 

bioassay is not performed.

Acute (single intake) >40 Initiate an internal dosimetry investigation. Further analyze the air 

sample to determine the isotopic mix and better define the potential 

exposure. Conduct special bioassay and perform an internal dose 

evaluation.

Cumulative in one 

calendar year

≤10 Enter the DAC-hr estimate into the DAC-hr tracking system and verify 

that the worker is on appropriate bioassay. The DAC-hr estimate is 

deleted from the tracking system once bioassay results are received or 

a permanent dose is recorded based on the DAC-hr estimate.

Cumulative in one 

calendar year

>10 Enter the DAC-hr estimate into the DAC-hr tracking system and verify 

that the worker is on appropriate bioassay. PNNL internal dosimetry 

determines if internal dose evaluation is needed. The DAC-hr estimate 

is deleted from the tracking system once bioassay results are received 

or a permanent dose is recorded based on an internal dose evaluation.

demonstrate that external radiation doses to individuals in certain areas did not exceed established 

thresholds and, therefore, that personnel dosimetry in those areas was not required. 

The PNNL radiological control program requires personnel dosimetry for: 

• Radiological workers who are likely to receive at least 100 mrem annually, and

• Declared pregnant workers, minors, and members of the public who are likely to receive at 

least 50 mrem annually.

The area dosimeters were exchanged and analyzed quarterly or annually depending on the location. 

For 2007 through 2011 the dosimeters were the Hanford standard dosimeter (Harshaw Model 8825) 

and the Hanford combination neutron dosimeter. The area dosimetry program was expected to 

transition to the Landauer InLight Model 2 and Model 2T optically stimulated dosimeters in 2012 
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(Bivins and Stoetzel 2012) but it is unknown if this actually occurred. PNNL transitioned to the same 

Landauer dosimeters for personnel monitoring as of January 2013 (see Section 6.0). 

The standard dosimeter was used in all monitored locations. Some locations were also monitored 

with neutron dosimeters in addition to the standard dosimeters. Neutron dosimeters were mounted on 

5-gal carboy bottles to allow for albedo response. 

Investigation levels for area dosimeter results were 40 mrem for quarterly reads and 160 mrem for 

annual reads for 2007 through 2009. In 2010 the investigation levels were increased to 50 mrem for 

quarterly exchange and 200 mrem for annual exchange. The investigation levels reflect an assumed 

occupancy rate of 2,000 hours per year, which equates to an occupancy fraction of 0.23 (given a year 

defined as 8,760 hours). Applying the assumed occupancy yields threshold values of 54 mrem per 

quarter and 217 mrem per year to reach 50 mrem in a 2,000 hour working year. 

In addition to the above occupancy fraction (0.23), PNNL also applies an occupancy factor when 

evaluating the area dosimeter readings relative to its 50-mrem threshold. The occupancy factor is 

defined so that a value of 1.0 equals 2,000 hours of occupancy in a year. PNNL also applies a 

background correction to the dosimeter readings, so the assigned annual dose results represent net 

values. A neutron correction factor is applied to neutron dose readings to account for the difference 

between the unmoderated neutron spectrum used for dosimeter calibration and the moderated or 

scattered neutron spectrum present in the monitoring locations. 

The selection criteria for locations for area dosimetry monitoring were as follows:

• All PNNL facilities in the Hanford 300 Area where staff worked at least 8 hours per month, 

• All PNNL facilities where staff performed radiological work (i.e. had a current radiological work 

permit), and

• All PNNL facilities within 15 m of another facility (including Hanford contractor facilities) that 

contained an indoor or outdoor radiological area.

Locations were selected for neutron monitoring as needed. Note that “PNNL” in this context is from 

the Laboratory’s perspective. PNNL facilities within the Hanford 300 Area are considered Hanford 

facilities under EEOICPA. 

None of the area dosimetry results for 2007 through 2011 exceeded 50 mrem in a year after 

correcting for occupancy. Information about the number of monitored locations and those selected for 

neutron monitoring is summarized below. This information was sourced from the annual reports from 

the PNNL area dosimetry program for 2007 through 2011 as noted below:

• 2007. 132 locations were monitored using the standard dosimeter. Additional neutron 

monitoring was performed in three locations in the Hanford 300 Area: one in Building 318 and 

two in Building 325 (Bivins and Stoetzel 200e8). 

• 2008. 135 locations were monitored using the standard dosimeter. Additional neutron 

monitoring was performed in three locations in the Hanford 300 Area: one in Building 318 and 

two in Building 325 (same locations as the year prior) (Bivins and Stoetzel 2009). 

• 2009. 135 locations were monitored using the standard dosimeter. Additional neutron 

monitoring was performed in three locations in the Hanford 300 Area: one in Building 318 and 

two in Building 325 (same locations as the year prior) (Bivins and Stoetzel 2010). 
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• 2010. 176 locations were monitored using the standard dosimeter. Additional neutron 

monitoring was performed in seven locations in the Hanford 300 Area and one location in the 

EMSL on the PNNL main campus. Neutron monitoring was performed in Building 318 (one 

location), Building 325 (two locations), Building 331 (four locations), and the EMSL (one 

location in Room 1344). The monitoring in the EMSL was prompted by a 252Cf sealed source 

stored in a safe in Room 1344 (Bivins and Stoetzel 2011). 

• 2011. 154 locations were monitored using the standard dosimeter. Additional neutron 

monitoring was performed in seven locations in the Hanford 300 Area, one location in the 

EMSL, and one location in Building 3440. The Hanford 300 Area and EMSL neutron 

monitoring locations were the same as those monitored in 2010. For 2011 an additional 

neutron area dosimeter was placed in Building 3440 due to neutron sources in a safe in Room 

1303 (Bivins and Stoetzel 2012). 
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3.0 OCCUPATIONAL MEDICAL DOSE

Occupational medical services for PNNL employees, including all medical X-rays required as a 

condition of employment, were provided by the Hanford site contractor from January 1, 2005, through 

September 30, 2012. Dose reconstruction information for energy employees that had PNNL 

employment during that period should therefore be obtained from ORAUT-TKBS-0006-3, Hanford Site 

– Occupational Medical Dose (ORAUT 2010).

Since October 1, 2012, occupational medical X-rays for PNNL employees have been performed at the 

Kadlec Medical Center in Richland, Washington. Kadlec Medical Center is not a covered facility 

under EEOICPA for this time period; therefore, radiation dose from occupational screening X-rays 

after September 30, 2012, should not be included in dose reconstructions.
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4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

4.1 INTRODUCTION

4.1.1 Purpose

This section addresses the occupational environmental dose applicable to the PNNL site from 2005 to 

the present. Occupational environmental dose refers to the radiation dose a worker received outside 

of buildings, but on the PNNL site, as a result of ambient airborne radionuclides or ionizing radiation.

4.1.2 Scope

Internal and external exposures to radionuclides in the outdoor environment are considered 

separately in Sections 4.2 and 4.3, respectively.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS

Both the Hanford and PNNL sites release airborne radioactive materials to the air as a result of 

operations. Both sites, therefore, engage in routine airborne effluent monitoring and issue annual 

reports detailing airborne releases and the associated radiation dose to workers and members of the 

public. These reports are provided, in part, to demonstrate compliance with U.S. Environmental 

Agency (EPA) and State of Washington requirements that the dose to a member of the public from 

radioactive air emissions not exceed 10 mrem/yr. Compliance is shown by calculating the dose to the 

nearest receptor or to an otherwise hypothetical maximally exposed individual. The dose calculations 

are performed using the site’s airborne effluent monitoring and local meteorological data as inputs to 

EPA-approved computer programs for that purpose.

From 2005 through 2009, the only radioactive air emissions from PNNL were associated with two 

Battelle-owned facilities, the main building of the RTL (RTL-520) and the LSL-II. During that time 

PNNL issued separate annual airborne emissions reports for PNNL facilities, which were limited to the 

EMSL and the Battelle-owned buildings. The Battelle-owned facilities operated under a radioactive 

materials license from the State of Washington.

As of 2010, several radiological facilities in the PSF complex came online. PNNL therefore began 

issuing an annual radionuclide air emissions report to address those facilities. It also continued to 

provide the annual airborne emissions reports for the two Battelle-owned facilities (RTL and LSL-II).

The Hanford site has provided annual radionuclide air emissions reports and detailed environmental 

reports since before 2005.

In addition to airborne releases from Hanford and PNNL, individuals on the PNNL campus might also 

be exposed to radioactive airborne effluents from a number of commercial facilities that operate on 

the Hanford Site and in and around Richland, Washington. These include the Columbia Generating 

Station (a commercial nuclear power reactor) operated by Energy Northwest, a low-level radioactive 

waste disposal facility operated by US Ecology, a nuclear fuel fabrication facility operated by AREVA, 

and other commercial waste handling and decontamination facilities. The annual compliance 

calculations performed by both Hanford and PNNL are based on measured effluent data and 

therefore do not reflect potential dose from other, non-DOE sources.

Table 4-1 presents annual calculated doses that demonstrate compliance with the 10-mrem/yr limit for 

radioactive airborne emissions from the Hanford Site, Battelle-owned buildings within the PNNL 

campus, and from the PNNL radiological facilities within the PSF complex (notably Building 3410). 

For the separately reported Battelle-owned facilities at PNNL, the calculated dose from the LSL-II is 
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negligible relative to that from the RTL. Those contributors are therefore denoted in the table as 

PNNL – RTL. References are given for each annual dose value. The asserted doses are effective 

dose equivalent.

Table 4-1. Annual doses from radioactive airborne emissions reported by Hanford and PNNL (mrem). 

Year Hanforda PNNL – RTLb PNNL – PSFc

2005 0.039 0.3 Not applicable

2006 0.10 0.3 Not applicable

2007 0.14 0.3 Not applicable

2008 0.11 0.5 Not applicable

2009 0.068 0.5 Not applicable

2010 0.075 0.6 8.0E-06

2011 0.042 4.6E-04 1.7E-05

a. Sources for Hanford values by year: Diediker et al. (2006) and Rokkan et al. (2007, 2008, 2009, 2010, 2011, 2012). 

b. Sources for RTL values by year: Anderson (2006, 2007, 2008, 2009, 2010, 2011, 2012). 

c. Sources for PSF values by year: Snyder et al. (2011), Snyder, Barnett, and Bisping (2012). 

The reported annual dose values for Hanford include inhalation dose from routine and nonroutine 

point source emissions, plus that from diffuse and fugitive sources including radon. They represent 

the inhalation dose to a hypothetical maximally exposed individual 1.4 km east of the 300 Area on the 

east side of the Columbia River. 

The relatively higher annual dose values for the RTL are due to the low elevation of the release point 

(5 m) and the close proximity of the assumed receiver (300 m in 2005 to 2007 and 200 m in 2008 to

2011). The reported dose values for the RTL also include contributions from pathways other than 

inhalation. The significant decrease in the reported dose for 2011 in relation to previous years 

appears to be due to a decrease in the amount of 137Cs that was released. 

There were no radioactive airborne releases from other PNNL facilities until 2010, when several 

facilities in the PSF complex began radiological operations. The major defined emissions points for 

the PSF complex are the 3410, 3420, and 3430 buildings. All of these facilities discharge radioactive 

airborne effluents through stacks ranging in height from 13.5 m to 15.5 m. The effective discharge 

heights are greater. For 2010, the location for the maximally exposed individual was a location 480 m 

south-southeast of Building 3410 (Snyder et al. 2011). In 2011 this distance was increased to 550 m 

due to a change in the stack operating parameters (Snyder, Barnett, and Bisping 2012). 

The data in Table 4-1 show that inhalation doses to PNNL workers from radioactive airborne effluents 

from Hanford or PNNL are trivial and can be neglected for dose reconstruction purposes. Calculated 

annual environmental internal doses that total less than 0.001 rem for a specific radiation type and 

energy interval are not required to be included in the Interactive RadioEpidemiological Program 

(IREP) input sheet (ORAUT 2004). For such cases, the dose reconstruction should include 

appropriate discussions. 

4.3 EXTERNAL EXPOSURE FROM ONSITE AMBIENT SOURCES

Before 2006, the environmental program implemented by the Hanford Site contractors included the 

deployment of environmental TLDs for measuring the ambient external dose in a number of locations 

on and around the site. The locations included some on the PNNL main campus. The Hanford 

environmental external dose monitoring was discontinued after 2005 due to funding reductions. The 

Hanford Site environmental report for 2006 notes that the many years’ worth of external dose data 

“indicate that current radiation levels are at or near background levels, and are stable or decreasing 

as onsite cleanup activities progress” (Poston et al. 2007). 
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Table 4-2 shows the reported ambient external dose data for 2005 for locations on the Hanford Site 

and its perimeter, corrected from continuous occupancy to a period of 2,000 hours (Poston et al. 

2006, Tables 10.13.4 and 10.13.5). The values in the table represent the maximum values in the 

environmental report, which are the highest readings obtained for the measurements within a given 

distance classification (i.e., for a given area).

Table 4-2. Annual dose rates (mrem/yr) for the PNNL main campus and Hanford site locations.

PNNL 100 Area 200 Area 300 Area 400 Area 600 Area Sitewide (Hanford)

22 21 22 20 20 23 23

The values for the PNNL main campus are those reported for the 11 Hanford perimeter locations. 

The data for the Hanford Site itself are included for cases in which a location other than the PNNL 

main campus might be indicated. The annual dose rates in Table 4-2 should be used for all years 

from 2005 to the present.
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5.0 INTERNAL DOSIMETRY

5.1 INTRODUCTION 

5.1.1 Purpose

The purpose of this section is to describe internal dosimetry practices at PNNL to support dose 

reconstructions under EEOICPA.

5.1.2 Scope

Between January 1, 2005, and September 30, 2012, PNNL and Hanford shared the same internal 

monitoring program. PNNL managed those services for the Hanford site contractors. As of 

October 1, 2012, internal monitoring at Hanford was taken over by its mission support contractor. 

PNNL purchased internal monitoring services from that contractor for the remainder of 2012 before 

transitioning to its own program as of January 1, 2013. Internal dosimetry information for PNNL 

workers should therefore be obtained from the Hanford site profile for the period from January 1, 

2005, through December 31, 2012. 

The information in the remainder of this section pertains to the PNNL internal dosimetry program from 

January 1, 2013, to the present. Although the PNNL internal monitoring program then became 

separate from the Hanford program, PNNL and Hanford still used the same contractor (GEL 

Laboratories) for in vitro analyses. PNNL also still used the Hanford in vivo counting facilities.

Section 5.2 describes the PNNL internal dosimetry program in general and how individuals are 

selected for bioassay. Section 5.3 discusses the in vitro and in vivo monitoring methods and the 

sensitivities of those measurements. Internal dosimetry records are discussed in Section 5.4, and 

Section 5.5 presents tables that define codes in the databases PNNL uses for its internal monitoring 

and dose evaluation records. 

Except as otherwise noted, all information in this section was derived from (Maclellan 2013) and (BMI 

2013c). 

5.2 PNNL INTERNAL DOSIMETRY PROGRAM

Like other site services, PNNL radiation protection services are provided under a tenant model under 

which users acquire needed services on a fee basis. The internal dosimetry program offers the 

following services to PNNL employees and contractors who are monitored under the PNNL radiation 

protection program: 

• Administering the bioassay monitoring program, 

• Investigating and documenting evaluations of potential intakes for exposure records files, 

• Arranging for in vitro analysis services and ensuring that the analytical services contractor 

conforms to contractually specified technical requirements, 

• Maintaining DOE Laboratory Accreditation Program (DOELAP) accreditation for in vitro 

analyses (accreditation for in vivo analyses is maintained by the Hanford in vivo monitoring 

program, under which PNNL maintains a secondary accreditation), 

• Selecting and applying models, procedures, and practices for evaluating internal radionuclide 

depositions and doses, and
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• Providing guidance and support to PNNL in technical matters pertaining to internal dosimetry. 

The primary means of identifying potential radionuclide intakes at PNNL are air sampling, 

contamination surveys, and other field monitoring techniques. Bioassay is considered the primary 

means for confirming intakes and a secondary means for identifying intakes. 

The internal monitoring program is designed to be able to identify and confirm an intake resulting in a 

CDE of 100 mrem. For some materials, such as plutonium or type S uranium, detecting intakes that 

would result in this dose requires nonroutine bioassay that is triggered promptly by workplace 

monitoring. 

5.2.1 Routine Bioassay

As of June of 2013, no PNNL workers were on a routine bioassay schedule. Rather, PNNL adopted a 

modified version of the process (HPS 2001, Appendix A) to identify workers who require bioassay. 

This approach is based on potential dose per occupancy day, which is derived from information in 

RWPs. Bioassay is required for any radiation worker whose work history for the past 12 months 

indicates the potential for intake of radioactive material sufficient to result in a CDE of 100 mrem or 

more from all radionuclides. A unity rule approach is used if a worker is expected to work under 

multiple RWPs. If the assessment of the potential dose per occupancy day indicates a need for 

internal monitoring, then bioassay requirements are assigned based on the radionuclides the 

individual will be working with. Bioassay intervals are assigned for each radionuclide or group of 

nuclides that indicate the time since the potential intake during which a valid bioassay result should be 

obtained. 

The need for bioassay is monitored routinely via the Bioassay Management Tool, which is used to 

perform monthly reviews of radiologically controlled area entries in conjunction with information from 

RWPs. Derived air concentrations, annual limits on intake, occupancy factors, dispersibility factors, 

etc. all factor in to the determination of whether bioassay is required. The DAC-hr tracking 

requirements and action levels are discussed in Section 2.4.1. 

If the Bioassay Management Tool identifies the need for bioassay, one or more selections are made 

from the bioassay categories in Table 5-1 based on radionuclide information from the RWPs. In 

general, PNNL assumes that no more than four bioassay analyses will be needed to monitor for the 

mixtures of radionuclides to which its staff might be exposed. PNNL makes use of indicator 

radionuclides in determining bioassay requirements for established radionuclide mixtures. Table 5-2 

gives the analysis codes that correspond to the bioassay categories shown in Table 5-1. The analysis 

codes in Table 5-2 do not include all analysis types. See Tables 5-3 through 5-8 for a complete listing 

of analysis codes. 

Bioassay is not available for the bioassay category codes Remainder and AC227. The PNNL 

guidance in these instances is to “document equivalent monitoring action.” The radium isotopic 

analysis (IRA in the tables) is not included in the contractual detection level data, which indicates this 

analysis is not currently available.

Table 5-3 presents the same information as Table 5-2 but formatted to show the analysis codes that 

can be associated with a given bioassay category.

In addition to the potential dose per day approach described above, PNNL might in some instances 

implement routine bioassay for PNNL workers who are working at other sites where there is a 

potential for radioactive material intake.
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Table 5-1. PNNL bioassay category codes and intervals.

RWP radionuclide

Bioassay category 

code

Bioassay 

interval (mo)

H-3 Tritium 3

H-3-tritide Tritide 3

Pu-238, Pu-239 IPU 12

Cm-242, Cm-244 ICM 12

Cs-137, Co-60, MFAP,a tank waste Gamma 12

Ra-224, Ra-226 IRA 12

Sr-85, Sr-89, SR-90 SR 12

Th-228, Th-229, Th-230, Th-232 ITH 12

U, U-234, U-235, U-238, UDEP, UNAT IUS 3

U, U-234, U-235, U-238, UDEP, UNAT IU 12

Ac-227 AC227 12

Np-237 NP237 12

Am-241 AM241 12

Am-243 AM243 12

Any nuclide not in the above groups Remainder 12

a. MFAP = mixed fission or activation products.

Table 5-2. Bioassay analysis codes and the corresponding bioassay categories.

Analysis 

code Analytes or analysis Applicable bioassay categories

H3 Tritium Tritium, tritide

SR Total strontium SR

SR90 Sr-90 SR

NP237 Np-237 NP237

AM241 Am-241 AM241

AM243 Am-243 AM243

IPIU Sequential: Pu isotopic + U isotopic IPU

ICM Cm-242, Cm-244 ICM

IPA Sequential: Pu isotopic + Am-241 IPU, AM241

ITPAC Sequential: Pu isotopic + Am-241 + Cm isotopic IPU, ICM, AM241

IPSA Sequential: Pu isotopic + Am-241 + Total strontium IPU, SR, AM241

IUPU Sequential: Pu isotopic + U-238 IPU

IPS Sequential: Pu isotopic + Total strontium IPU, SR

IPU Pu isotopic IPU

IRA Ra isotopic IRA

ITH Th isotopic ITH

IU U isotopic IUS, IU

ICA Sequential: Cm isotopic + Am-241 ICM

GAMMA Gamma spectroscopy GAMMA

WB Whole-body counting GAMMA

5.2.2 Baseline and Termination Bioassay

Not all PNNL workers receive baseline or termination bioassays. Baseline bioassays might be 

ordered for personnel who have had previous intakes that might interfere with interpretation of future 

bioassay results or who have the potential to receive intakes of radioactive materials that might be 

present in urine from nonoccupational sources (e.g., uranium).

Termination bioassay is required for workers who participated in or qualified for participation in 

bioassay monitoring, unless it is documented that the worker was not potentially exposed to 

dispersible radioactive materials after the most recent bioassay or if a special evaluation of the 
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Table 5-3. Bioassay analyses applicable to each bioassay 

category.

Bioassay 

category Applicable analysis codes

Tritium H3

Tritide H3

IPU IPIU, IPA, ITPAC, IPSA, IUPU, IPS, 

IPU

ICM ICM, ITPAC, ICA

Gamma GAMMA, WB

IRA IRA

SR SR, SR90, IPSA, IPS

ITH ITH

IUS IU

IU IU

NP237 NP237

AM241 AM241, IPA, ITPAC, IPSA

AM243 AM243

Table 5-4. MDLs for routine processing of samples (dpm/sample unless otherwise noted).a

Analysis

Analysis 

code Constituents reported Urinalysis Fecal analysis

Pu isotopic IPU Pu-238, Pu-239+240 0.02 0.2

Pu isotopic IPUL Pu-238, Pu-239+240 0.005 Not applicable

Am-241 AM241 Am-241 0.02 0.2

Am-243 AM243 Am-243 0.02 0.2

Cm isotopic ICM Cm-242, Cm-244 0.02 Not applicable

U isotopic IU U-234, U-235, U-238 0.02 Not applicable

Th isotopic ITH Th-228, Th-229, Th-230, Th-232 0.1 1

Np-237 NP237 Np-237 0.02 0.1

Tritium H3 H-3 20 dpm/mL Not applicable

Sr – total SR sum: Sr-89 + Sr-90 10 Not applicable

Sr-90 SR90 Sr-90 10 Not applicable

Gamma spec. LEPD Am-241 5 Not applicable

U-236 mass U236 U-236 140 μg/sample Not applicable

U-238 mass U238 U-238 0.06 μg/sample 0.3 μg/sample

Pm-147 PM147 Pm-147 50 200

a. Use the MDAs for the individual analyses in the case of sequential analyses.

individual’s radiologically controlled area entries since the most recent bioassay has been completed 

showing that additional bioassay was not necessary. 

5.2.3 Special Bioassay

PNNL recommends special bioassay under the following circumstances, unless they are caused by 

radon progeny:

• Facial contamination that indicates a potential intake,

• Nasal contamination is present,

• Air monitoring indicates a potential for an intake that would result in a committed effective dose 

of 100 mrem or more,

• An unplanned intake is suspected for any reason, and

• Skin contamination is in excess of established thresholds for alpha or beta/gamma emitters.
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Table 5-5. MDLs for priority processing of samples (dpm/sample unless otherwise noted).a

Analysis

Analysis 

code Constituents reported Urinalysis Fecal analysis

Pu isotopic IPU Pu-238, Pu-239+240 0.02 0.2

Pu isotopic IPUL Pu-238, Pu-239+240 0.005 Not applicable

Cm isotopic ICM Cm-242, Cm-244 0.02 0.2

U isotopic IU U-234, U-235, U-238 0.02 0.3

Ra isotopic IRA Ra-224, Ra-226 0.3 1.5

Np-237 NP237 Np-237 0.02 0.1

Am-241 AM241 Am-241 0.02 0.2

Am-243 AM243 Am-243 0.02 0.2

Th isotopic ITH Th-228, Th-229, Th-230, Th-232 0.1 1

U-236 mass U236 U-236 140 μg/sample Not applicable

U-238 mass U238 U-238 0.06 μg/sample 0.3 μg/sample

Tritium H3 H-3 20 dpm/mL Not applicable

C-14 C14 C-14 10 dpm/mL 200

Sr – total SR sum: Sr-89 + Sr-90 10 30

Sr isotopic ISR Sr-89, Sr-90 30 45, 30 respectively

Sr-90 SR90 Sr-90 10 30

Pm-147 PM147 Pm-147 50 200

Pu-241 PU241 Pu-241 10 10

Gamma spec. LEPD Am-241 5 5

a. Use the MDAs for the individual analyses in the case of sequential analyses.

Table 5-6. MDLs for expedited processing of samples (dpm/sample unless otherwise noted).a

Analysis

Analysis 

code Constituents reported Urinalysis Fecal analysis

Pu isotopic IPU Pu-238, Pu-239+240 0.08 3

Cm isotopic ICM Cm-242, Cm-244 1.2 70

U isotopic IU U-234, U-235, U-238 0.12 4

Ra isotopic IRA Ra-224, Ra-226 0.3 3

Am-241 AM241 Am-241 0.08 6

Am-243 AM243 Am-243 0.08 6

Np-237 NP237 Np-237 0.02 0.1

Th isotopic ITH Th-228, Th-229, Th-230, Th-232 0.1 1

U-238 mass U238 U-238 0.5 μg/sample 5 μg/sample

Tritium H3 H-3 100 dpm/mL 100 dpm/mL

C-14 C14 C-14 20 dpm/mL 2,000

Pm-147 PM147 Pm-147 200 2,000

Sr – total SR sum: Sr-89 + Sr-90 50 150

Gamma spec. LEPD Am-241 5 5

a. Use the MDAs for the individual analyses in the case of sequential analyses.

In addition to the special bioassay drivers itemized above, PNNL workers can request bioassay.

5.3 PNNL INTERNAL MONITORING METHODS AND SENSITIVITIES

5.3.1 In Vitro Analyses

In vitro analyses for PNNL are performed by its analytical services contractor, a commercial vendor in 

Richland, Washington. The services include urinalysis, fecal analysis, and other nuclide-specific 

analyses.
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Table 5-7. MDLs for emergency processing of samples (dpm/sample unless otherwise noted).

Analysis

Analysis 

code Constituents reported Urinalysis Fecal analysis

Pu isotopic IPU Pu-238, Pu-239+240 0.5 9

Cm isotopic ICM Cm-242, Cm-244 10 240

U isotopic IU U-234, U-235, U-238 1 12

Ra isotopic IRA Ra-224, Ra-226 2 10

Am-241 AM241 Am-241 0.5 2

Am-243 AM243 Am-243 1 20

Np-237 NP237 Np-237 1 20

Th isotopic ITH Th-228, Th-229, Th-230, Th-232 1 10

U-238 mass U238 U-238 7 μg/sample 8 μg/sample

Tritium H3 H-3 100 dpm/mL Not applicable

C-14 C14 C-14 100 dpm/mL 10,000

Pm-147 PM147 Pm-147 1,000 10,000

Sr – total SR sum: Sr-89 + Sr-90 80 450

Gamma spec. LEPD Am-241 20 20

Tritium

The tritium monitoring program is based on tritium in the oxide form because that is the typically 

encountered type. Tritium bioassay is performed via liquid scintillation counting of tritiated water 

vapor in urine. Only 1 mL is required for analysis, so samples of essentially any volume can be used. 

Samples are collected as single void or simulated 12-hour.

Strontium

Samples are counted using a gas proportional counter after chemical processing. Separation 

chemistry for yttrium is performed if 90Sr identification is desired.

Uranium

Urinalysis for uranium can be done on a mass basis or isotopically by alpha spectrometry. Massbased analysis is used for DU, NU, and RU. A screening level of 0.2 μg/d is used for the upper range 

of the normally expected excretion rate, reflecting an occupationally attributable excretion rate of 

0.18 μg/d above the geometric mean environmental level of 0.02 μg/d established for the PNNL 

workforce.

Simulated 24-hour samples are collected for uranium urinalysis via alpha spectrometry. Screening 

levels for uranium isotopic analysis are 0.16 dpm for 233+234U, 0.007 dpm for 235U, and 0.15 dpm for 

238U. These values correspond to an excretion rate of 0.2 μg/d for NU. Values above the screening 

levels trigger an investigation and perhaps a dose evaluation. 

Transuranic Materials

Urinalyses can be performed for plutonium, americium, neptunium, and curium. After chemical 

processing, appropriate tracers are used to determine the chemical yield of the process. Split 

samples or multiple samples are used to accommodate analyses for multiple isotopes for a given 

individual. 

Other Radionuclides

In addition to those above, PNNL has included the following analyses in the urinalysis sensitivity 

requirements it specifies to its analytical services contractor: 

• Radium isotopic, 

• Thorium isotopic, 

• 14C, 

• 147

Pm, 
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• 241Pu, and

• Gamma spectrometry for 241Am. 

5.3.2 In Vivo Counting

In vivo measurements for PNNL are performed by the Hanford mission support contractor at the 747A 

Building. The in vivo counting systems available to PNNL through this arrangement include the 

following: 

• Whole-body counting using a stand-up counter containing multiple sodium iodide (NaI) 

detectors, 

• Whole-body counting or scanning using a coaxial germanium system, and

• An array of planar hyperpure germanium (HPGe) detectors that can be used in multiple 

configurations. 

The stand-up NaI system is also known as the preview counter. For whole-body counts, PNNL has 

the option of requesting a screening count using the preview counter, or a 10-minute count using the 

coaxial germanium system. 

The HPGe detectors can be used in a variety of geometries including chest counting, skull counting, 

liver counting, thyroid counting, lymph node counting, and wound counting. 

Chest Counting

Chest counting is performed by placing an array of four planar HPGe detectors anteriorly over the 

individual’s lungs, in light contact with the chest. Chest wall thickness correction is made based on 

height-to-weight ratio. Ultrasound measurement of chest wall thickness is used in cases of known 

depositions. Chest count measurements that have been corrected for contributions from deposition in 

the skeleton and, perhaps, the liver, represent lung burden assays.

A chest count result can be considered to be a conservative estimate of a lung burden except for 

short periods after intake. The need for correcting a chest count result for skeletal deposition is 

usually determined from a skull count once a lung deposition has been confirmed. Likewise, liver 

counts can be used to determine if a correction for liver deposition is needed. Liver counts are

primarily used for long-term follow-up and to confirm deposition estimates from urinalysis 

(Maclellan 2013). 

The Hanford chest counting system routinely reports 241Am, 235U, and 234Th. A peak search algorithm 

is used to identify the presence of other gamma lines. The normal counting time is 3,000 seconds, 

but longer counts can be used if greater sensitivity is needed. If a radionuclide is detected, the 

individual is asked to shower and change into clean coveralls to be recounted. Recounts are typically

3,600 seconds in duration. The 3,000- and 3,600-second counts can then be summed, if desired, for 

greater sensitivity. 

Uranium detection is achieved by measuring photon emissions from 235U and 234Th. The latter is 

normally assumed to be in secular equilibrium for uranium contamination at PNNL. However, this 

assumption is not correct if the uranium source material has been separated from its progeny 

elements within about 6 months. PNNL has determined that in vivo measurement of 234Th and 235U 

are roughly equal in terms of their ability to indicate intakes of RU. Thorium-234 is a more sensitive 

indicator for natural or DU. 
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Other In Vivo Counting Geometries

In addition to chest counting, the Hanford planar HPGe system is also used in the following 

geometries: 

• Skull counts for skeletal deposition are performed by placing planar germanium detectors on 

the forehead. A typical count time is 3,000 seconds.

• Liver counts are performed by placing an array of HPGe detectors anteriorly over the liver. 

The routine counting time is 3,000 seconds. The liver counting geometry is calibrated using 

241Am.

• Thyroid counts are performed by placing a single HPGe detector 10 cm above the thyroid. 

The routine counting time is 600 seconds.

• Wound counts are performed using a single HPGe detector. The typical count time is 

10 minutes. The Hanford in vivo program also has portable germanium systems for 

performing wound counts in field locations.

Other types of in vivo counts performed for plutonium at Hanford include lymph node counts and 

scanning lung counts. 

5.3.3 Bioassay Measurement Sensitivities

5.3.3.1 In Vitro Analyses

PNNL specifies reporting levels to its analytical services contractor and requires the contractor to 

provide prompt notification in the event a routine bioassay result exceeds a specified level. Analytical 

requests to the contractor include one of four processing categories: routine, priority, expedited, and 

emergency. All results are to be promptly reported in the case of priority, expedited, or emergency 

analysis requests. PNNL compares the analytical results with its screening levels to determine if 

additional investigation or dose assessment are warranted. 

Tables 5-8 to 5-10 list the contractually mandated minimum detection levels (MDLs) PNNL has 

established with its analytical services contractor as of April of 2013. Values are given for each of the 

four processing categories. These values should be used for missed dose assignments in cases in 

which an analysis-specific detection limit or MDA is not reported. 

In addition to meeting the detection levels for gamma spectrometry analyses in Table 5-8, PNNL also 

requires that the analytical services contractor “resolve and quantify” gamma-emitting radionuclides in 

submitted urine samples at a nominal detection level of 20 dpm relative to the energy and photon 

abundance for 137Cs. That criterion applies to routine processing for isotopes having a half-life greater 

than 30 days and energy greater than 100 keV. For priority processing the detection level criterion is 

twice the propagated uncertainty. 

5.3.3.2 In Vivo Analyses

For whole-body counts, PNNL has the option of requesting a screening count using the preview 

counter (a stand-up NaI detector system), or a 10-minute count using the coaxial germanium system. 

Nominal MDA values for each system are given in Table 5-9.
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Table 5-8. Contractual MDLs for processing 

of gamma spectroscopy samples 

(dpm/sample).

a

Isotope MDL

Co-60 35

Cs-134 20

Cs-137 20

Mn-54 20

Na-22 25

Ru-106 120

a. The MDL is in dpm/L for sample volumes ≥1 L.

Table 5-9. Nominal MDAs for the Hanford whole-body counters (nCi).

Isotope Preview countera Coaxial germanium counterb

K-40 10 7

Co-60 1.2 0.8

Cs-137 1.3 1.0

Eu-154 7.0 2.0

a. Count time = 200 s.

b. Count time = 600 s.

The Hanford in vivo facility includes an array of planar HPGe detectors that can be used in a variety of 

geometries including chest counting, skull counting, liver counting, and thyroid counting (single 

detector). Nominal MDA values for these configurations are given in Table 5-10.

Table 5-10. Nominal MDA values for the Hanford planar 

HPGe detectors in various configurations (nCi).a

Geometry and isotope MDA

Normal chest, Am-241 0.16

Normal chest, U-235 0.09

Normal chest, Th-234 1.5

Skull for skeletal burden, Am-241 0.5

Liver, Am-241 0.17

Thyroid, I-125 0.1

Thyroid, I-129 0.2

Thyroid, I-131 0.26

a. Normal chest: 3,000-s count, four detectors, avg. size subject.

Skull: 3,000-s count, two detectors on forehead.

Liver: 3,000-s count, three detectors over liver, avg. size subject.

Thyroid: 600-s measurement, one 38-cm2 detector 10 cm above 

thyroid.

5.4 INTERNAL DOSIMETRY RECORDS

Before 2013, radiological exposure records and reports for PNNL workers were maintained by the 

Hanford mission support contractor; PNNL has managed them since then.

The primary PNNL record of internal dose is the internal dose evaluation report. These reports are 

issued for each assessed internal exposure and are stored in the radiation exposure files by the 

Radiation Protection Division.

Dose information from the internal dose evaluation report, including summary intake information, is 

maintained by the Radiation Protection Division in the Sentinel database. The Sentinel database also 

includes bioassay results and other information. The Sentinel database contains information for 

PNNL-monitored workers as of 2013. Before then such information is found in the Hanford REX 
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database, and a subset of that database known as the Internal Dose Tracking System (INTERTRAC) 

database. 

The following tables list the definitions and meanings of various codes that might be encountered 

when reviewing internal dosimetry data from the Sentinel and INTERTRAC databases (Tables 5-11 to 

5-26 and Tables 5-27 to 5-30, respectively).

Table 5-11. Sentinel analysis codes and reported 

isotopes.

Code Isotopes

AM241 Am-241

AM243 Am-243

C14 C-14

H3 H-3

IPIU Pu-238, Pu-239, U-234, U-235, U-238

ICA Am-241, Cm-242, Cm-244

ICM Cm-242, Cm-244

IPA Pu-238, Pu-239, Am-241

IPIU U-234, U-235, U-238, Pu-238, Pu-239

IPS Pu-238, Pu-239, SR

IPSA Pu-238, Pu-239, SR, Am-241

IPU Pu-238, Pu-239

IPUB Pu-238, Pu-239, Pu-241

IPUBA Pu-238, Pu-239, Pu-241, Am-241

IPUL Pu-238, Pu-239

IRA Ra-224, Ra-226

GAMMA See Table 5-8

ISR Sr-89, SR-90

ITH Th-228, Th-229, Th-230, Th-232

ITPAC Pu-238, Pu-239, Am-241, Cm-242, Cm-244

IU U-234, U-235, U-238

IUPU U-238, Pu-238, Pu-239

LEPD Am-241

NP237 Np-237

PM147 Pm-147

SR Total Sr

SR90 SR-90

U236 U-236

U238 U-238

Table 5-12. Sentinel analytical laboratory codes.

Code Laboratory

GL General Engineering Laboratories (GEL Laboratories)

IT IT Analytical Services – Richland

LA Los Alamos National Laboratory

OR Oak Ridge National Laboratory

PL PNNL Analytical Chemistry Laboratory

QN Quanterra

RE Reynolds Electrical and Engineering Company, Nevada Test Site)

ST Severn Trent Laboratories – Richland, Washington

TA TMA/Norcal, Richmond, California

WH Westinghouse Hanford Company 222-S Lab
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 39 of 57

Table 5-13. Sentinel bioassay 

frequency codes.

Code Bioassay frequency

A Annual

B Biennial

D Special day

F 5 yr

Q Quarterly

S Semiannual

M Monthly

W Weekly

X Biweekly

Table 5-14. Sentinel 

sample type codes.

Code Type of sample

B Blood

F Feces

S Sputum

T Tissue

U Urine

W In vivo

Table 5-15. Sentinel bioassay measurement reason codes.

Code Name Description

BL Baseline Measurement is performed to establish a reference level against which subsequent 

measurements are compared. Generally, this might be for new employees or for 

established employees before commencing work with radioactive materials, 

beginning a specific type of radiation zone work, or making an offsite trip where 

potential intakes could occur.

PR Periodic Periodic measurement is performed at a regularly scheduled interval.

EA End of 

assignment

Measurement is performed after completion of specific work assignment, but not 

end of employment.

SP Special Measurement is performed as part of a specific investigation of potential internal 

dose. Might include response to off-normal work conditions or follow-up of 

abnormal periodic measurements.

CR Contractor 

request

Measurement requested by employer for reasons other than periodic, baseline, end 

of assignment, or special investigation.

RA Reanalysis A First repeat in vivo measurement or second aliquot analysis of an excreta sample.

RB Reanalysis B Second repeat in vivo measurement or third aliquot analysis of an excreta sample.

RC Reanalysis C Third repeat aliquot analysis of an excreta sample.

R1 Recount 1 First recount of original excreta sample or repeat in vivo examination.

R2 Recount 2 Second recount of original excreta sample or repeat in vivo examination.

QR Quality and 

research

Measurement performed as part of quality control, quality assurance, or research 

work.

TM Termination Final bioassay at termination of employment.

12 Contract 

work

In vivo measurement performed under contract to customers rather than for Hanford 

employees.

20 Source count In vivo source count made for system calibration or as a function check, usually 

using a known check source.

30 Background 

count

In vivo system background measurement performed for system calibration or as a 

functional check.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 40 of 57

Table 5-16. Sentinel bioassay sample kit codes and sample date conventions.

Kit code for 

delivery and 

retrieval

Kit code for 

pick-up Media Description

1 P Urine Approximate 24-hour urine collection. Collected at home over a 

2-d period. Used for routine sampling and when a larger volume 

sample is desired. Designated sample date is the day after kit 

delivery to the employee.

2 Q Urine Approximate 12-hr urine collection for termination sampling only. 

Collected at home overnight. Designated sample date is the day 

after kit delivery to the employee.

3 R Urine Total 24-hr urine collection. Collected at home and at work (if 

necessary) to collect all urine voided during a 24-hour period. 

Generally used for sampling immediately after an occurrence or 

for work restriction sampling. Designated sample date is the day 

after delivery or the date on which the sample collection began.

4 S Urine Single void (spot urine) collection. Collection in a single bottle, 

used for initial indications of an intake or when small sample 

volumes are adequate. Designated sample date is the date of 

voiding.

5 T Feces Collection of a single fecal voiding usually for investigation of a 

potential intake. Sample date is the day after kit delivery or date 

on which the sample was voided.

6 U Urine Partial day or approximate 12-hr collection. Usually collected at 

home overnight. Used for collection after an occurrence or when 

a large volume urine sample is not necessary. Designated 

sample date is the date of delivery to the employee.

7 V Urine Approximate 12-hr collection Sunday-Monday sample (Friday 

delivery only). Generally used for workers chronically exposed to 

soluble uranium. Designated sample date is the Sunday in the 

sampling period.

8 W Feces Collection of a single fecal voiding used for a special program for 

plutonium oxide workers. Designated sample date for shift 

workers is the Tuesday of long shift change, and for day workers 

is the appropriate Sunday.

9 X Urine Kit designed for collection of urine outside the local service area. 

Transportation is handled by private carrier. Generally used for 

termination samples not collected locally.

A Y Urine Simulated 48-hr urine collection. Collected at home over a 4-d 

period. Used for IPUL sampling. Designated sample date is 2 d 

after kit delivery to the employee.

B Not 

applicable

Urine 12-hr urine collection for termination sampling only. Collected at 

home overnight. Kit delivered in normal manner, but brought to a 

designated onsite location by worker for pick-up by Contractor. 

Designated sample date is the day after the date of kit delivery to 

the employee. Delivery only, no home pick-up required.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 41 of 57

Table 5-17. Sentinel sample status codes.

Code Description

CN Container not out

DL Kit delivered

FA Failed analysis

IS Insufficient volume

LC Lost container

ND Kit not delivered

NE Not evaluated

NS No sample

OR Order received by laboratory

PE Pending (after 03/31/2005: awaiting instructions for analysis)

PN Pending (before 04/1/2005)

RV Valid sample received

Table 5-18. Sentinel bioassay 

sample processing codes.

Code Description

R Routine

P Priority

X Expedited

E Emergency

Table 5-19. Sentinel no-sample codes.

Code Description

CS Cancelled sample/analysis.

CT Sample lost due to bioassay analysis contract termination.

FA Failed analysis. A valid analytical result could not be obtained.

IS Insufficient sample volume.

LC Lost container. Sample kit not retrieved.

ND Not delivered. Sample scheduled but kit never delivered.

NE Not evaluated. Sample was collected but not analyzed. Typically 

used when a backup sample was obtained but was not needed.

NS No sample. Kit retrieved but worker did not provide a sample.

Table 5-20. Sentinel unit codes.

Code Units

DPM/S dpm/sample

UG/S μg/sample

UCI/L μCi/L

NCI nCi

UCI μCi/sample

DPM/ML dpm/mL

Table 5-21. Sentinel person codes.

Code Description

E Employee

F Fetus

N Nonresident

S Subcontractor (this code is inactive)

V Visitor
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Table 5-22. Sentinel in vivo analysis request codes.

Code Analysis

CA Chest count for Am-241 only

CC Chest count for Am-241, U-235, and Th-234

CU Chest count for U-235 and Th-234

Table 5-23. Sentinel in vivo schedule (measurement) type 

codes.

Code Measurement type

C Chest count

C2 Extended chest count

HC Head and chest count

HD Head count

H2 Head and extended chest count

LC Liver and chest count

LV Liver count

LY Lymph node count

TC Thyroid and chest count

TH Thyroid count

WB Whole-body count

WC Whole-body count: coaxial germanium counter

WD Wound count

Table 5-24. Sentinel in vivo detector codes.

Code Type of detector or counting cell

CC Coaxial germanium counter for whole-body counts

DS Stainless Steel Room with digital signal processing

LD Lead Room for special counting geometries

CH Lead Room for special counts

SU Stand-up whole-body counter

SS Stainless Steel Room for lung count

IR Iron Room counter for lung count

The above in vivo detector codes are those that were typically in use by the Hanford in vivo counting 

group as of August 2003. See the Hanford site profile for additional information on these detectors 

and facilities.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 43 of 57

Table 5-25. Sentinel body location codes for in vivo counts.

Code Body location

ABD Abdomen

CA1 Chest – americium

CA2 Chest – americium corrected by ultrasound

CC1 Chest – combination: americium and uranium

CC2 Chest – combination: americium and uranium corrected by ultrasound

CHT Chest result

CH1 Chest result

CH2 Chest result corrected for chest wall thickness via ultrasound

CU1 Chest – ultrasound

CU2 Chest – uranium corrected by ultrasound

HND Hand

KNE Knee

LG1 Lung result: chest result corrected for skeleton burden interference

LG2 Lung result: chest result corrected for skeleton and liver burden interference

LV1 Liver

LV2 Liver result corrected for skeleton burden interference

LV3 Liver result corrected for skeleton and lung burden interference

LYM Lymph nodes

SK1 Skeleton result based on a head count

SK2 Skeleton result based on something other than a head count

SPL Special

THX Thorax

THY Thyroid

TRY Throat

WBD Whole body

WND Wound

Table 5-26. Sentinel in vivo no-result codes.

Code Reason for no result

C External contamination other than radon detected on the subject. Measurement invalid; no 

results obtained.

F Failure of equipment or faulty setup of equipment. Measurement invalid; no results 

obtained.

I Interference from localized activity in another part of the subject’s body. Measurement 

invalid; no results obtained.

L Location of internal or external activity was qualitatively determined by mapping, masking, 

or collimating. May include one or more measurement counts. These measurements are 

qualitative for identifying location of activity and do not yield quantifiable estimates of 

activity.

M Medically administered radioactivity interfered with measurement. Measurement invalid; no 

results obtained.

N No show. Worker did not meet appointment.

P Preliminary count, when followed by a more accurate record count. Used to indicate 

measurement taken, but not a record count.

R Radon interference from subject's clothing, hair, or skin. Measurement invalid; no results 

obtained.

S The subject’s actions interrupted completion of the count. Measurement invalid; no results 

obtained.

W Waived. Scheduled exam was waived based on needs review.

X Measurement invalid; no results obtained. Other no-result codes do not apply. See 

comment field for a brief description.

Z Test case.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 44 of 57

 

Table 5-27. INTERTRAC mode of intake codes.

Code Mode of intake

ABS Absorption

ING Ingestion

INH Inhalation

NON None (no intake)

UNK Unknown

WND Wound

Table 5-28. INTERTRAC evaluation reason codes.

Code Reason for evaluation

A Annual chronic intake evaluation

C Contractor requested evaluation

D DAC-hours evaluation

H High routine bioassay evaluation

I Incident evaluation

N New hire measurement or previous employment 

record indicated exposure before Hanford 

employment

R Reevaluation

Table 5-29. INTERTRAC source of intake codes.

Code Source of intake

DHE Intake at DOE site while employed at Hanford

HAN Intake at Hanford

NHE Intake at non-DOE site while employed at Hanford

NOC Nonoccupational intake

PTH Intake occurred before Hanford employment

Table 5-30. INTERTRAC miscellaneous codes.

Code Type Code Description

Intake confirmed Y Yes (occupational intake)

Intake confirmed N No

Nature of intake A Acute

Nature of intake C Chronic

Recorded dose Y Yes (occupational intake)

Recorded dose N No

Recorded dose O Undetermined: old evaluation assessing body 

burden rather than dose, or an evaluation in process

Recorded dose Z Recorded dose is zero mrem

Source known Y Yes

Source known N No

Type of evaluation P Preliminary

Type of evaluation F Final
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6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 INTRODUCTION

6.1.1 Purpose

The purpose of this section is to detail historical external dosimetry programs, systems, and practices 

at PNNL. This information may be used by dose reconstructors as needed to evaluate external 

occupational doses for EEOICPA claimants.

6.1.2 Scope

This section describes PNNL external monitoring practices as of January 2013. Before then a 

common external monitoring program was in place for both PNNL and Hanford. Information about 

external monitoring at PNNL during 2005 through 2012 should therefore be obtained from the Hanford 

site profile. 

6.2 BACKGROUND

Before 2012, PNNL or one of its predecessors had provided dosimetry services for both Hanford and 

PNNL since Hanford’s beginning in the early 1940s. A single DOELAP accreditation had been 

maintained for external dosimetry at both sites since the DOELAP program began in the late 1980s. 

This arrangement continued until October 1, 2012, when the responsibility for radiation dosimetry 

services for Hanford transferred from PNNL to the mission support contractor for the Hanford site. At 

that time PNNL began maintaining its own DOELAP accreditation and providing external dosimetry 

services to PNNL, the DOE Pacific Northwest Site Office, and their subcontractors. The PNNL 

DOELAP accreditation is sponsored by this office (Rathbone 2012). 

Between October 1, 2012, and January 1, 2013, PNNL purchased dosimetry services from the 

Hanford mission support contractor. PNNL began using its own commercial vendor for external 

monitoring on January 1, 2013. The PNNL external dosimetry program is asserted to comply with 

DOE requirements in 10 CFR Part 835 and DOE Standard 1095-2011 for DOELAP (DOE 2011) as 

well as guidance in DOE G 441.1-1C, Radiation Protection Programs Guide (Rathbone 2012). 

6.2 DOSE RECONSTRUCTION PARAMETERS

6.2.1 Administrative Practices

The PNNL radiation control program requires that personnel dosimeters be provided to and used by 

the following groups (BMI 2013b, Article 511.1): 

• Radiological workers who, under typical conditions, are likely to receive one or more of the 

following:

– An effective dose to the whole body of 100 mrem or more in a year, 

– An equivalent dose to the skin or to any extremity of 5 rem or more in a year, or

– An equivalent dose to the lens of the eye of 1.5 rem or more in a year.

• Declared pregnant workers who are likely to receive from external sources an equivalent dose 

to the embryo/fetus in excess of 10% of the applicable limit at 10 CFR 835.206(a). 

• Occupationally exposed minors likely to receive a dose in excess of 50% of the applicable 

limits at 10 CFR 835.207 in a year from external sources.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 46 of 57

 

• Members of the public entering a controlled area likely to receive a dose in excess of 50% of 

the limit at 10 CFR 835.208 in a year from external sources.

• Individuals entering a high or very high radiation area.

In addition, the PNNL radiation control program requires that neutron dosimetry be provided if any of 

the above dose criteria are likely to be exceeded from neutron exposure. Further, the program 

manual states that DOELAP-accredited neutron dosimetry should be used whenever an individual is 

likely to meet or exceed the applicable dose criteria and 10% or more of the dose is likely to be from 

neutron exposure (BMI 2013b, Article 511.2). 

Article 512 of the PNNL radiation control program requires that external dose monitoring programs

implemented to demonstrate compliance with 10 CFR 835.402(a) shall be adequate to demonstrate 

compliance with the dose limits established in Subpart C of 10 CFR Part 835 and shall be either 

accredited, or excepted from accreditation, in accordance with the DOE Laboratory Accreditation 

Program for Personnel Dosimetry; or determined by the Secretarial Officer responsible for 

environment, safety and health matters to have performance substantially equivalent to that of 

programs accredited under the DOE Laboratory Accreditation Program for Personnel Dosimetry.

In general, PNNL does not engage in radiological operations that have a potential for significant 

external dose. Even with the Hanford 300 Area are included, over 80% of external dose results for 

PNNL workers are zero (i.e. are less than the reporting level). PNNL maintains an area dosimetry 

program to demonstrate that external doses to individuals working in radiological buffer areas or other 

areas where personnel dosimetry is not required do not exceed an effective dose of 50 mrem annually 

(see Section 2.4.2). 

Transit Dosimeters

PNNL maintains a manual reader system for its dosimeters that it uses to evaluate transit dosimeters 

included in shipments from its dosimetry vendor. The onsite manual reader can also be used to read 

personnel dosimeters in cases when an individual might be nearing an administrative dose limit 

(Rathbone 2012). 

6.2.2 Site Dosimetry Technology

6.2.2.1 Whole-Body Dosimeters

PNNL uses an external vendor (Landauer) for its personnel monitoring program. Two Landauer 

dosimeters are used:

• The InLight Model 2 (for electrons and photons), and

• The InLight Model 2T (for electrons, photons, and neutrons).

The InLight Model 2 and Model 2T are identical except the latter includes a CR-39 track-etch neutron 

detector. Both designs use optically stimulated luminescence, which is nondestructive, meaning a 

dosimeter can be reread or read on an interim basis without significantly altering the final result. 

The InLight dosimeters use Al2O3:C as the phosphor. Al2O3:C is not tissue-equivalent and 

significantly overresponds (relative to soft tissue) at photon energies less than 100 keV. Element 

filters and a dose calculation algorithm are therefore used to obtain accurate results. Each dosimeter 

contains four elements filtered by plastic, copper, or aluminum. There is also an open window with 

minimal filtration. Table 6-1 gives the filter thicknesses from the front and the rear for each of the four 

dosimeter elements (Rathbone 2012). 
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 47 of 57

Table 6-1. Front and rear filtration thickness for the InLight Model 2 and 

Model 2T dosimeter elements (mg/cm2

).

Aspect

Open 

window Plastic Aluminum Copper

Front 30 298 398 568

Rear 16 284 384 554

In addition to the same four elements of the InLight Model 2 dosimeter, the Model 2T version also 

contains a Landauer Neutrak 144 CR-39 detector capable of measuring neutron dose from both fast 

and thermal neutrons. The CR-39 detector is a plastic foil made of poly-allyl diglycol carbonate 

(C12H18O7). This material is commonly called CR-39, which is a trade name given to it by one of the 

original manufacturers. Fast neutrons are detected in CR-39 by recording tracks of recoil protons

from (n,p) reactions. Thermal neutrons are detected through (n,α) reactions in a boron-loaded Teflon 

filter (also called a radiator) that covers part of the CR-39 chip. The other part of the chip is covered 

by a polyethylene filter. The area behind the polyethylene filter is sensitive to fast neutrons in the 

energy range from 40 keV to 40 MeV. The area behind the boron-loaded Teflon filter responds to 

both fast neutrons and thermal neutrons in the energy range from 0.25 eV to 40 eV (Rathbone 2012).

The CR-39 neutron dosimeter in the InLight Model 2T dosimeter has a useful measurement range of 

20 mrem to 25 rem for fast neutrons and 10 mrem to 5 rem for thermal neutrons (Rathbone 2012).

The InLight Model 2 and Model 2T are capable of detecting photons (X and gamma rays) with 

energies between 5 keV and 20 MeV, with a useful measurement range from 5 mrem to 1,000 rem. 

The InLight Model 2 and Model 2T are capable of detecting beta particles with energies greater than 

150 keV (average) with a useful measurement range from 20 mrem to 1,000 rem.

There is limited potential for neutron exposures at PNNL. Sources of Cf-252 are known to have been 

stored in safes in Room 1344 in the EMSL and Room 1303 in Building 3440. The PNNL area 

dosimetry program included neutron monitoring in those areas to show any neutron dose received by 

unmonitored personnel was less than 50 mrem per year. There could be some potential for neutron 

generation associated with accelerator operations depending on the particles and energies involved. 

PNNL’s criteria for neutron monitoring are discussed in Section 6.2.1. In addition to the other criteria 

for personnel external monitoring, neutron monitoring is recommended whenever 10% or more of an 

individual’s dose is likely to be from neutron exposure.

6.2.2.2 Extremity Dosimetry

PNNL issues extremity dosimetry to individuals who are likely to be exposed to nonuniform radiation 

fields where the shallow dose to the extremities is likely to significantly exceed the shallow dose to the 

whole body. The PNNL extremity dosimeters are calibrated to measure Hp(0.07) when worn on the 

finger.

PNNL uses the Landauer U-Ring dosimeter. It contains a single Harshaw TLD-100 (LiF) chip in a 

polystyrene cap that provides a density thickness of 40 mg/cm2

. The useful range for the U-Ring 

dosimeter is considered to be 30 mrem to 500 rem shallow dose equivalent (SDE) (Rathbone 2012).

6.3 DOSE CALCULATION AND REPORTING

6.3.1 Dose Algorithm

The dose algorithm used with the InLight Model 2 whole-body dosimeter was designed to measure 

total personal dose equivalent at depths of 7,300 and 1,000 mg/cm2 from radiation fields including 

photons from 20 keV to over 6 MeV and high and low energy beta radiations. Additionally, the 
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 48 of 57

 

algorithm is capable of accurate assessment of mixed fields including mixtures of the photons or 

photons and betas from the above fields. The dose algorithm was based upon correction factors 

derived from the American National Standard HPS N13.11-2001 (HPS 2001) which incorporates the 

standard beams listed in ISO 4037-4:2004 (ISO 2004). 

6.3.2 Exposure Energy Spectra

Dose reconstruction under the NIOSH program requires estimates of exposure percentages within 

specific energy bands for each type of radiation as follows (NIOSH 2007):

• Photons: 

– <30 keV, 

– 30–250 keV, and

– >250 keV; 

• Beta particles: 

– <15 keV, and

– >15 keV; 

• Neutrons: 

– <10 keV, 

– 10–100 keV, 

– 100 keV–2 MeV, 

– 2–20 MeV, and

– >20 MeV. 

For external exposures, betas with energies of less than 15 keV are not applicable. Otherwise, the 

energy spectrum choices that are most favorable to the claimant should be used when performing 

organ dose conversion from reported external dosimetry results from PNNL unless other, exposurespecific spectral information is available. For photon exposures, the 30 – 250 keV energy range is 

typically the most favorable to the claimant for likely noncompensable claims (i.e. when maximizing 

assumptions are desired). For likely compensable claims (minimizing assumptions) a photon energy 

split of 25% less than 30 keV photons and 75% greater than 250 keV photons should be used. For 

neutron exposures, the 0.1 – 2 MeV energy range is typically the most favorable to the claimant 

(ORAUT 2006). 

There is an exception to the above favorable to the claimant photon energy group selections for cases 

involving dose to the skin from work with or near isotopes of plutonium. In that case, since the dose 

conversion factor for the skin is considered to be 1.0 for all energies, one should select the less than 

30 keV photon energy group given the higher risk factor for that group relative to the 30 – 250 keV 

group in the Interactive RadioEpidemiological Program (ORAUT 2006). 

The dosimetry results reported by Landauer are in units of personal dose equivalent and already 

reflect current values for the applicable radiation and tissue weighting factors. Hence, beginning in 

2013 when PNNL adopted its own external dosimetry program, no adjustment should be needed 

before converting the reported results to organ dose. The external dosimetry program at use at PNNL 

(which was the same as the Hanford program prior to 2013) adopted the neutron radiation weighting 

factors from ICRP 60 as of 2011 (ORAUT 2011). For neutron results reported prior to 2011, 

corrections should be applied for any differences between the neutron radiation weighting factors (or 
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 49 of 57

 

quality factors) used by the Hanford program and those from ICRP 60, as discussed in the Hanford 

site profile or other, applicable project guidance. 

6.3.3 Missed Dose

Lower limit of detection (LLD) data for the InLight Model 2T dosimeter are presented in Table 6-2 for 

exchange periods of 45, 75, and 100 days. The data were calculated from DOELAP testing results for 

low-energy photon, high-energy photon, general beta particle, DU beta, and unmoderated neutron 

fields. 

Table 6-2. LLD values for the InLight Model 2T dosimeter (mrem).a

Irradiation category

45-d

SDE

45-d

DDE

45-d

neutron

75-d

SDE

75-d

DDE

75-d

neutron

100-d

SDE

100-d

DDE

100-d

neutron

Low-energy photon 11 9 N/A 18 17 N/A 21 17 N/A

High-energy photon 8 8 N/A 13 14 N/A 16 14 N/A

Beta – general 13 N/A N/A 22 N/A N/A 25 N/A N/A

Beta – DU 13 N/A N/A N/A N/A N/A N/A N/A N/A

Unmoderated neutron N/A N/A 13 N/A N/A 11 N/A N/A 17

a. N/A = not applicable.

Landauer’s reporting levels for dose results from the InLight Model 2T dosimeter are:

• SDE, 5 mrem; 

• Deep dose equivalent (DDE), 5 mrem; 

• DDE from fast neutrons, 20 mrem; and

• DDE from thermal neutrons, 10 mrem. 

For photon and beta exposures it is anticipated that the LLD value that is most favorable to the 

claimant should be selected in most cases. For SDE this value is 25 mrem. For DDE it is 17 mrem.

The LLD values for neutron exposures in Table 6-2 are below the reporting level of 20 mrem DDE for 

fast neutrons, so 20 mrem should be used as the LLD for fast neutron exposures. 

Landauer’s reporting level for the U-Ring extremity dosimeter is 30 mrem. This value is greater than 

the accepted value for the LLD for this dosimeter and should therefore be applied as the LLD when 

needed. 

Reporting levels given on dosimetry reports take precedence over the values given in this site profile. 

PNNL uses a combination of monthly, quarterly, or annual dosimeter exchange frequencies 

depending on the expected dose. Supplemental dosimetry is used to track dose between primary 

dosimeter reads. The actual exchange frequency or frequencies need to be ascertained from claim 

records (i.e. from the monitoring period given on the dosimetry reports). 

7.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0027 Revision No. 02 Effective Date: 09/12/2016 Page 50 of 57

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127453]

Anderson, B. G., 2008, Calendar Year 2007 Battelle Richland Private Buildings Radioactive Airborne 

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127526]

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Laboratory, Richland, Washington. 

Bivins, S. R., and G. A. Stoetzel, 2008, Area Monitoring Dosimeter Program for the Pacific Northwest 

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National Laboratory, Richland, Washington, July. [SRDB Ref ID: 127388]

Bivins, S. R., and G. A. Stoetzel, 2009, Area Monitoring Dosimeter Program for the Pacific Northwest 

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National Laboratory, Richland, Washington, August. [SRDB Ref ID: 127393]

Bivins, S. R., and G. A. Stoetzel, 2011, Area Monitoring Dosimeter Program for the Pacific Northwest 

National Laboratory: Results for CY 2010, Battelle Memorial Institute, Pacific Northwest 

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Bivins, S. R., and G. A. Stoetzel, 2012, Area Monitoring Dosimeter Program for the Pacific Northwest 

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National Laboratory, Richland, Washington, September. [SRDB Ref ID: 127395]
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BMI (Battelle Memorial Institute), 2010b, Ultra Trace Analysis Building (3430), PNNL Site Facility Use 

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BMI (Battelle Memorial Institute), 2011a, PSF Trailer A and B (3455 & 3465), PNNL Site Facility Use

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BMI (Battelle Memorial Institute), 2011b, Laboratory Support Warehouse (3475), PNNL Site Facility 

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BMI (Battelle Memorial Institute), 2011c, Radiation Detection Building (3420), PNNL Site Facility Use 

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BMI (Battelle Memorial Institute), 2012a, Environmental Molecular Sciences Laboratory (EMSL), 3020 

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BMI (Battelle Memorial Institute), 2012b, Life Sciences Laboratory II (LSL-II), Chemical Storage & 

Transfer Facility (LSLA), and the Portable Chemical Storage Facility (LSLB) Buildings, RCHN 

Area Facility Use Agreement, FUA-LSLII-R6, Pacific Northwest National Laboratory, Richland, 

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BMI (Battelle Memorial Institute), 2012c, Materials Science and Technology (MS&T), 3410 Building 

and Central Utility Plant (CUP), PNNL Site Facility Use Agreement, FUA-3410/CUP-R4, 

Pacific Northwest National Laboratory, Richland, Washington, December. [SRDB Ref ID: 

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BMI (Battelle Memorial Institute), 2012d, Research Technology Laboratory (RTL-520) Building, RCHN 

Area (Including RTL-510, 524, 530, & 560) Facility Use Agreement, FUA-RTL520-R8, Pacific 

Northwest National Laboratory, Richland, Washington, August. [SRDB Ref ID: 127607]

BMI (Battelle Memorial Institute), 2013a, Large Detector Laboratory (3440), PNNL Site Facility Use 

Agreement, FUA-3440-R1, Pacific Northwest National Laboratory, Richland, Washington, 

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BMI (Battelle Memorial Institute), 2013b, Radiological Control Program Description, Pacific Northwest 

National Laboratory, Richland, Washington, April. [SRDB Ref ID: 127591]

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Diediker, L. P., D. J. Rokkan, K. Rhoads, and L. H. Staven, 2006, Radionuclide Air Emissions Report 

for the Hanford Site, Calendar Year 2005, DOE/RL-2006-01, Rev. 0, U.S. Department of 

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for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, DOE-ST-1027-92 

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HPS (Health Physics Society), 2001, An American National Standard, Design of Internal Dosimetry 

Programs, ANSI/HPS N13.39-2001, American National Standards Institute, New York, New 

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Calibrating Dosemeters and Doserate Meters and for Determining Their Response as a 

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GLOSSARY

absorption type

Categories for materials according to their rates of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization).

air sampling

Collection of samples of the ambient atmosphere to detect or measure the presence of 

radioactive material in the air.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

derived air concentration (DAC)

Annual limit of intake of a radionuclide divided by the volume of air inhaled by Reference Man 

in a working year (2.4 × 103 cubic meters). A DAC-hour is the exposure to a person breathing 

the DAC for 1 hour. 

dose reconstruction 

Process of analyzing the available information including evaluation of historical methods and 

data to estimate the dose a person could have received from one or more radiation exposures.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. 

dosimetry

Measurement and calculation of internal and external radiation doses.

effluent

Liquid or gaseous waste released to the environment.

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 
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some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. For 

example, the isotope of uranium that contains 92 protons and 143 neutrons can appear as 

235U, U-235, or uranium-235. See periodic table of the elements and radioactive isotope. 

enriched uranium (EU)

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.

enrichment

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in 

235U. See depleted uranium, enriched uranium, and tails. 

external dose

Dose received from radiation emitted by sources outside the body.

fissionable

Capable of undergoing fission by capturing neutrons, including fast neutrons. Uranium-238 is 

fissionable. Fissionable indicates both spontaneous and induced fission.

geometric mean (GM)

The nth root of the product of all the members of a set of positive numbers, where n is the 

number of members. The sample geometric mean is designed for averaging ratio or 

proportion data. It is equivalent to taking logarithms of the sample values (i.e., transforming 

the sample), finding the arithmetic mean of the logs, and then retransforming back to the 

original scale (by taking antilogs). It can only be used when all the sample values are greater 

than zero.

glovebox

Enclosure with special rubber gloves through which an operator can handle radioactive or 

toxic material without risk of injury or contamination, normally operated at a slightly reduced 

pressure so that air leakage, if any, is inward.

high-efficiency particulate air (HEPA) filter

Dense filter that removes contaminants from air flows before return to the working environment 

or discharge to the outside air (exhaust).

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

Measurements of radioactive material in the human body using instrumentation that detects 

radiation emitted from the radioactive material in the body.
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internal dose

Dose received from radioactive material in the body.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element.

lower limit of detection (LLD)

See minimum detectable activity.

mass spectrometer

Instrument that determines the relative concentrations of elements or molecules of various 

masses through electromagnetic separation of ions (spectrometry).

minimum detectable activity (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error). See action level, decision level, and minimum reporting level.

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

periodic table of the elements

Arrangement of the chemical elements in order of increasing atomic number from left to right 

and by similar chemical properties vertically. Elements of similar properties occur one under 

the other, which yields groups or families of elements.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

progeny

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products. See parent.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radioactive

Of, caused by, or exhibiting radioactivity.

radioactive isotope

Natural or synthetic form of an atom that emits radioactivity when it decays. See isotope.
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radionuclide

Radioactive nuclide. See radioactive and nuclide.

recycled uranium (RU)

Uranium first irradiated in a reactor then recovered through chemical separation and 

purification. RU contains minor amounts of transuranic material (e.g., plutonium and 

neptunium) and fission products (e.g., technetium) or uranium products (e.g., 236U) after 

purification.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

sealed source

Radioactive material encased in a capsule designed to prevent leakage or escape of the

material.

transuranic elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium.

tritium (hydrogen-3, 3

H, T)

Radioactive isotope of hydrogen that contains one proton and two neutrons in its nucleus. It 

decays by beta emission and has a radioactive half-life of about 12.5 years. 

vitrification

Incorporation of radioactive waste products, particularly from nuclear fuel reprocessing, into 

glass. Also called glassification.